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1

Nuclear Fuels: Promise and Limitations  

SciTech Connect (OSTI)

From 1950 through 1980, scientists, engineers and national leaders confidently predicted an early twenty-first century where fast breeder reactors and commercial nuclear fuel reprocessing were commonplace. Such a scenario seemed necessary for a world with the more than 1000 GWe of nuclear energy needed to meet such an ever-increasing thirst for energy. Thirty years later uranium reserves are increasing on pace with consumption, the growth of nuclear power has been slowed, commercial breeder reactors have yet to enter the marketplace, and less than a handful of commercial reprocessing plants operate. As Nobel Laureate Niels Bohr famously said, “Prediction is very difficult, especially if it’s about the future.” The programme for IChemE’s 2012 conference on the nuclear fuel cycle features a graphic of an idealized nuclear fuel cycle that symbolizes the quest for a closed nuclear fuel cycle featuring careful husbanding of precious resources while minimizing the waste footprint. Progress toward achieving this ideal has been disrupted by technology innovations in the mining and petrochemical industries, as well as within the nuclear industry.

Harold F. McFarlane

2012-03-01T23:59:59.000Z

2

Use of Integrated Decay Heat Limits to Facilitate Spent Nuclear Fuel Loading to Yucca Mountain  

SciTech Connect (OSTI)

As an alternative to the use of the linear loading or areal power density (APD) concept, using integrated decay heat limits based on the use of mountain-scale heat transfer analysis is considered to represent the thermal impact from the deposited spent nuclear fuel (SNF) to the Yucca Mountain repository. Two different integrated decay heat limits were derived to represent both the short-term (up to 50 years from the time of repository closure) and the long-term decay heat effect (up to 1500 years from the time of repository closure). The derived limits were found to appropriately represent the drift wall temperature limit (200 deg. C) and the midway between adjacent drifts temperature limit (96 deg. C) as long as used fuel is uniformly loaded into the mountain. These limits can be a useful practical guide to facilitate the loading of used fuel into Yucca Mountain. (authors)

Li, Jun; Yim, Man-Sung; McNelis, David [Department of Nuclear Engineering, North Carolina State University (United States); Piet, Steven [Idaho National Laboratory (United States)

2007-07-01T23:59:59.000Z

3

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

4

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

5

Critical Analysis of Dry Storage Temperature Limits for Zircaloy-Clad Spent Nuclear Fuel Based on Diffusion Controlled Cavity Growth  

SciTech Connect (OSTI)

Interim dry storage of spent nuclear fuel (SNF) rods is of critical concern because a shortage of existing SNF wet storage capacity combined with delays in the availability of a permanent disposal repository has led to an increasing number of SNF rods being placed into interim dry storage. Safe interim dry storage must be maintained for a minimum of twenty years according to the Standard Review Plan for Dry Cask Storage Systems [1] and the Code of Federal Regulations, 10 CFR Part 72 [2]. Interim dry storage licensees must meet certain safety conditions when storing SNF rods to ensure that there is a ''very low probability (e.g. 0.5%) of cladding breach during long-term storage'' [1]. Commercial SNF typically consists of uranium oxide pellets surrounded by a thin cladding. The cladding is usually an {alpha}-zirconium based alloy know as ''Zircaloy''. In dry storage, the SNF rods are confined in one of several types of cask systems approved by the Nuclear Regulatory Commission (NRC). ''The cask system must be designed to prevent degradation of fuel cladding that results in a type of cladding breach, such as axial-splits or ductile fracture, where irradiated UO{sub 2} particles may be released. In addition, the fuel cladding should not degrade to the point where more than one percent of the fuel rods suffer pinhole or hairline crack type failure under normal storage conditions [1].'' The NRC has approved two models [3,4] for use by proposed dry storage licensees to determine the maximum initial temperature limit for nuclear fuel rods in dry storage that supposedly meet the above criteria and yield consistent temperature limits. Though these two models are based on the same fundamental failure theory, different assumptions have been made including the choice of values for material constants in the failure equation. This report will examine and compare the similarities and inconsistencies of these two models. It will illustrate some of the shortcomings of the current models and suggest modifications as well as some experiments that should be started in the near future. This report will also discuss changes in the current NRC standards with regard to the adoption of a strain-based model to be used to determine maximum allowable temperatures of the SNF.

Hayes, T.A.; Rosen, R.S.; Kassner, M.E.

1999-12-01T23:59:59.000Z

6

What is spent nuclear fuel?  

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

What is Spent Nuclear Fuel? Spent nuclear fuel (SNF) is irradiated fuel or targets containing uranium, plutonium, or thorium that is permanently withdrawn from a nuclear reactor or...

7

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...

8

National Spent Nuclear Fuel Program  

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

need to safely and efficiently manage all DOE-owned spent nuclear fuel and high level waste and prepare it for disposal. The National Spent Nuclear Fuel Program is addressing...

9

Nuclear fuel electrorefiner  

DOE Patents [OSTI]

The present invention relates to a nuclear fuel electrorefiner having a vessel containing a molten electrolyte pool floating on top of a cadmium pool. An anodic fuel dissolution basket and a high-efficiency cathode are suspended in the molten electrolyte pool. A shroud surrounds the fuel dissolution basket and the shroud is positioned so as to separate the electrolyte pool into an isolated electrolyte pool within the shroud and a bulk electrolyte pool outside the shroud. In operation, unwanted noble-metal fission products migrate downward into the cadmium pool and form precipitates where they are removed by a filter and separator assembly. Uranium values are transported by the cadmium pool from the isolated electrolyte pool to the bulk electrolyte pool, and then pass to the high-efficiency cathode where they are electrolytically deposited thereto.

Ahluwalia, Rajesh K.; Hua, Thanh Q.

2004-02-10T23:59:59.000Z

10

FUEL CASK IMPACT LIMITER VULNERABILITIES  

SciTech Connect (OSTI)

Cylindrical fuel casks often have impact limiters surrounding just the ends of the cask shaft in a typical 'dumbbell' arrangement. The primary purpose of these impact limiters is to absorb energy to reduce loads on the cask structure during impacts associated with a severe accident. Impact limiters are also credited in many packages with protecting closure seals and maintaining lower peak temperatures during fire events. For this credit to be taken in safety analyses, the impact limiter attachment system must be shown to retain the impact limiter following Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) impacts. Large casks are often certified by analysis only because of the costs associated with testing. Therefore, some cask impact limiter attachment systems have not been tested in real impacts. A recent structural analysis of the T-3 Spent Fuel Containment Cask found problems with the design of the impact limiter attachment system. Assumptions in the original Safety Analysis for Packaging (SARP) concerning the loading in the attachment bolts were found to be inaccurate in certain drop orientations. This paper documents the lessons learned and their applicability to impact limiter attachment system designs.

Leduc, D; Jeffery England, J; Roy Rothermel, R

2009-02-09T23:59:59.000Z

11

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

12

Modeling the Nuclear Fuel Cycle  

SciTech Connect (OSTI)

A review of existing nuclear fuel cycle systems analysis codes was performed to determine if any existing codes meet technical and functional requirements defined for a U.S. national program supporting the global and domestic assessment, development and deployment of nuclear energy systems. The program would be implemented using an interconnected architecture of different codes ranging from the fuel cycle analysis code, which is the subject of the review, to fundamental physical and mechanistic codes. Four main functions are defined for the code: (1) the ability to characterize and deploy individual fuel cycle facilities and reactors in a simulation, while discretely tracking material movements, (2) the capability to perform an uncertainty analysis for each element of the fuel cycle and an aggregate uncertainty analysis, (3) the inclusion of an optimization engine able to optimize simultaneously across multiple objective functions, and (4) open and accessible code software and documentation to aid in collaboration between multiple entities and facilitate software updates. Existing codes, categorized as annualized or discrete fuel tracking codes, were assessed according to the four functions and associated requirements. These codes were developed by various government, education and industrial entities to fulfill particular needs. In some cases, decisions were made during code development to limit the level of detail included in a code to ease its use or to focus on certain aspects of a fuel cycle to address specific questions. The review revealed that while no two of the codes are identical, they all perform many of the same basic functions. No code was able to perform defined function 2 or several requirements of functions 1 and 3. Based on this review, it was concluded that the functions and requirements will be met only with development of a new code, referred to as GENIUS.

Jacob J. Jacobson; Mary Lou Dunzik-Gougar; Christopher A. Juchau

2010-08-01T23:59:59.000Z

13

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...

14

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

15

Advanced Nuclear Fuel Cycle Options  

SciTech Connect (OSTI)

A systematic evaluation has been conducted of the potential for advanced nuclear fuel cycle strategies and options to address the issues ascribed to the use of nuclear power. Issues included nuclear waste management, proliferation risk, safety, security, economics and affordability, and sustainability. The two basic strategies, once-through and recycle, and the range of possibilities within each strategy, are considered for all aspects of the fuel cycle including options for nuclear material irradiation, separations if needed, and disposal. Options range from incremental changes to today’s implementation to revolutionary concepts that would require the development of advanced nuclear technologies.

Roald Wigeland; Temitope Taiwo; Michael Todosow; William Halsey; Jess Gehin

2010-06-01T23:59:59.000Z

16

Nuclear Fuels | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the ContributionsArms Control R&D Consortium includesEnergy Nuclear Fuels

17

6 Nuclear Fuel Designs  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLasDelivered‰PNGExperience hands-onASTROPHYSICS H.CarbonMarchTHEMaterials and1663 January

18

Risk and Responsibility Sharing in Nuclear Spent Fuel Management  

E-Print Network [OSTI]

With the Nuclear Waste Policy Act of 1982, the responsibility of American utilities in the long-term management of spent nuclear fuel was limited to the payment of a fee. This narrow involvement did not result in faster ...

De Roo, Guillaume

19

Spent Nuclear Fuel (SNF) Project Product Specification  

SciTech Connect (OSTI)

This document establishes the limits and controls for the significant parameters that could potentially affect the safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for processing, transport, and storage. The product specifications in this document cover the SNF packaged in Multi-Canister Overpacks to be transported throughout the SNF Project.

PAJUNEN, A.L.

2000-01-20T23:59:59.000Z

20

The Effectiveness of Full Actinide Recycle as a Nuclear Waste Management Strategy when Implemented over a Limited Timeframe – Part II: Thorium Fuel Cycle  

E-Print Network [OSTI]

water reactors (LWRs) or sodium-cooled fast reactors (SFRs) is considered for uranium (U) fuel cycles. With full actinide recycling, at least 6 generations of SFRs are required in a gradual phase-out of nuclear power to achieve transmutation performance...

Lindley, Benjamin A.; Fiorina, Carlo; Gregg, Robert; Franceschini, Fausto; Parks, Geoffrey T.

2014-12-06T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear fuels limited" 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

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 has developed for a ready source or information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained has been obtained from nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops; and so forth. Sources do not agree completely with each other, and the data listed herein does not reflect any one single source but frequently is consolidation/combination of information. Lack of space as well as the intent and purpose of the Fact Book limit the given information to that pertaining to the Nuclear Fuel Cycle and to data considered of primary interest or most helpful to the majority of users.

Leigh, I.W.

1988-01-01T23:59:59.000Z

22

Proliferation Resistant Nuclear Reactor Fuel  

SciTech Connect (OSTI)

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

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

2011-02-18T23:59:59.000Z

23

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

24

Transportation capabilities study of DOE-owned spent nuclear fuel  

SciTech Connect (OSTI)

This study evaluates current capabilities for transporting spent nuclear fuel owned by the US Department of Energy. Currently licensed irradiated fuel shipping packages that have the potential for shipping the spent nuclear fuel are identified and then matched against the various spent nuclear fuel types. Also included are the results of a limited investigation into other certified packages and new packages currently under development. This study is intended to support top-level planning for the disposition of the Department of Energy`s spent nuclear fuel inventory.

Clark, G.L.; Johnson, R.A.; Smith, R.W. [Packaging Technology, Inc., Tacoma, WA (United States); Abbott, D.G.; Tyacke, M.J. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

1994-10-01T23:59:59.000Z

25

Spent nuclear fuel sampling strategy  

SciTech Connect (OSTI)

This report proposes a strategy for sampling the spent nuclear fuel (SNF) stored in the 105-K Basins (105-K East and 105-K West). This strategy will support decisions concerning the path forward SNF disposition efforts in the following areas: (1) SNF isolation activities such as repackaging/overpacking to a newly constructed staging facility; (2) conditioning processes for fuel stabilization; and (3) interim storage options. This strategy was developed without following the Data Quality Objective (DQO) methodology. It is, however, intended to augment the SNF project DQOS. The SNF sampling is derived by evaluating the current storage condition of the SNF and the factors that effected SNF corrosion/degradation.

Bergmann, D.W.

1995-02-08T23:59:59.000Z

26

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

27

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

28

Categorization of Used Nuclear Fuel Inventory in Support of a...  

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

Categorization of Used Nuclear Fuel Inventory in Support of a Comprehensive National Nuclear Fuel Cycle Strategy Categorization of Used Nuclear Fuel Inventory in Support of a...

29

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

30

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network [OSTI]

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

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

31

Impact of Nuclear Energy Futures on Advanced Fuel Cycle Options  

SciTech Connect (OSTI)

The Nuclear Waste Policy Act requires the Secretary of Energy to inform Congress before 2010 on the need for a second geologic repository for spent nuclear fuel. By that time, the spent fuel discharged from current commercial reactors will exceed the statutory limit of the first repository. There are several approaches to eliminate the need for another repository in this century. This paper presents a high-level analysis of these spent fuel management options in the context of a full range of possible nuclear energy futures. The analysis indicates the best option to implement varies depending on the nuclear energy future selected.

Dixon, B.W.; Piet, S.J.

2004-10-03T23:59:59.000Z

32

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.

33

Simulated nuclear reactor fuel assembly  

DOE Patents [OSTI]

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

Berta, Victor T. (Idaho Falls, ID)

1993-01-01T23:59:59.000Z

34

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

35

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

36

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network [OSTI]

capacity and operating efficiency of nuclear plants [31,operating efficiency of nuclear plants in the past decades.cost of the fuel Nuclear Plant Capacity Factor Nuclear

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

37

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

38

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

39

Nuclear Fuel Cycle | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergyDepartment of Energy Advanced1, 2014Nuclear Facilities Nuclear

40

Nuclear Fuel Cycle | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergyDepartment of Energy Advanced1, 2014Nuclear Facilities NuclearCycle

Note: This page contains sample records for the topic "nuclear fuels limited" 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 Fuel Facts: Uranium | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergyDepartment of Energy Advanced1, 2014Nuclear Facilities NuclearCycleFacts:

42

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 has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was 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 OECN/NEA activities reports; not reflect any one single source but frequently represent a consolidation/combination of information.

Leigh, I.W.; Patridge, M.D.

1991-05-01T23:59:59.000Z

43

Annotated Bibliography for Drying Nuclear Fuel  

SciTech Connect (OSTI)

Internationally, the nuclear industry is represented by both commercial utilities and research institutions. Over the past two decades many of these entities have had to relocate inventories of spent nuclear fuel from underwater storage to dry storage. These efforts were primarily prompted by two factors: insufficient storage capacity (potentially precipitated by an open-ended nuclear fuel cycle) or deteriorating quality of existing underwater facilities. The intent of developing this bibliography is to assess what issues associated with fuel drying have been identified, to consider where concerns have been satisfactorily addressed, and to recommend where additional research would offer the most value to the commercial industry and the U. S. Department of Energy.

Rebecca E. Smith

2011-09-01T23:59:59.000Z

44

Nuclear Symbiosis - A Means to Achieve Sustainable Nuclear Growth while Limiting the Spread of Sensititive Nuclear Technology  

SciTech Connect (OSTI)

Global growth of nuclear energy in the 21st century is creating new challenges to limit the spread of nuclear technology without hindering adoption in countries now considering nuclear power. Independent nuclear states desire autonomy over energy choices and seek energy independence. However, this independence comes with high costs for development of new indigenous fuel cycle capabilities. Nuclear supplier states and expert groups have proposed fuel supply assurance mechanisms such as fuel take-back services, international enrichment services and fuel banks in exchange for recipient state concessions on the development of sensitive technologies. Nuclear states are slow to accept any concessions to their rights under the Non-Proliferation Treaty. To date, decisions not to develop indigenous fuel cycle capabilities have been driven primarily by economics. However, additional incentives may be required to offset a nuclear state’s perceived loss of energy independence. This paper proposes alternative economic development incentives that could help countries decide to forgo development of sensitive nuclear technologies. The incentives are created through a nuclear-centered industrial complex with “symbiotic” links to indigenous economic opportunities. This paper also describes a practical tool called the “Nuclear Materials Exchange” for identifying these opportunities.

David Shropshire

2009-09-01T23:59:59.000Z

45

apex nuclear fuel: Topics by E-print Network  

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

fuel fabrication, in-core physics and fuel management of uranium, thorium and other fuel types, ... Kazimi, Mujid S. 19 Nuclear Waste Imaging and Spent Fuel Verification by...

46

Composite construction for nuclear fuel containers  

DOE Patents [OSTI]

An improved method for producing nuclear fuel containers of a composite construction having components providing therein a barrier system for resisting destructive action by volatile fission products or impurities and also interdiffusion of metal constituents, and the product thereof. The composite nuclear fuel containers of the method comprise a casing of zirconium or alloy thereof with a layer of copper overlying an oxidized surface portion of the zirconium or alloy thereof.

Cheng, Bo-Ching (Fremont, CA); Rosenbaum, Herman S. (Fremont, CA); Armijo, Joseph S. (Saratoga, CA)

1987-01-01T23:59:59.000Z

47

Review of Used Nuclear Fuel Storage and Transportation Technical...  

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

action based on the comparison. Review of Used Nuclear Fuel Storage and Transportation Technical Gap Analysis More Documents & Publications Review of Used Nuclear Fuel...

48

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,...

49

Influence of Nuclear Fuel Cycles on Uncertainty of Long Term...  

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

Influence of Nuclear Fuel Cycles on Uncertainty of Long Term Performance of Geologic Disposal Systems Influence of Nuclear Fuel Cycles on Uncertainty of Long Term Performance of...

50

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

51

Dry Processing of Used Nuclear Fuel  

SciTech Connect (OSTI)

Dry (non-aqueous) separations technologies have been used for treatment of used nuclear fuel since the 1960s, and they are still being developed and demonstrated in many countries. Dry technologies offer potential advantages compared to traditional aqueous separations including: compactness, resistance to radiation effects, criticality control benefits, compatibility with advanced fuel types, and ability to produce low purity products. Within the Department of Energy’s Advanced Fuel Cycle Initiative, an electrochemical process employing molten salts is being developed for recycle of fast reactor fuel and treatment of light water reactor oxide fuel to produce a feed for fast reactors. Much of the development of this technology is based on treatment of used Experimental Breeder Reactor II (EBR-II) fuel, which is metallic. Electrochemical treatment of the EBR-II fuel has been ongoing in the Fuel Conditioning Facility, located at the Materials and Fuel Complex of Idaho National Laboratory since 1996. More than 3.8 metric tons of heavy metal of metallic fast reactor fuel have been treated using this technology. This paper will summarize the status of electrochemical development and demonstration activities with used nuclear fuel, including high-level waste work. A historic perspective on the background of dry processing will also be provided.

K. M. Goff; M. F. Simpson

2009-09-01T23:59:59.000Z

52

PWR Fuel Shipping Limits & RNP Core Design  

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

Excellent Operational Record No radioactive spills or releases No radiation exposure to train personnel or general public 8 Safety - Success Factors Robust package design Nuclear...

53

Renewable Fuels Limited RFL | Open Energy Information  

Open Energy Info (EERE)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia:FAQ < RAPID Jump to: navigation, searchVirginia Blue Ridge AndREII Jump to:RFSL Jump to: navigation,RFL Jump

54

advanced nuclear fuels: Topics by E-print Network  

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

Fuel Cycles University of California eScholarship Repository Summary: uranium or thorium ores and production of nuclear fuel, anynuclear fuel strontium Sievert Trivalent...

55

advanced nuclear fuel: Topics by E-print Network  

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

Fuel Cycles University of California eScholarship Repository Summary: uranium or thorium ores and production of nuclear fuel, anynuclear fuel strontium Sievert Trivalent...

56

Dry Transfer Systems for Used Nuclear Fuel  

SciTech Connect (OSTI)

The potential need for a dry transfer system (DTS) to enable retrieval of used nuclear fuel (UNF) for inspection or repackaging will increase as the duration and quantity of fuel in dry storage increases. This report explores the uses for a DTS, identifies associated general functional requirements, and reviews existing and proposed systems that currently perform dry fuel transfers. The focus of this paper is on the need for a DTS to enable transfer of bare fuel assemblies. Dry transfer systems for UNF canisters are currently available and in use for transferring loaded canisters between the drying station and storage and transportation casks.

Brett W. Carlsen; Michaele BradyRaap

2012-05-01T23:59:59.000Z

57

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

58

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 has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was 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/NEA activities reports; proceedings of conferences and workshops, etc. The data listed do not reflect any one single source but frequently represent a consolidation/combination of information.

Leigh, I W; Mitchell, S J

1990-01-01T23:59:59.000Z

59

International nuclear fuel cycle fact book. [Contains glossary  

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 has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained has been obtained from nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops; and so forth. Sources do not agree completely with each other, and the data listed herein does not reflect any one single source but frequently is a consolidation/combination of information. Lack of space as well as the intent and purpose of the Fact Book limit the given information to that pertaining to the Nuclear Fuel Cycle and to data considered of primary interest or most helpful to the majority of users.

Leigh, I.W.; Lakey, L.T.; Schneider, K.J.; Silviera, D.J.

1987-01-01T23:59:59.000Z

60

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

Note: This page contains sample records for the topic "nuclear fuels limited" 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

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

62

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

63

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.degree. 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.

Heubeck, Norman B. (Schenectady, NY)

1998-01-01T23:59:59.000Z

64

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

65

The limits of the nuclear landscape  

SciTech Connect (OSTI)

In 2011, 100 new nuclides were discovered1. They joined the approximately 3,000 stable and radioactive nuclides that either occur naturally on Earth or are synthesized in the laboratory2,3. Every atomic nucleus, characterized by a specific number of protons and neutrons, occupies a spot on the chart of nuclides, which is bounded by drip lines indicating the values of neutron and proton number at which nuclear binding ends. The placement of the neutron drip line for the heavier elements is based on theoretical predictions using extreme extrapolations, and so is uncertain. However, it is not known how uncertain it is or how many protons and neutrons can be bound in a nucleus. Here we estimate these limits of the nuclear landscape and provide statistical and systematic uncertainties for our predictions. We use nuclear density functional theory, several Skyrme interactions and high-performance computing, and find that the number of bound nuclides with between 2 and 120 protons is around 7,000. We find that extrapolations for drip-line positions and selected nuclear properties, including neutron separation energies relevant to astrophysical processes, are very consistent between the models used.

Nazarewicz, Witold [ORNL; Erler, J. [Oak Ridge National Laboratory (ORNL); Birge, N. [University of Tennessee, Knoxville (UTK); Kortelainen, E. M. [University of Tennessee (UTK) and Oak Ridge National Laboratory (ORNL); Olsen, E. [University of Tennessee, Knoxville (UTK); Perhac, A. [University of Tennessee, Knoxville (UTK); Stoitsov, M. [University of Tennessee (UTK) and Oak Ridge National Laboratory (ORNL)

2012-01-01T23:59:59.000Z

66

Locking support for nuclear fuel assemblies  

DOE Patents [OSTI]

A locking device for supporting and locking a nuclear fuel assembly within a cylindrical bore formed by a support plate, the locking device including a support and locking sleeve having upwardly extending fingers forming wedge shaped contact portions arranged for interaction between an annular tapered surface on the fuel assembly and the support plate bore as well as downwardly extending fingers having wedge shaped contact portions arranged for interaction between an annularly tapered surface on the support plate bore and the fuel assembly whereby the sleeve tends to support and lock the fuel assembly in place within the bore by its own weight while facilitating removal and/or replacement of the fuel assembly.

Ledin, Eric (San Diego, CA)

1980-01-01T23:59:59.000Z

67

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

68

Spent Nuclear Fuel (SNF) Project Product Specification  

SciTech Connect (OSTI)

The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major sub-system. Section 4.0--Specific technical basis description for each product specification. The scope of this product specification does not include data collection requirements to support accountability or environmental compliance activities.

PAJUNEN, A.L.

2000-12-07T23:59:59.000Z

69

Impact of Nuclear Energy Futures on Advanced Fuel Cycle Options  

SciTech Connect (OSTI)

The Nuclear Waste Policy Act requires the Secretary of Energy to inform Congress before 2010 on the need for a second geologic repository for spent nuclear fuel. By that time, the spent fuel discharged from current commercial reactors will exceed the statutory limit of the first repository (63,000 MTiHM commercial, 7,000 MT non-commercial). There are several approaches to eliminate the need for another repository in this century. This paper presents a high-level analysis of these spent fuel management options in the context of a full range of possible nuclear energy futures. The analysis indicates the best option to implement varies depending on the nuclear energy future selected. The first step in understanding the need for different spent fuel management approaches is to understand the size of potential spent fuel inventories. A full range of potential futures for domestic commercial nuclear energy is considered. These energy futures are as follows: 1. Existing License Completion - Based on existing spent fuel inventories plus extrapolation of future plant-by-plant discharges until the end of each operating license, including known license extensions. 2. Extended License Completion - Based on existing spent fuel inventories plus a plant-by-plant extrapolation of future discharges assuming on all operating plants having one 20-year extension. 3. Continuing Level Energy Generation - Based on extension of the current ~100 GWe installed commercial base and average spent fuel discharge of 2100 MT/yr through the year 2100. 4. Continuing Market Share Generation – Based on a 1.8% compounded growth of the electricity market through the year 2100, matched by growing nuclear capacity and associated spent fuel discharge. 5. Growing Market Share Generation - Extension of current nuclear capacity and associated spent fuel discharge through 2100 with 3.2% growth representing 1.5% market growth (all energy, not just electricity) and 1.7% share growth. Share growth results in tripling market share by 2100 from the current 8.4% to 25%, equivalent to continuing the average market growth of last 50 years for an additional 100 years. Five primary spent fuel management strategies are assessed against each of the energy futures to determine the number of geological repositories needed and how the first repository would be used. The geological repository site at Yucca Mountain, Nevada, has the physical potential to accommodate all the spent fuel that will be generated by the current fleet of domestic commercial nuclear reactors, even with license extensions. If new nuclear plants are built in the future as replacements or additions, the United States will need to adopt spent fuel treatment to extend the life of the repository. Should a significant number of new nuclear plants be built, advanced fuel recycling will be needed to fully manage the spent fuel within a single repository. The analysis also considers the timeframe for most efficient implementation of new spent fuel management strategies. The mix of unprocessed spent fuel and processed high level waste in Yucca Mountain varies with each future and strategy. Either recycling must start before there is too much unprocessed waste emplaced or unprocessed waste will have to be retrieved later with corresponding costs. For each case, the latest date to implement reprocessing without subsequent retrieval is determined.

Brent W. Dixon; Steven J. Piet

2004-10-01T23:59:59.000Z

70

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

Office of Environmental Management (EM)

Nuclear Fuel Storage and Transportation Planning Project Overview DOE Office of Nuclear Energy Task Force for Strategic Developments to Blue Ribbon Commission Recommendations...

71

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

Energy Savers [EERE]

Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport - Demonstration of Approach and Results of Used Fuel Performance Characterization Used...

72

Financing Strategies for Nuclear Fuel Cycle Facility  

SciTech Connect (OSTI)

To help meet our nation’s energy needs, reprocessing of spent nuclear fuel is being considered more and more as a necessary step in a future nuclear fuel cycle, but incorporating this step into the fuel cycle will require considerable investment. This report presents an evaluation of financing scenarios for reprocessing facilities integrated into the nuclear fuel cycle. A range of options, from fully government owned to fully private owned, was evaluated using a DPL (Dynamic Programming Language) 6.0 model, which can systematically optimize outcomes based on user-defined criteria (e.g., lowest life-cycle cost, lowest unit cost). Though all business decisions follow similar logic with regard to financing, reprocessing facilities are an exception due to the range of financing options available. The evaluation concludes that lowest unit costs and lifetime costs follow a fully government-owned financing strategy, due to government forgiveness of debt as sunk costs. Other financing arrangements, however, including regulated utility ownership and a hybrid ownership scheme, led to acceptable costs, below the Nuclear Energy Agency published estimates. Overwhelmingly, uncertainty in annual capacity led to the greatest fluctuations in unit costs necessary for recovery of operating and capital expenditures; the ability to determine annual capacity will be a driving factor in setting unit costs. For private ventures, the costs of capital, especially equity interest rates, dominate the balance sheet; the annual operating costs dominate the government case. It is concluded that to finance the construction and operation of such a facility without government ownership could be feasible with measures taken to mitigate risk, and that factors besides unit costs should be considered (e.g., legal issues, social effects, proliferation concerns) before making a decision on financing strategy.

David Shropshire; Sharon Chandler

2005-12-01T23:59:59.000Z

73

Standard guide for drying behavior of spent nuclear fuel  

E-Print Network [OSTI]

1.1 This guide is organized to discuss the three major components of significance in the drying behavior of spent nuclear fuel: evaluating the need for drying, drying spent nuclear fuel, and confirmation of adequate dryness. 1.1.1 The guide addresses drying methods and their limitations in drying spent nuclear fuels that have been in storage at water pools. The guide discusses sources and forms of water that remain in SNF, its container, or both, after the drying process and discusses the importance and potential effects they may have on fuel integrity, and container materials. The effects of residual water are discussed mechanistically as a function of the container thermal and radiological environment to provide guidance on situations that may require extraordinary drying methods, specialized handling, or other treatments. 1.1.2 The basic issue in drying is to determine how dry the SNF must be in order to prevent issues with fuel retrievability, container pressurization, or container corrosion. Adequate d...

American Society for Testing and Materials. Philadelphia

2008-01-01T23:59:59.000Z

74

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

75

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...

76

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

Office of Environmental Management (EM)

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...

77

Summary of nuclear fuel reprocessing activities around the world  

SciTech Connect (OSTI)

This review of international practices for nuclear fuel reprocessing was prepared to provide a nontechnical summary of the current status of nuclear fuel reprocessing activities around the world. The sources of information are widely varied.

Mellinger, P.J.; Harmon, K.M.; Lakey, L.T.

1984-11-01T23:59:59.000Z

78

HIGH EFFICIENCY GENERATION OF HYDROGEN FUELS USING NUCLEAR POWER  

SciTech Connect (OSTI)

OAK B202 HIGH EFFICIENCY GENERATION OF HYDROGEN FUELS USING NUCLEAR POWER. Combustion of fossil fuels, used to power transportation, generate electricity, heat homes and fuel industry provides 86% of the world's energy. Drawbacks to fossil fuel utilization include limited supply, pollution, and carbon dioxide emissions. Carbon dioxide emissions, thought to be responsible for global warming, are now the subject of international treaties. Together, these drawbacks argue for the replacement of fossil fuels with a less-polluting potentially renewable primary energy such as nuclear energy. Conventional nuclear plants readily generate electric power but fossil fuels are firmly entrenched in the transportation sector. Hydrogen is an environmentally attractive transportation fuel that has the potential to displace fossil fuels. Hydrogen will be particularly advantageous when coupled with fuel cells. Fuel cells have higher efficiency than conventional battery/internal combustion engine combinations and do not produce nitrogen oxides during low-temperature operation. Contemporary hydrogen production is primarily based on fossil fuels and most specifically on natural gas. When hydrogen is produced using energy derived from fossil fuels, there is little or no environmental advantage. There is currently no large scale, cost-effective, environmentally attractive hydrogen production process available for commercialization, nor has such a process been identified. The objective of this work is to find an economically feasible process for the production of hydrogen, by nuclear means, using an advanced high-temperature nuclear reactor as the primary energy source. Hydrogen production by thermochemical water-splitting (Appendix A), a chemical process that accomplishes the decomposition of water into hydrogen and oxygen using only heat or, in the case of a hybrid thermochemical process, by a combination of heat and electrolysis, could meet these goals. Hydrogen produced from fossil fuels has trace contaminants (primarily carbon monoxide) that are detrimental to precious metal catalyzed fuel cells, as is now recognized by many of the world's largest automobile companies. Thermochemical hydrogen will not contain carbon monoxide as an impurity at any level. Electrolysis, the alternative process for producing hydrogen using nuclear energy, suffers from thermodynamic inefficiencies in both the production of electricity and in electrolytic parts of the process. The efficiency of electrolysis (electricity to hydrogen) is currently about 80%. Electric power generation efficiency would have to exceed 65% (thermal to electrical) for the combined efficiency to exceed the 52% (thermal to hydrogen) calculated for one thermochemical cycle. Thermochemical water-splitting cycles have been studied, at various levels of effort, for the past 35 years. They were extensively studied in the late 70s and early 80s but have received little attention in the past 10 years, particularly in the U.S. While there is no question about the technical feasibility and the potential for high efficiency, cycles with proven low cost and high efficiency have yet to be developed commercially. Over 100 cycles have been proposed, but substantial research has been executed on only a few. This report describes work accomplished during a three-year project whose objective is to ''define an economically feasible concept for production of hydrogen, by nuclear means, using an advanced high temperature nuclear reactor as the energy source.'' The emphasis of the first phase was to evaluate thermochemical processes which offer the potential for efficient, cost-effective, large-scale production of hydrogen from water in which the primary energy input is high temperature heat from an advanced nuclear reactor and to select one (or, at most three) for further detailed consideration. During Phase 1, an exhaustive literature search was performed to locate all cycles previously proposed. The cycles located were screened using objective criteria to determine which could benefit, in terms of efficien

BROWN,LC; BESENBRUCH,GE; LENTSCH,RD; SCHULTZ,KR; FUNK,JF; PICKARD,PS; MARSHALL,AC; SHOWALTER,SK

2003-06-01T23:59:59.000Z

79

Nuclear power generation and fuel cycle report 1996  

SciTech Connect (OSTI)

This report presents the current status and projections through 2015 of nuclear capacity, generation, and fuel cycle requirements for all countries using nuclear power to generate electricity for commercial use. It also contains information and forecasts of developments in the worldwide nuclear fuel market. Long term projections of U.S. nuclear capacity, generation, and spent fuel discharges for two different scenarios through 2040 are developed. A discussion on decommissioning of nuclear power plants is included.

NONE

1996-10-01T23:59:59.000Z

80

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

Note: This page contains sample records for the topic "nuclear fuels limited" 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

Gamma Ray Mirrors for Direct Measurement of Spent Nuclear Fuel  

SciTech Connect (OSTI)

Direct measurement of the amount of Pu and U in spent nuclear fuel represents a challenge for the safeguards community. Ideally, the characteristic gamma-ray emission lines from different isotopes provide an observable suitable for this task. However, these lines are generally lost in the fierce flux of radiation emitted by the fuel. The rates are so high that detector dead times limit measurements to only very small solid angles of the fuel. Only through the use of carefully designed view ports and long dwell times are such measurements possible. Recent advances in multilayer grazing-incidence gamma-ray optics provide one possible means of overcoming this difficulty. With a proper optical and coating design, such optics can serve as a notch filter, passing only narrow regions of the overall spectrum to a fully shielded detector that does not view the spent fuel directly. We report on the design of a mirror system and a number of experimental measurements.

Pivovaroff, Dr. Michael J. [Lawrence Livermore National Laboratory (LLNL)] [Lawrence Livermore National Laboratory (LLNL); Ziock, Klaus-Peter [ORNL] [ORNL; Harrison, Mark J [ORNL] [ORNL; Soufli, Regina [Lawrence Livermore National Laboratory (LLNL)] [Lawrence Livermore National Laboratory (LLNL)

2014-01-01T23:59:59.000Z

82

Pyroprocess for processing spent nuclear fuel  

DOE Patents [OSTI]

This is a pyroprocess for processing spent nuclear fuel. The spent nuclear fuel is chopped into pieces and placed in a basket which is lowered in to a liquid salt solution. The salt is rich in ZrF.sub.4 and containing alkali or alkaline earth fluorides, and in particular, the salt chosen was LiF-50 mol % ZrF.sub.4 with a eutectic melting point of 500.degree. C. Prior to lowering the basket, the salt is heated to a temperature of between 550.degree. C. and 700.degree. C. in order to obtain a molten solution. After dissolution the oxides of U, Th, rare earth and other like oxides, the salt bath solution is subject to hydro-fluorination to remove the oxygen and then to a fluorination step to remove U as gaseous UF.sub.6. In addition, after dissolution, the basket contains PuO.sub.2 and undissolved parts of the fuel rods, and the basket and its contents are processed to remove the Pu.

Miller, William E. (Naperville, IL); Tomczuk, Zygmunt (Lockport, IL)

2002-01-01T23:59:59.000Z

83

Nuclear fuel cycles for mid-century development  

E-Print Network [OSTI]

A comparative analysis of nuclear fuel cycles was carried out. Fuel cycles reviewed include: once-through fuel cycles in LWRs, PHWRs, HTGRs, and fast gas cooled breed and burn reactors; single-pass recycle schemes: plutonium ...

Parent, Etienne, 1977-

2003-01-01T23:59:59.000Z

84

Method and apparatus for close packing of nuclear fuel assemblies  

DOE Patents [OSTI]

The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

Newman, D.F.

1993-03-30T23:59:59.000Z

85

Method and apparatus for close packing of nuclear fuel assemblies  

DOE Patents [OSTI]

The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

Newman, Darrell F. (Richland, WA)

1993-01-01T23:59:59.000Z

86

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

87

Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay  

E-Print Network [OSTI]

Security of the National Nuclear Security Administration, USof Energys National Nuclear Security Administration (NNSA)

Quiter, Brian

2012-01-01T23:59:59.000Z

88

BIOMASS FOR HYDROGEN AND OTHER TRANSPORT FUELS -POTENTIALS, LIMITATIONS & COSTS  

E-Print Network [OSTI]

BIOMASS FOR HYDROGEN AND OTHER TRANSPORT FUELS - POTENTIALS, LIMITATIONS & COSTS Senior scientist - "Towards Hydrogen Society" ·biomass resources - potentials, limits ·biomass carbon cycle ·biomass for hydrogen - as compared to other H2- sources and to other biomass paths #12;BIOMASS - THE CARBON CYCLE

89

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

SciTech Connect (OSTI)

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

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

2012-09-01T23:59:59.000Z

90

Spent Nuclear Fuel Project path forward: nuclear safety equivalency to comparable NRC-licensed facilities  

SciTech Connect (OSTI)

This document includes the Technical requirements which meet the nuclear safety objectives of the NRC regulations for fuel treatment and storage facilities. These include requirements regarding radiation exposure limits, safety analysis, design and construction. This document also includes administrative requirements which meet the objectives of the major elements of the NRC licensing process. These include formally documented design and safety analysis, independent technical review, and oppportunity for public involvement.

Garvin, L.J.

1995-11-01T23:59:59.000Z

91

Spent nuclear fuel project - criteria document spent nuclear fuel final safety analysis report  

SciTech Connect (OSTI)

The criteria document provides the criteria and planning guidance for developing the Spent Nuclear Fuel (SNF) Final Safety Analysis Report (FSAR). This FSAR will support the US Department of Energy, Richland Operations Office decision to authorize the procurement, installation, installation acceptance testing, startup, and operation of the SNF Project facilities (K Basins, Cold Vacuum Drying Facility, and Canister Storage Building).

MORGAN, R.G.

1999-02-23T23:59:59.000Z

92

Commercial Spent Nuclear Fuel Waste Package Misload Analysis  

SciTech Connect (OSTI)

The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M&O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis Department. Before using the results of this calculation, the reader is cautioned to verify that the assumptions made in this calculation regarding the waste stream, the loading process, and the staging of the spent nuclear fuel assemblies are applicable.

A. Alsaed

2005-07-28T23:59:59.000Z

93

Commercial Spent Nuclear Fuel Waste Package Misload Analysis  

SciTech Connect (OSTI)

The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M&O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis Department. Before using the results of this calculation, the reader is cautioned to verify that the assumptions made in this calculation regarding the waste stream, the loading process, and the staging of the spent nuclear fuel assemblies are applicable.

J.K. Knudson

2003-10-02T23:59:59.000Z

94

INVESTIGATION ON THE FLAME EXTINCTION LIMIT OF FUEL BLENDS  

SciTech Connect (OSTI)

Lean flame extinction limits of binary fuel mixtures of methane (CH{sub 4}), propane (C{sub 3}H{sub 8}), and ethane (C{sub 2}H{sub 6}) were measured using a twin-flame counter-flow burner. Experiments were conducted to generate an extinction equivalence ratio vs. global stretch rate plot and an extrapolation method was used to calculate the equivalence ratio corresponding to an experimentally unattainable zero-stretch condition. The foregoing gases were selected because they are the primary constitutes of natural gas, which is the primary focus of the present study. To validate the experimental setup and methodology, the flame extinction limit of pure fuels at zero stretch conditions were also estimated and compared with published values. The lean flame extinction limits of methane (f{sub ext} = 4.6%) and propane (f{sub ext} = 2.25%) flames measured in the present study agreed with the values reported in the literature. It was observed that the flame extinction limit of fuel blends have a polynomial relation with the concentration of component fuels in the mixture. This behavior contradicts with the commonly used linear Le Chatelier's approximation. The experimentally determined polynomial relations between the flame extinction limits of fuel blends (i.e. methane-propane and methane-ethane) and methane concentration are as follows: (1) Methane-Propane--%f{sub ext} = (1.05 x 10{sup -9}) f{sup 5}-(1.3644 x 10{sup -7}) f{sup 4}+(6.40299 x 10{sup -6}) f{sup 3}-(1.2108459 x 10{sup -4}) f{sup 2}+(2.87305329 x 10{sup -3}) f+2.2483; (2) Methane-Ethane--%f{sub ext} = (2.1 x 10{sup -9})f{sup 5}-(3.5752 x 10{sup -7}) f{sup 4}+(2.095425 x 10{sup -5}) f{sup 3}-(5.037353 x 10{sup -4}) f{sup 2} + 6.08980409 f + 2.8923. Where f{sub ext} is the extinction limits of methane-propane and methane-ethane fuel blends, and f is the concentration (% volume) of methane in the fuel mixture. The relations were obtained by fitting fifth order curve (polynomial regression) to experimentally measured extinction limits at different mixture conditions. To extend the study to a commercial fuel, the flame extinction limit for Birmingham natural gas (a blend of 95% methane, 5% ethane and 5% nitrogen) was experimentally determined and was found to be 3.62% fuel in the air-fuel mixture.

Ahsan R. Choudhuri

2005-02-01T23:59:59.000Z

95

Dynamic Systems Analysis Report for Nuclear Fuel Recycle  

SciTech Connect (OSTI)

This report examines the time-dependent dynamics of transitioning from the current United States (U.S.) nuclear fuel cycle where used nuclear fuel is disposed in a repository to a closed fuel cycle where the used fuel is recycled and only fission products and waste are disposed. The report is intended to help inform policy developers, decision makers, and program managers of system-level options and constraints as they guide the formulation and implementation of advanced fuel cycle development and demonstration efforts and move toward deployment of nuclear fuel recycling infrastructure.

Brent Dixon; Sonny Kim; David Shropshire; Steven Piet; Gretchen Matthern; Bill Halsey

2008-12-01T23:59:59.000Z

96

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

97

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

98

Method of manufacturing nuclear fuel bundle spacers  

SciTech Connect (OSTI)

This patent describes a method of manufacturing nuclear fuel bundle spacers on an automated production line basis. It comprises: cutting elongated tubing stock into shorter tubular ferrules; checking the length of each ferrule and rejecting those ferrules of unacceptable lengths; cutting predetermined features in the sidewall of each ferrule; forming the sidewall of each ferrule to impart predetermined surface formations thereto; checking a critical dimension of each sidewall surface formation of each ferrule and rejecting those of unacceptable dimensions; assembling successive pairs of ferrules into subassemblies; assembling successive subassemblies into a spacer assembly fixture; assembling a peripheral band in the spacer assembly fixture; conjoining the ferrules to each other and to the peripheral band to create a structurally rigid, finished spacer; and providing a separate controller for automatically controlling and monitoring the performances of these steps.

White, D.W.; Muncy, D.G.; Schoenig, F.C. Jr.

1989-09-26T23:59:59.000Z

99

Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay  

E-Print Network [OSTI]

of the Institute of Nuclear Material Management, Tucson, AZ,Assay, Institute of Nuclear Materials Management 51st Annual

Quiter, Brian

2012-01-01T23:59:59.000Z

100

Nuclear fuel cycle facility accident analysis handbook  

SciTech Connect (OSTI)

The purpose of this Handbook is to provide guidance on how to calculate the characteristics of releases of radioactive materials and/or hazardous chemicals from nonreactor nuclear facilities. In addition, the Handbook provides guidance on how to calculate the consequences of those releases. There are four major chapters: Hazard Evaluation and Scenario Development; Source Term Determination; Transport Within Containment/Confinement; and Atmospheric Dispersion and Consequences Modeling. These chapters are supported by Appendices, including: a summary of chemical and nuclear information that contains descriptions of various fuel cycle facilities; details on how to calculate the characteristics of source terms for releases of hazardous chemicals; a comparison of NRC, EPA, and OSHA programs that address chemical safety; a summary of the performance of HEPA and other filters; and a discussion of uncertainties. Several sample problems are presented: a free-fall spill of powder, an explosion with radioactive release; a fire with radioactive release; filter failure; hydrogen fluoride release from a tankcar; a uranium hexafluoride cylinder rupture; a liquid spill in a vitrification plant; and a criticality incident. Finally, this Handbook includes a computer model, LPF No.1B, that is intended for use in calculating Leak Path Factors. A list of contributors to the Handbook is presented in Chapter 6. 39 figs., 35 tabs.

NONE

1998-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear fuels limited" 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

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

102

Spent Nuclear Fuel project integrated safety management plan  

SciTech Connect (OSTI)

This document is being revised in its entirety and the document title is being revised to ``Spent Nuclear Fuel Project Integrated Safety Management Plan.

Daschke, K.D.

1996-09-17T23:59:59.000Z

103

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

104

Thermomechanical analysis of innovative nuclear fuel pin designs  

E-Print Network [OSTI]

One way to increase the power of a nuclear reactor is to change the solid cylindrical fuel to Internally and Externally Cooled (I&EC) annular fuel, and adjust the flow and the core inlet coolant temperature. The switch to ...

Lerch Andrew (Andrew J.)

2010-01-01T23:59:59.000Z

105

Privacy Act Fees and Time Limits | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartmentNational Nuclearhas 'Natitude' | NationalADRWindNuclearSecurityNuclearExemptions

106

International nuclear fuel cycle fact book. Revision 6  

SciTech Connect (OSTI)

The International Fuel Cycle 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. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2.

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

1986-01-01T23:59:59.000Z

107

Fabrication of high exposure nuclear fuel pellets  

DOE Patents [OSTI]

A method is disclosed for making a fuel pellet for a nuclear reactor. A mixture is prepared of PuO.sub.2 and UO.sub.2 powders, where the mixture contains at least about 30% PuO.sub.2, and where at least about 12% of the Pu is the Pu.sup.240 isotope. To this mixture is added about 0.3 to about 5% of a binder having a melting point of at least about 250.degree. F. The mixture is pressed to form a slug and the slug is granulated. Up to about 4.7% of a lubricant having a melting point of at least about 330.degree. F. is added to the granulated slug. Both the binder and the lubricant are selected from a group consisting of polyvinyl carboxylate, polyvinyl alcohol, naturally occurring high molecular weight cellulosic polymers, chemically modified high molecular weight cellulosic polymers, and mixtures thereof. The mixture is pressed to form a pellet and the pellet is sintered.

Frederickson, James R. (Richland, WA)

1987-01-01T23:59:59.000Z

108

World nuclear capacity and fuel cycle requirements, November 1993  

SciTech Connect (OSTI)

This analysis report presents the current status and projections of nuclear capacity, generation, and fuel cycle requirements for all countries in the world using nuclear power to generate electricity for commercial use. Long-term projections of US nuclear capacity, generation, fuel cycle requirements, and spent fuel discharges for three different scenarios through 2030 are provided in support of the Department of Energy`s activities pertaining to the Nuclear Waste Policy Act of 1982 (as amended in 1987). The projections of uranium requirements also support the Energy Information Administration`s annual report, Domestic Uranium Mining and Milling Industry: Viability Assessment.

Not Available

1993-11-30T23:59:59.000Z

109

Economics of nuclear fuel cycles : option valuation and neutronics simulation of mixed oxide fuels  

E-Print Network [OSTI]

In most studies aiming at the economic assessment of nuclear fuel cycles, a primary concern is to keep scenarios economically comparable. For Uranium Oxide (UOX) and Mixed Oxide (MOX) fuels, a traditional way to achieve ...

De Roo, Guillaume

2009-01-01T23:59:59.000Z

110

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

111

Technology Readiness Levels for Advanced Nuclear Fuels and Materials Development  

SciTech Connect (OSTI)

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

Jon Carmack

2014-01-01T23:59:59.000Z

112

PRESENT LIMITATIONS OF CdTe DETECTORS IN NUCLEAR MEDICINE  

E-Print Network [OSTI]

365 PRESENT LIMITATIONS OF CdTe DETECTORS IN NUCLEAR MEDICINE R. ALLEMAND, P. BOUTEILLER, M. LAVAL quality criteria, it is necessary to compare Cd-Te detectors results (or estimated characteristics) with other methods (i. e. 8cintillation cameras) in order to know the effective interest of Cd-Te in nuclear

Boyer, Edmond

113

Spent nuclear fuel discharges from U.S. reactors 1994  

SciTech Connect (OSTI)

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

NONE

1996-02-01T23:59:59.000Z

114

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites  

SciTech Connect (OSTI)

This report fulfills the M2 milestone M2FT-13PN0912022, “Stranded Sites De-Inventorying Report.” In January 2013, the U.S. Department of Energy (DOE) issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (DOE 2013). 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 (BRC 2012). Shutdown sites are defined as those commercial nuclear power reactor sites where the nuclear power reactors have been shut down and the site has been decommissioned or is undergoing decommissioning. In this report, a preliminary evaluation of removing used nuclear fuel from 12 shutdown sites was conducted. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, and San Onofre. These sites have no other operating nuclear power reactors at their sites and have also notified the U.S. Nuclear Regulatory Commission that their reactors have permanently ceased power operations and that nuclear fuel has been permanently removed from their reactor vessels. Shutdown reactors at sites having other operating reactors are not included in this evaluation.

Maheras, Steven J. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Best, Ralph E.; Ross, Steven B. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Buxton, Kenneth A. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); England, Jeffery L. [Savannah River National Laboratory, Aiken, SC (United States); McConnell, Paul E. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

2013-09-30T23:59:59.000Z

115

Benefits and concerns of a closed nuclear fuel cycle  

SciTech Connect (OSTI)

Nuclear power can play an important role in our energy future, contributing to increasing electricity demand while at the same time decreasing carbon dioxide emissions. However, the nuclear fuel cycle in the United States today is unsustainable. As stated in the 1982 Nuclear Waste Policy Act, the U.S. Department of Energy is responsible for disposing of spent nuclear fuel generated by commercial nuclear power plants operating in a “once-through” fuel cycle in the deep geologic repository located at Yucca Mountain. However, unyielding political opposition to the site has hindered the commissioning process to the extant that the current administration has recently declared the unsuitability of the Yucca Mountain site. In light of this the DOE is exploring other options, including closing the fuel cycle through recycling and reprocessing of spent nuclear fuel. The possibility of closing the fuel cycle is receiving special attention because of its ability to minimize the final high level waste (HLW) package as well as recover additional energy value from the original fuel. The technology is, however, still very controversial because of the increased cost and proliferation risk it can present. To lend perspective on the closed fuel cycle alternative, this presents the arguments for and against closing the fuel cycle with respect to sustainability, proliferation risk, commercial viability, waste management, and energy security.

Widder, Sarah H.

2010-11-17T23:59:59.000Z

116

Coupon Surveillance For Corrosion Monitoring In Nuclear Fuel Basin  

SciTech Connect (OSTI)

Aluminum and stainless steel coupons were put into a nuclear fuel basin to monitor the effect of water chemistry on the corrosion of fuel cladding. These coupons have been monitored for over ten years. The corrosion and pitting data is being used to model the kinetics and estimate the damage that is occurring to the fuel cladding.

Mickalonis, J. I.; Murphy, T. R.; Deible, R.

2012-10-01T23:59:59.000Z

117

Limited Test Ban Treaty | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartmentNational NuclearhasAdministration77NuclearSecurityAdministrator |Life Extension|

118

Spent nuclear fuel discharges from US reactors 1993  

SciTech Connect (OSTI)

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

Not Available

1995-02-01T23:59:59.000Z

119

Methods for making a porous nuclear fuel element  

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

2014-12-30T23:59:59.000Z

120

Thermoacoustic sensor for nuclear fuel temperaturemonitoring and heat transfer enhancement  

SciTech Connect (OSTI)

A new acoustical sensing system for the nuclear power industry has been developed at The Pennsylvania State University in collaboration with Idaho National Laboratories. This sensor uses the high temperatures of nuclear fuel to convert a nuclear fuel rod into a standing-wave thermoacoustic engine. When a standing wave is generated, the sound wave within the fuel rod will be propagated, by acoustic radiation, through the cooling fluid within the reactor or spent fuel pool and can be monitored a remote location external to the reactor. The frequency of the sound can be correlated to an effective temperature of either the fuel or the surrounding coolant. We will present results for a thermoacoustic resonator built into a Nitonic-60 (stainless steel) fuel rod that requires only one passive component and no heat exchangers.

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

2013-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear fuels limited" 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

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

122

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

123

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

124

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

125

Method of increasing the deterrent to proliferation of nuclear fuels  

DOE Patents [OSTI]

A process of recycling protactinium-231 to enhance the utilization of radioactively hot uranium-232 in nuclear fuel for the purpose of making both fresh and spent fuel more resistant to proliferation. The uranium-232 may be obtained by the irradiation of protactinium-231 which is normally found in the spent fuel rods of a thorium base nuclear reactor. The production of protactinium-231 and uranium-232 would be made possible by the use of the thorium uranium-233 fuel cycle in power reactors.

Rampolla, Donald S. (Pittsburgh, PA)

1982-01-01T23:59:59.000Z

126

Sandia National Laboratories: Nuclear Energy and Fuel Cycle Programs  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 -theErik Spoerke SSLSMolten-Salt StorageNo More GreenWorkshops Nuclear

127

TEPP - Spent Nuclear Fuel | Department of Energy  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNGInternational EnergyCommitteeRenewable Energy,Section 180(c)CHARTER

128

Used Nuclear Fuels Storage, Transportation, and Disposal Analysis...  

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

Used Nuclear Fuels Storage, Transportation, and Disposal Analysis Resource and Data System (UNF-ST&DARDS) Apr 08 2014 10:00 AM - 11:00 AM John M. Scaglione, ORNL staff, Oak Ridge...

129

Inventory of LWR spent nuclear fuel in the 324 Building  

SciTech Connect (OSTI)

This document contains the results of calculations to estimate the decay heat, neutron source term, photon source term, and radioactive inventory of light-water-reactor spent nuclear fuel in the 324 Building at Pacific Northwest National Laboratory.

Jenquin, U.P.

1996-09-01T23:59:59.000Z

130

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

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

Used nuclear fuel (UNF) must maintain its integrity during the storage period in such a way that it can withstand the physical forces of handling and transportation associated with...

131

Department of Mechanical and Nuclear Engineering Spring 2012 Fuel Efficient Stoves to Achieve Fuel Security  

E-Print Network [OSTI]

PENNSTATE Department of Mechanical and Nuclear Engineering Spring 2012 Fuel Efficient Stoves to Achieve Fuel Security Overview Tanzanians living near the Udzungwa Mountains National Park have 100,000 villagers without an available fuel source. One possible solution to alleviate this crisis

Demirel, Melik C.

132

What to Expect When Readying to Move Spent Nuclear Fuel from...  

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

What to Expect When Readying to Move Spent Nuclear Fuel from Commercial Nuclear Power Plants What to Expect When Readying to Move Spent Nuclear Fuel from Commercial Nuclear Power...

133

Pyroprocessing of Fast Flux Test Facility Nuclear Fuel  

SciTech Connect (OSTI)

Used nuclear fuel from the Fast Flux Test Facility (FFTF) was recently transferred to the Idaho National Laboratory and processed by pyroprocessing in the Fuel Conditioning Facility. Approximately 213 kg of uranium from sodium-bonded metallic FFTF fuel was processed over a one year period with the equipment previously used for the processing of EBR-II used fuel. The peak burnup of the FFTF fuel ranged from 10 to 15 atom% for the 900+ chopped elements processed. Fifteen low-enriched uranium ingots were cast following the electrorefining and distillation operations to recover approximately 192 kg of uranium. A material balance on the primary fuel constituents, uranium and zirconium, during the FFTF campaign will be presented along with a brief description of operating parameters. Recoverable uranium during the pyroprocessing of FFTF nuclear fuel was greater than 95% while the purity of the final electrorefined uranium products exceeded 99%.

B.R. Westphal; G.L. Fredrickson; G.G. Galbreth; D. Vaden; M.D. Elliott; J.C. Price; E.M. Honeyfield; M.N. Patterson; L. A. Wurth

2013-10-01T23:59:59.000Z

134

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

135

Pyroprocessing of fast flux test facility nuclear fuel  

SciTech Connect (OSTI)

Used nuclear fuel from the Fast Flux Test Facility (FFTF) was recently transferred to the Idaho National Laboratory and processed by pyroprocessing in the Fuel Conditioning Facility. Approximately 213 kg of uranium from sodium-bonded metallic FFTF fuel was processed over a one year period with the equipment previously used for the processing of EBR-II used fuel. The peak burnup of the FFTF fuel ranged from 10 to 15 atom% for the 900+ chopped elements processed. Fifteen low-enriched uranium ingots were cast following the electrorefining and distillation operations to recover approximately 192 kg of uranium. A material balance on the primary fuel constituents, uranium and zirconium, during the FFTF campaign will be presented along with a brief description of operating parameters. Recoverable uranium during the pyroprocessing of FFTF nuclear fuel was greater than 95% while the purity of the final electro-refined uranium products exceeded 99%. (authors)

Westphal, B.R.; Wurth, L.A.; Fredrickson, G.L.; Galbreth, G.G.; Vaden, D.; Elliott, M.D.; Price, J.C.; Honeyfield, E.M.; Patterson, M.N. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID, 83415 (United States)

2013-07-01T23:59:59.000Z

136

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.

James A. Smith; David L. Cottle; Barry H. Rabin

2013-07-01T23:59:59.000Z

137

A Perspective on the U.S. Nuclear Fuel Cycle  

SciTech Connect (OSTI)

There has been a resurgence of interest in the possibility of processing the US spent nuclear fuel, instead of burying it in a geologic repository. Accordingly, key topical findings from three relevant EPRI evaluations made in the 1990-1995 time-frame are recapped and updated to accommodate a few developments over the subsequent ten years. Views recently expressed by other US entities are discussed. Processing aspects thereby addressed include effects on waste disposal and on geologic repository capacity, impacts on the economics of the nuclear fuel cycle and of the overall nuclear power scenario, alternative dispositions of the plutonium separated by the processing, impacts on the structure of the perceived weapons proliferation risk, and challenges for the immediate future and for the current half-century. Currently, there is a statutory limit of 70,000 metric tons on the amount of nuclear waste materials that can be accepted at Yucca Mountain. The Environmental Impact Statement (EIS) for the project analyzed emplacement of up to 120,000 metric tons of nuclear waste products in the repository. Additional scientific analyses suggest significantly higher capacity could be achieved with changes in the repository configuration that use only geology that has already been characterized and do not deviate from existing design parameters. Conservatively assuming the repository capacity postulated in the EIS, the need date for a second repository is essentially deferrable until that determined by a potential new nuclear plant deployment program. A further increase in technical capacity of the first repository (and further and extensive delay to the need date for a second repository) is potentially achievable by processing the spent fuel to remove the plutonium (and at least the americium too), provided the plutonium and the americium are then comprehensively burnt. The burning of some of the isotopes involved would need fast reactors (discounting for now a small possibility that one of several recently postulated alternatives will prove superior overall). However, adoption of processing would carry a substantial cost burden and reliability of the few demonstration fast reactors built to-date has been poor. Trends and developments could remove these obstacles to the processing scenario, possibly before major decisions on a second repository become necessary, which need not be until mid-century at the earliest. Pending the outcomes of these long-term trends and developments, economics and reliability encourage us to stay with non-processing for the near term at least. Besides completing the Yucca Mountain program, the two biggest and inter-related fuel-cycle needs today are for a nationwide consensus on which processing technology offers the optimum mix of economic competitiveness and proliferation resistance and for a sustained effort to negotiate greater international cooperation and safeguards. Equally likely to control the readiness schedule is development/demonstration of an acceptable, reliable and affordable fast reactor. (authors)

Rodwell, Ed; Machiels, Albert [Electric Power Research Institute, Inc. - EPRI, 3420 Hillview Avenue, Palo Alto, California 94304 (United States)

2006-07-01T23:59:59.000Z

138

MANAGING SPENT NUCLEAR FUEL WASTES AT THE IDAHO NATIONAL LABORATORY  

SciTech Connect (OSTI)

The Idaho National Engineering Laboratory (INL) has a large inventory of diverse types of spent nuclear fuel (SNF). This legacy is in part due to the history of the INL as the National Reactor Testing Station, in part to its mission to recover highly enriched uranium from SNF and in part to it’s mission to test and examine SNF after irradiation. The INL also has a large diversity of SNF storage facility, some dating back 50 years in the site history. The success of the INL SNF program is measured by its ability to: 1) achieve safe existing storage, 2) continue to receive SNF from other locations, both foreign and domestic, 3) repackage SNF from wet storage to interim dry storage, and 4) prepare the SNF for dispositioning in a federal repository. Because of the diversity in the SNF and the facilities at the INL, the INL is addressing almost very condition that may exist in the SNF world. Many of solutions developed by the INL are applicable to other SNF storage sites as they develop their management strategy. The SNF being managed by the INL are in a variety of conditions, from intact assemblies to individual rods or plates to powders, rubble, and metallurgical mounts. Some of the fuel has been in wet storage for over forty years. The fuel is stored bare, or in metal cans and either wet under water or dry in vaults, caissons or casks. Inspections have shown varying degrees of corrosion and degradation of the fuel and the storage cans. Some of the fuel has been recanned under water, and the conditions of the fuel inside the second or third can are unknown. The fuel has been stored in one of 10 different facilities: five wet pools and one casks storage pad, one vault, two generations of caisson facilities, and one modular Independent Spent Fuel Storage Installation (ISFSI). The wet pools range from forty years old to the most modern pool in the US Department of Energy (DOE) complex. The near-term objective is moving the fuel in the older wet storage facilities to interim dry storage facilities, thus permitting the shutdown and decommission of the older facilities. Two wet pool facilities, one at the Idaho Nuclear Technology and Engineering Center and the other at Test Area North, were targeted for initial SNF movements since these were some of the oldest at the INL. Because of the difference in the SNF materials different types of drying processes had to be developed. Passive drying, as is done with typical commercial SNF was not an option because on the condition of some of the fuel, the materials to be dried, and the low heat generation of some of the SNF. There were also size limitations in the existing facility. Active dry stations were designed to address the specific needs of the SNF and the facilities.

Hill, Thomas J

2005-09-01T23:59:59.000Z

139

A review of nuclear fuel cycle options for developing nations  

SciTech Connect (OSTI)

A study of several nuclear reactor and fuel cycle options for developing nations was performed. All reactor choices were considered under a GNEP framework. Two advanced alternative reactor types, a nuclear battery-type reactor and a fuel reprocessing fast reactor were examined and compared with a conventional Generation III+ LWR reactor. The burn of nuclear fuel was simulated using ORIGEN 2.2 for each reactor type and the resulting information was used to compare the options in terms of waste produced, waste quality and repository impact. The ORIGEN data was also used to evaluate the economics of the fuel cycles using unit costs, discount rates and present value functions with the material balances. The comparison of the fuel cycles and reactors developed in this work provides a basis for the evaluation of subsidy programs and cost-benefit comparisons for various reactor parameters such as repository impact and proliferation risk versus economic considerations. (authors)

Harrison, R.K.; Scopatz, A.M.; Ernesti, M. [The University of Texas at Austin, Pickle Research Campus, Building 159, Austin, TX 78712 (United States)

2007-07-01T23:59:59.000Z

140

Managing Spent Nuclear Fuel at the Idaho National Laboratory  

SciTech Connect (OSTI)

The Idaho National Laboratory (INL) has a large inventory of diverse types of spent nuclear fuel (SNF). This legacy derives from the history of the INL as the National Reactor Testing Station, and from its mission to recover HEU from SNF and to test and examine SNF after irradiation. The INL also has a large diversity of SNF storage facilities, some 50 years old. SNF at INL has many forms—from intact assemblies down to metallurgical mounts, and some fuel has been wet stored for over 40 years. SNF is stored bare or in metal cans under water, or dry in vaults, caissons or casks. Inspection shows varying corrosion and degradation of the SNF and its storage cans. SNF has been stored in 10 different facilities: 5 pools, one cask storage pad, one vault, two generations of caisson facilities, and one modular Independent Spent Fuel Storage Installation (ISFSI). The pools range in age from 40 years old to the most modern in the US Department of Energy (DOE) complex. The near-term objective is to move SNF from older pools to interim dry storage, allowing shutdown and decommissioning of the older facilities. This move involves drying methods that are dependent on fuel type. The long-term objective is to have INL SNF in safe dry storage and ready to be shipped to the National Repository. The unique features of the INL SNF requires special treatments and packaging to meet the proposed repository acceptance criteria and SNF will be repackaged in standardized canisters for shipment and disposal in the National Repository. Disposal will use the standardized canisters that can be co-disposed with High Level Waste glass logs to limit the total fissile material in a repository waste package. The DOE standardized canister also simplifies the repository handling of the multitude of DOE SNF sizes and shapes.

Thomas Hill; Denzel L. Fillmore

2005-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear fuels limited" 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

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

142

Reprocessing of nuclear fuels at the Savannah River Plant  

SciTech Connect (OSTI)

For more than 30 years, the Savannah River Plant (SRP) has been a major supplier of nuclear materials such as plutonium-239 and tritium-3 for nuclear and thermonuclear weapons, plutonium-238 for space exploration, and isotopes of americium, curium, and californium for use in the nuclear research community. SRP is a complete nuclear park, providing most of the processes in the nuclear fuel cycle. Key processes involve fabrication and cladding of the nuclear fuel, target, and control assemblies; rework of heavy water for use as reactor moderator; reactor loading, operation, and unloading; chemical recovery of the reactor transmutation products and spent fuels; and management of the gaseous, liquid, and solid nuclear and chemical wastes; plus a host of support operations. The site's history and the key processes from fabrication of reactor fuels and targets to finishing of virgin plutonium for use in the nuclear weapons complex are reviewed. Emphasis has been given to the chemistry of the recovery and purification of weapons grade plutonium from irradiated reactor targets.

Gray, L.W.

1986-10-04T23:59:59.000Z

143

Global nuclear energy partnership fuels transient testing at the Sandia National Laboratories nuclear facilities : planning and facility infrastructure options.  

SciTech Connect (OSTI)

The Global Nuclear Energy Partnership fuels development program is currently developing metallic, oxide, and nitride fuel forms as candidate fuels for an Advanced Burner Reactor. The Advance Burner Reactor is being designed to fission actinides efficiently, thereby reducing the long-term storage requirements for spent fuel repositories. Small fuel samples are being fabricated and evaluated with different transuranic loadings and with extensive burnup using the Advanced Test Reactor. During the next several years, numerous fuel samples will be fabricated, evaluated, and tested, with the eventual goal of developing a transmuter fuel database that supports the down selection to the most suitable fuel type. To provide a comparative database of safety margins for the range of potential transmuter fuels, this report describes a plan to conduct a set of early transient tests in the Annular Core Research Reactor at Sandia National Laboratories. The Annular Core Research Reactor is uniquely qualified to perform these types of tests because of its wide range of operating capabilities and large dry central cavity which extents through the center of the core. The goal of the fuels testing program is to demonstrate that the design and fabrication processes are of sufficient quality that the fuel will not fail at its design limit--up to a specified burnup, power density, and operating temperature. Transient testing is required to determine the fuel pin failure thresholds and to demonstrate that adequate fuel failure margins exist during the postulated design basis accidents.

Kelly, John E.; Wright, Steven Alan; Tikare, Veena; MacLean, Heather J. (Idaho National Laboratory, Idaho Falls, ID); Parma, Edward J., Jr.; Peters, Curtis D.; Vernon, Milton E.; Pickard, Paul S.

2007-10-01T23:59:59.000Z

144

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

145

Energy Return on Investment from Recycling Nuclear Fuel  

SciTech Connect (OSTI)

This report presents an evaluation of the Energy Return on Investment (EROI) from recycling an initial batch of 800 t/y of used nuclear fuel (UNF) through a Recycle Center under a number of different fuel cycle scenarios. The study assumed that apart from the original 800 t of UNF only depleted uranium was available as a feed. Therefore for each subsequent scenario only fuel that was derived from the previous fuel cycle scenario was considered. The scenarios represent a good cross section of the options available and the results contained in this paper and associated appendices will allow for other fuel cycle options to be considered.

None

2011-08-17T23:59:59.000Z

146

Nuclear Fuels Storage & Transportation Planning Project Documents |  

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

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147

Nuclear fuel recycling in 4 minutes | Argonne National Laboratory  

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

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148

Sandia National Laboratories: Nuclear Energy and Fuel Systems Programs  

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

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149

Breeding nuclear fuels with accelerators: replacement for breeder reactors  

SciTech Connect (OSTI)

One application of high energy particle accelerators has been, and still is, the production of nuclear fuel for the nuclear energy industry; tantalizing because it would create a whole new industry. This approach to producing fissile from fertile material was first considered in the early 1950's in the context of the nuclear weapons program. A considerable development effort was expended before discovery of uranium ore in New Mexico put an end to the project. Later, US commitment to the Liquid Metal Fast Breeder Reactors (LMFBR) killed any further interest in pursuing accelerator breeder technology. Interest in the application of accelerators to breed nuclear fuels, and possibly burn nuclear wastes, revived in the late 1970's, when the LMFBR came under attack during the Carter administration. This period gave the opportunity to revisit the concept in view of the present state of the technology. This evaluation and the extensive calculational modeling of target designs that have been carried out are promising. In fact, a nuclear fuel cycle of Light Water Reactors and Accelerator Breeders is competitive to that of the LMFBR. At this time, however, the relative abundance of uranium reserves vs electricity demand and projected growth rate render this study purely academic. It will be for the next generation of accelerator builders to demonstate the competitiveness of this technology versus that of other nuclear fuel cycles, such as LMFBR's or Fusion Hybrid systems. 22 references, 1 figure, 5 tables.

Grand, P.; Takahashi, H.

1984-01-01T23:59:59.000Z

150

Depleted uranium as a backfill for nuclear fuel waste package  

DOE Patents [OSTI]

A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

Forsberg, Charles W. (Oak Ridge, TN)

1998-01-01T23:59:59.000Z

151

Depleted uranium as a backfill for nuclear fuel waste package  

DOE Patents [OSTI]

A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

Forsberg, C.W.

1998-11-03T23:59:59.000Z

152

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

153

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

154

Advanced nuclear fuel cycles - Main challenges and strategic choices  

SciTech Connect (OSTI)

A graphical conceptual model of the uranium fuel cycles has been developed to capture the present, anticipated, and potential (future) nuclear fuel cycle elements. The once-through cycle and plutonium recycle in fast reactors represent two basic approaches that bound classical options for nuclear fuel cycles. Chief among these other options are mono-recycling of plutonium in thermal reactors and recycling of minor actinides in fast reactors. Mono-recycling of plutonium in thermal reactors offers modest savings in natural uranium, provides an alternative approach for present-day interim management of used fuel, and offers a potential bridging technology to development and deployment of future fuel cycles. In addition to breeder reactors' obvious fuel sustainability advantages, recycling of minor actinides in fast reactors offers an attractive concept for long-term management of the wastes, but its ultimate value is uncertain in view of the added complexity in doing so,. Ultimately, there are no simple choices for nuclear fuel cycle options, as the selection of a fuel cycle option must reflect strategic criteria and priorities that vary with national policy and market perspectives. For example, fuel cycle decision-making driven primarily by national strategic interests will likely favor energy security or proliferation resistance issues, whereas decisions driven primarily by commercial or market influences will focus on economic competitiveness.

Le Biez, V. [Corps des Mines, 35 bis rue Saint-Sabin, F-75011 Paris (France); Machiels, A.; Sowder, A. [Electric Power Research Institute, Inc. 3420, Hillview Avenue, Palo Alto, CA 94304 (United States)

2013-07-01T23:59:59.000Z

155

Nuclear Waste Imaging and Spent Fuel Verification by Muon Tomography  

E-Print Network [OSTI]

This paper explores the use of cosmic ray muons to image the contents of shielded containers and detect high-Z special nuclear materials inside them. Cosmic ray muons are a naturally occurring form of radiation, are highly penetrating and exhibit large scattering angles on high Z materials. Specifically, we investigated how radiographic and tomographic techniques can be effective for non-invasive nuclear waste characterization and for nuclear material accountancy of spent fuel inside dry storage containers. We show that the tracking of individual muons, as they enter and exit a structure, can potentially improve the accuracy and availability of data on nuclear waste and the contents of Dry Storage Containers (DSC) used for spent fuel storage at CANDU plants. This could be achieved in near real time, with the potential for unattended and remotely monitored operations. We show that the expected sensitivity, in the case of the DSC, exceeds the IAEA detection target for nuclear material accountancy.

Jonkmans, G; Jewett, C; Thompson, M

2012-01-01T23:59:59.000Z

156

Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs, Draft Environmental Impact Statement. Volume 1, Appendix D: Part A, Naval Spent Nuclear Fuel Management  

SciTech Connect (OSTI)

Volume 1 to the Department of Energy`s Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Management Programs Environmental Impact Statement evaluates a range of alternatives for managing naval spent nuclear fuel expected to be removed from US Navy nuclear-powered vessels and prototype reactors through the year 2035. The Environmental Impact Statement (EIS) considers a range of alternatives for examining and storing naval spent nuclear fuel, including alternatives that terminate examination and involve storage close to the refueling or defueling site. The EIS covers the potential environmental impacts of each alternative, as well as cost impacts and impacts to the Naval Nuclear Propulsion Program mission. This Appendix covers aspects of the alternatives that involve managing naval spent nuclear fuel at four naval shipyards and the Naval Nuclear Propulsion Program Kesselring Site in West Milton, New York. This Appendix also covers the impacts of alternatives that involve examining naval spent nuclear fuel at the Expended Core Facility in Idaho and the potential impacts of constructing and operating an inspection facility at any of the Department of Energy (DOE) facilities considered in the EIS. This Appendix also considers the impacts of the alternative involving limited spent nuclear fuel examinations at Puget Sound Naval Shipyard. This Appendix does not address the impacts associated with storing naval spent nuclear fuel after it has been inspected and transferred to DOE facilities. These impacts are addressed in separate appendices for each DOE site.

Not Available

1994-06-01T23:59:59.000Z

157

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

158

Impact of alternative nuclear fuel cycle options on infrastructure and fuel requirements, actinide and waste inventories, and economics  

E-Print Network [OSTI]

The nuclear fuel once-through cycle (OTC) scheme currently practiced in the U.S. leads to accumulation of uranium, transuranic (TRU) and fission product inventories in the spent nuclear fuel. Various separation and recycling ...

Guérin, Laurent, S.M. Massachusetts Institute of Technology

2009-01-01T23:59:59.000Z

159

Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors  

DOE Patents [OSTI]

Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). 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, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

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

2010-02-23T23:59:59.000Z

160

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 "nuclear fuels limited" 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

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

162

Next-generation nuclear fuel withstands high-temperature accident  

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

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163

Sandia National Laboratories: Recent Sandia International Used Nuclear Fuel  

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

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164

Sandia National Laboratories: Nuclear Fuel Cycle Options Catalog  

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

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165

Nuclear Fuels Storage and Transportation Planning Project (NFST) Program  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNG |September 15, 2010 PrintingNeed| Department ofDC. |NuclearFacts:Department

166

International Nuclear Fuel Cycle Fact Book. Revision 12  

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

167

International nuclear fuel cycle fact book: Revision 9  

SciTech Connect (OSTI)

The International Nuclear Fuel Cycle Fact Book has been compiled in an effort to provide current data concerning fuel cycle and waste management facilities, R and D programs and key personnel. The Fact Book contains: national summaries in which a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; and international agencies in which a section for each of the international agencies which has significant fuel cycle involvement, and a listing of nuclear societies. The national summaries, in addition to the data described above, feature a small map for each country as well as some general information. The latter is presented from the perspective of the Fact Book user in the United States.

Leigh, I.W.

1989-01-01T23:59:59.000Z

168

Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 1, Appendix C, Savannah River Site Spent Nuclear Fuel Mangement Program  

SciTech Connect (OSTI)

The US Department of Energy (DOE) is engaged in two related decision making processes concerning: (1) the transportation, receipt, processing, and storage of spent nuclear fuel (SNF) at the DOE Idaho National Engineering Laboratory (INEL) which will focus on the next 10 years; and (2) programmatic decisions on future spent nuclear fuel management which will emphasize the next 40 years. DOE is analyzing the environmental consequences of these spent nuclear fuel management actions in this two-volume Environmental Impact Statement (EIS). Volume 1 supports broad programmatic decisions that will have applicability across the DOE complex and describes in detail the purpose and need for this DOE action. Volume 2 is specific to actions at the INEL. This document, which limits its discussion to the Savannah River Site (SRS) spent nuclear fuel management program, supports Volume 1 of the EIS. Following the introduction, Chapter 2 contains background information related to the SRS and the framework of environmental regulations pertinent to spent nuclear fuel management. Chapter 3 identifies spent nuclear fuel management alternatives that DOE could implement at the SRS, and summarizes their potential environmental consequences. Chapter 4 describes the existing environmental resources of the SRS that spent nuclear fuel activities could affect. Chapter 5 analyzes in detail the environmental consequences of each spent nuclear fuel management alternative and describes cumulative impacts. The chapter also contains information on unavoidable adverse impacts, commitment of resources, short-term use of the environment and mitigation measures.

Not Available

1994-06-01T23:59:59.000Z

169

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

170

Nuclear Fuel Cycle Reasoner: PNNL FY13 Report  

SciTech Connect (OSTI)

In Fiscal Year 2012 (FY12) PNNL implemented a formal reasoning framework and applied it to a specific challenge in nuclear nonproliferation. The Semantic Nonproliferation Analysis Platform (SNAP) was developed as a preliminary graphical user interface to demonstrate the potential power of the underlying semantic technologies to analyze and explore facts and relationships relating to the nuclear fuel cycle (NFC). In Fiscal Year 2013 (FY13) the SNAP demonstration was enhanced with respect to query and navigation usability issues.

Hohimer, Ryan E.; Strasburg, Jana D.

2013-09-30T23:59:59.000Z

171

Long-term tradeoffs between nuclear- and fossil-fuel burning  

SciTech Connect (OSTI)

A global energy/economics/environmental (E{sup 3}) model has been adapted with a nuclear energy/materials model to understand better {open_quotes}top-level{close_quotes}, long-term trade offs between civilian nuclear power, nuclear-weapons proliferation, fossil-fuel burning, and global economic welfare. Using a {open_quotes}business-as-usual{close_quotes} (BAU) point-of-departure case, economic, resource, proliferation-risk implications of plutonium recycle in LAIRs, greenhouse-gas-mitigating carbon taxes, and a range of nuclear energy costs (capital and fuel) considerations have been examined. After describing the essential elements of the analysis approach being developed to support the Los Alamos Nuclear Vision Project, preliminary examples of parametric variations about the BAU base-case scenario are presented. The results described herein represent a sampling from more extensive results collected in a separate report. The primary motivation here is: (a) to compare the BAU basecase with results from other studies; (b) to model on a regionally resolved global basis long-term (to year {approximately}2100) evolution of plutonium accumulation in a variety of forms under a limited range of fuel-cycle scenarios; and (c) to illustrate a preliminary connectivity between risks associated with nuclear proliferation and fossil-fuel burning (e.g., greenhouse-gas accumulations).

Krakowski, R.A.

1996-12-31T23:59:59.000Z

172

Apparatus for injection casting metallic nuclear energy fuel rods  

DOE Patents [OSTI]

Molds for making metallic nuclear fuel rods are provided which present reduced risks to the environment by reducing radioactive waste. In one embodiment, the mold is consumable with the fuel rod, and in another embodiment, part of the mold can be re-used. Several molds can be arranged together in a cascaded manner, if desired, or several long cavities can be integrated in a monolithic multiple cavity re-usable mold.

Seidel, Bobby R. (Idaho Falls, ID); Tracy, Donald B. (Firth, ID); Griffiths, Vernon (Butte, MT)

1991-01-01T23:59:59.000Z

173

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

174

Nuclear Fuel Cycle Reasoner: PNNL FY12 Report  

SciTech Connect (OSTI)

Building on previous internal investments and leveraging ongoing advancements in semantic technologies, PNNL implemented a formal reasoning framework and applied it to a specific challenge in nuclear nonproliferation. The Semantic Nonproliferation Analysis Platform (SNAP) was developed as a preliminary graphical user interface to demonstrate the potential power of the underlying semantic technologies to analyze and explore facts and relationships relating to the nuclear fuel cycle (NFC). In developing this proof of concept prototype, the utility and relevancy of semantic technologies to the Office of Defense Nuclear Nonproliferation Research and Development (DNN R&D) has been better understood.

Hohimer, Ryan E.; Pomiak, Yekaterina G.; Neorr, Peter A.; Gastelum, Zoe N.; Strasburg, Jana D.

2013-05-03T23:59:59.000Z

175

Air Shipment of Spent Nuclear Fuel from Romania to Russia  

SciTech Connect (OSTI)

Romania successfully completed the world’s first air shipment of spent nuclear fuel transported in Type B(U) casks under existing international laws and without shipment license special exceptions when the last Romanian highly enriched uranium (HEU) spent nuclear fuel was transported to the Russian Federation in June 2009. This air shipment required the design, fabrication, and licensing of special 20 foot freight containers and cask tiedown supports to transport the eighteen TUK 19 shipping casks on a Russian commercial cargo aircraft. The new equipment was certified for transport by road, rail, water, and air to provide multi modal transport capabilities for shipping research reactor spent fuel. The equipment design, safety analyses, and fabrication were performed in the Russian Federation and transport licenses were issued by both the Russian and Romanian regulatory authorities. The spent fuel was transported by truck from the VVR S research reactor to the Bucharest airport, flown by commercial cargo aircraft to the airport at Yekaterinburg, Russia, and then transported by truck to the final destination in a secure nuclear facility at Chelyabinsk, Russia. This shipment of 23.7 kg of HEU was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in close cooperation with the Rosatom State Atomic Energy Corporation and the International Atomic Energy Agency, and was managed in Romania by the National Commission for Nuclear Activities Control (CNCAN). This paper describes the planning, shipment preparations, equipment design, and license approvals that resulted in the safe and secure air shipment of this spent nuclear fuel.

Igor Bolshinsky; Ken Allen; Lucian Biro; Alexander Buchelnikov

2010-10-01T23:59:59.000Z

176

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

177

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

178

Impact of Fuel Failure on Criticality Safety of Used Nuclear Fuel  

SciTech Connect (OSTI)

Commercial used nuclear fuel (UNF) in the United States is expected to remain in storage for considerably longer periods than originally intended (e.g., <40 years). Extended storage (ES) time and irradiation of nuclear fuel to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, can result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF is not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on criticality safety of UNF in storage and transportation casks. This effort is primarily motivated by concerns related to the potential for fuel degradation during ES periods and transportation following ES. The criticality analyses consider representative UNF designs and cask systems and a range of fuel enrichments, burnups, and cooling times. The various failed-fuel configurations considered are designed to bound the anticipated effects of individual rod and general cladding failure, fuel rod deformation, loss of neutron absorber materials, degradation of canister internals, and gross assembly failure. The results quantify the potential impact on criticality safety associated with fuel reconfiguration and may be used to guide future research, design, and regulatory activities. Although it can be concluded that the criticality safety impacts of fuel reconfiguration during transportation subsequent to ES are manageable, the results indicate that certain configurations can result in a large increase in the effective neutron multiplication factor, k{sub eff}. Future work to inform decision making relative to which configurations are credible, and therefore need to be considered in a safety evaluation, is recommended.

Marshall, William BJ J [ORNL] [ORNL; Wagner, John C [ORNL] [ORNL

2012-01-01T23:59:59.000Z

179

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

180

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

Note: This page contains sample records for the topic "nuclear fuels limited" 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

Changing Biomass, Fossil, and Nuclear Fuel Cycles for Sustainability  

SciTech Connect (OSTI)

The energy and chemical industries face two great sustainability challenges: the need to avoid climate change and the need to replace crude oil as the basis of our transport and chemical industries. These challenges can be met by changing and synergistically combining the fossil, biomass, and nuclear fuel cycles.

Forsberg, Charles W [ORNL

2007-01-01T23:59:59.000Z

182

Enduring Nuclear Fuel Cycle, Proceedings of a panel discussion  

SciTech Connect (OSTI)

The panel reviewed the complete nuclear fuel cycle in the context of alternate energy resources, energy need projections, effects on the environment, susceptibility of nuclear materials to theft, diversion, and weapon proliferation. We also looked at ethical considerations of energy use, as well as waste, and its effects. The scope of the review extended to the end of the next century with due regard for world populations beyond that period. The intent was to take a long- range view and to project, not forecast, the future based on ethical rationales, and to avoid, as often happens, long-range discussions that quickly zoom in on only the next few decades. A specific nuclear fuel cycle technology that could satisfy these considerations was described and can be applied globally.

Walter, C. E., LLNL

1997-11-18T23:59:59.000Z

183

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

184

Polymer Electrolyte Fuel Cell Lifetime Limitations: The Role of  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustion | Department ofT ib l L d FNEPA/309 Reviewers | Department of EnergyDepartment

185

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

186

Utilization of Minor Actinides as a Fuel Component for Ultra-Long Life Bhr Configurations: Designs, Advantages and Limitations  

SciTech Connect (OSTI)

This project assessed the advantages and limitations of using minor actinides as a fuel component to achieve ultra-long life Very High Temperature Reactor (VHTR) configurations. Researchers considered and compared the capabilities of pebble-bed and prismatic core designs with advanced actinide fuels to achieve ultra-long operation without refueling. Since both core designs permit flexibility in component configuration, fuel utilization, and fuel management, it is possible to improve fissile properties of minor actinides by neutron spectrum shifting through configuration adjustments. The project studied advanced actinide fuels, which could reduce the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository and enable recovery of the energy contained in spent fuel. The ultra-long core life autonomous approach may reduce the technical need for additional repositories and is capable to improve marketability of the Generation IV VHTR by allowing worldwide deployment, including remote regions and regions with limited industrial resources. Utilization of minor actinides in nuclear reactors facilitates developments of new fuel cycles towards sustainable nuclear energy scenarios.

Dr. Pavel V. Tsvetkov

2009-05-20T23:59:59.000Z

187

Nuclear-Renewable Hybrid System Economic Basis for Electricity, Fuel, and Hydrogen  

SciTech Connect (OSTI)

Concerns about climate change and altering the ocean chemistry are likely to limit the use of fossil fuels. That implies a transition to a low-carbon nuclear-renewable electricity grid. Historically variable electricity demand was met using fossil plants with low capital costs, high operating costs, and substantial greenhouse gas emissions. However, the most easily scalable very-low-emissions generating options, nuclear and non-dispatchable renewables (solar and wind), are capital-intensive technologies with low operating costs that should operate at full capacities to minimize costs. No combination of fully-utilized nuclear and renewables can meet the variable electricity demand. This implies large quantities of expensive excess generating capacity much of the time. In a free market this results in near-zero electricity prices at times of high nuclear renewables output and low electricity demand with electricity revenue collapse. Capital deployment efficiency—the economic benefit derived from energy systems capital investment at a societal level—strongly favors high utilization of these capital-intensive systems, especially if low-carbon nuclear renewables are to replace fossil fuels. Hybrid energy systems are one option for better utilization of these systems that consumes excess energy at times of low prices to make some useful product.The economic basis for development of hybrid energy systems is described for a low-carbon nuclear renewable world where much of the time there are massivequantities of excess energy available from the electric sector.Examples include (1) high-temperature electrolysis to generate hydrogen for non-fossil liquid fuels, direct use as a transport fuel, metal reduction, etc. and (2) biorefineries.Nuclear energy with its concentrated constant heat output may become the enabling technology for economically-viable low-carbon electricity grids because hybrid nuclear systems may provide an economic way to produce dispatachable variable electricity with economic base-load operation of the reactor.

Charles Forsberg; Steven Aumeier

2014-04-01T23:59:59.000Z

188

TOWARDS BENCHMARK MEASUREMENTS FOR USED NUCLEAR FUEL ASSAY USING A LEAD SLOWING-DOWN SPECTROMETER  

E-Print Network [OSTI]

for spent fuel testing. The characterization of spent fuel is particularly important for nuclear safeguardsTOWARDS BENCHMARK MEASUREMENTS FOR USED NUCLEAR FUEL ASSAY USING A LEAD SLOWING-DOWN SPECTROMETER B) is considered as a possible option for non- destructive assay of fissile material in used nuclear fuel

Danon, Yaron

189

Polymer Electrolyte Fuel Cell Lifetime Limitations: The Role...  

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

More Documents & Publications Non-Platinum Bimetallic Cathode Electrocatalysts Testing Oxygen Reduction Reaction Activity with the Rotating Disc Electrode Technique DOE Fuel Cell...

190

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

191

Uncanistered Spent Nuclear fuel Disposal Container System Description Document  

SciTech Connect (OSTI)

The Uncanistered Spent Nuclear Fuel (SNF) Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in the emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Multiple boiling water reactor (BWR) and pressurized water reactor (PWR) disposal container designs are needed to accommodate the expected range of spent fuel assemblies and provide long-term confinement of the commercial SNF. The disposal container will include outer and inner cylinder walls, outer cylinder lids (two on the top, one on the bottom), inner cylinder lids (one on the top, one on the bottom), and an internal metallic basket structure. Exterior labels will provide a means by which to identify the disposal container and its contents. The two metal cylinders, in combination with the cladding, Emplacement Drift System, drip shield, and natural barrier, will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel and the outer cylinder and outer cylinder lid will be made of high-nickel alloy. The basket will assist criticality control, provide structural support, and improve heat transfer. The Uncanistered SNF Disposal Container System interfaces with the emplacement drift environment and internal waste by transferring heat from the SNF to the external environment and by protecting the SFN assemblies and their contents from damage/degradation by the external environment. The system also interfaces with the SFN by limiting access of moderator and oxidizing agents of the SFN. The waste package interfaces with the Emplacement Drift System's emplacement drift pallets upon which the wasted packages are placed. The disposal container interfaces with the Assembly Transfer System, Waste Emplacement/Retrieval System, Disposal Container Handling System, and Waste Package Remediation System during loading, handling, transfer, emplacement and retrieval of the disposal container/waste package.

NONE

2000-10-12T23:59:59.000Z

192

Safe Advantage on Dry Interim Spent Nuclear Fuel Storage  

SciTech Connect (OSTI)

This paper aims to present the advantages of dry cask storage in comparison with the wet storage (cooling water pools) for SNF. When the nuclear fuel is removed from the core reactor, it is moved to a storage unit and it wait for a final destination. Generally, the spent nuclear fuel (SNF) remains inside water pools within the reactors facility for the radioactive activity decay. After some period of time in pools, SNF can be sent to a definitive deposition in a geological repository and handled as radioactive waste or to reprocessing facilities, or still, wait for a future solution. Meanwhile, SNF remains stored for a period of time in dry or wet facilities, depending on the method adopted by the nuclear power plant or other plans of the country. Interim storage, up to 20 years ago, was exclusively wet and if the nuclear facility had to be decommissioned another storage solution had to be found. At the present time, after a preliminary cooling of the SNF elements inside the water pool, the elements can be stored in dry facilities. This kind of storage does not need complex radiation monitoring and it is safer then wet one. Casks, either concrete or metallic, are safer, especially on occurrence of earthquakes, like that occurred at Kashiwazaki-Kariwa nuclear power plant, in Japan on July 16, 2007. (authors)

Romanato, L.S. [Centro Tecnologico da Marinha em S.Paulo, Brazilian Navy Technological Center, Sao Paulo (Brazil)

2008-07-01T23:59:59.000Z

193

22.351 Systems Analysis of the Nuclear Fuel Cycle, Spring 2003  

E-Print Network [OSTI]

In-depth technical and policy analysis of various options for the nuclear fuel cycle. Topics include uranium supply, enrichment fuel fabrication, in-core physics and fuel management of uranium, thorium and other fuel types, ...

Kazimi, Mujid S.

194

E-Print Network 3.0 - alternative nuclear fuel Sample Search...  

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

Summary: of electricity from nuclear power plants is far less than any of the alternative energy technologies now contem... Processing of Nuclear Fuel, EGRN 430 ...

195

Transient Testing of Nuclear Fuels and Materials in United States  

SciTech Connect (OSTI)

The US Department of Energy (DOE) has been engaged in an effort to develop and qualify next generation LWR fuel with enhanced performance and safety and reduced waste generation since 2010. This program, which has emphasized collaboration between the DOE, U.S. national laboratories and nuclear industry, was refocused from enhanced performance to enhanced accident tolerance following the events at Fukushima in 2011. Accident tolerant fuels have been specifically described as fuels that, in comparison with standard UO2-Zircaloy, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, as well as design-basis and beyond design-basis events. The program maintains an ambitious goal to insert a lead test assembly (LTA) of the new design into a commercial power reactor by 2022 .

Daniel M. Wachs

2012-12-01T23:59:59.000Z

196

Laser shockwave technique for characterization of nuclear fuel plate interfaces  

SciTech Connect (OSTI)

The US National Nuclear Security Agency is tasked with minimizing the worldwide use of high-enriched uranium. One aspect of that effort is the conversion of research reactors to monolithic fuel plates of low-enriched uranium. The manufacturing process includes hot isostatic press bonding of an aluminum cladding to the fuel foil. The Laser Shockwave Technique (LST) is here evaluated for characterizing the interface strength of fuel plates using depleted Uranium/Mo foils. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves and is therefore well adapted to the quality assurance of this process. Preliminary results show a clear signature of well-bonded and debonded interfaces and the method is able to classify/rank the bond strength of fuel plates prepared under different HIP conditions.

Perton, M.; Levesque, D.; Monchalin, J.-P.; Lord, M. [National Research Council Canada, 75 de Mortagne Blvd, Boucherville, Quebec, J4B 6Y4 (Canada); Smith, J. A.; Rabin, B. H. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

2013-01-25T23:59:59.000Z

197

Laser Shockwave Technique For Characterization Of Nuclear Fuel Plate Interfaces  

SciTech Connect (OSTI)

The US National Nuclear Security Agency is tasked with minimizing the worldwide use of high-enriched uranium. One aspect of that effort is the conversion of research reactors to monolithic fuel plates of low-enriched uranium. The manufacturing process includes hot isostatic press bonding of an aluminum cladding to the fuel foil. The Laser Shockwave Technique (LST) is here evaluated for characterizing the interface strength of fuel plates using depleted Uranium/Mo foils. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves and is therefore well adapted to the quality assurance of this process. Preliminary results show a clear signature of well-bonded and debonded interfaces and the method is able to classify/rank the bond strength of fuel plates prepared under different HIP conditions.

James A. Smith; Barry H. Rabin; Mathieu Perton; Daniel Lévesque; Jean-Pierre Monchalin; Martin Lord

2012-07-01T23:59:59.000Z

198

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

199

Spent Nuclear Fuel (SNF) Project Design Basis Capacity Study  

SciTech Connect (OSTI)

This study of the design basis capacity of process systems was prepared by Fluor Federal Services for the Spent Nuclear Fuel Project. The evaluation uses a summary level model of major process sub-systems to determine the impact of sub-system interactions on the overall time to complete fuel removal operations. The process system model configuration and time cycle estimates developed in the original version of this report have been updated as operating scenario assumptions evolve. The initial document released in Fiscal Year (FY) 1996 varied the number of parallel systems and transport systems over a wide range, estimating a conservative design basis for completing fuel processing in a two year time period. Configurations modeling planned operations were updated in FY 1998 and FY 1999. The FY 1998 Base Case continued to indicate that fuel removal activities at the basins could be completed in slightly over 2 years. Evaluations completed in FY 1999 were based on schedule modifications that delayed the start of KE Basin fuel removal, with respect to the start of KW Basin fuel removal activities, by 12 months. This delay resulted in extending the time to complete all fuel removal activities by 12 months. However, the results indicated that the number of Cold Vacuum Drying (CVD) stations could be reduced from four to three without impacting the projected time to complete fuel removal activities. This update of the design basis capacity evaluation, performed for FY 2000, evaluates a fuel removal scenario that delays the start of KE Basin activities such that staffing peaks are minimized. The number of CVD stations included in all cases for the FY 2000 evaluation is reduced from three to two, since the scenario schedule results in minimal time periods of simultaneous fuel removal from both basins. The FY 2000 evaluation also considers removal of Shippingport fuel from T Plant storage and transfer to the Canister Storage Building for storage.

CLEVELAND, K.J.

2000-08-17T23:59:59.000Z

200

Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579  

SciTech Connect (OSTI)

General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2 discharge reuse. The EM2 waste disposal profile is effectively only fission products, which reduces the mass (about 3% vs LWR), average half life, heat and long term radio-toxicity of the disposal. Widespread implementation of EM2 fuel cycle is highly significant as it would increase world energy reserves; the remaining energy in U.S. LWR SNF alone exceeds that in the U.S. natural gas reserves. Unlike many LWR SNF disposition concepts, the EM2 fuel cycle conversion of SNF produces energy and associated revenue such that the overall project is cost effective. By providing conversion of SNF to fission products the fuel cycle is closed and a non-repository LWR SNF disposition path is created and overall repository requirements are significantly reduced. (authors)

Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D. [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)] [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)

2013-07-01T23:59:59.000Z

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


201

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

202

Mechanical Properties of Nuclear Fuel Surrogates using Picosecond Laser Ultrasonics  

SciTech Connect (OSTI)

Detailed understanding between microstructure evolution and mechanical properties is important for designing new high burnup nuclear fuels. In this presentation we discuss the use of picosecond ultrasonics to measure localize changes in mechanical properties of fuel surrogates. We develop measurement techniques that can be applied to investigate heterogeneous elastic properties caused by localize changes in chemistry, grain microstructure caused by recrystallization, and mechanical properties of small samples prepared using focused ion beam sample preparation. Emphasis is placed on understanding the relationship between microstructure and mechanical properties

David Hurley; Marat Khafizov; Farhad Farzbod; Eric Burgett

2013-05-01T23:59:59.000Z

203

Molten tin reprocessing of spent nuclear fuel elements  

DOE Patents [OSTI]

A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support the liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

Heckman, Richard A. (Castro Valley, CA)

1983-01-01T23:59:59.000Z

204

Financing Strategies For A Nuclear Fuel Cycle Facility  

SciTech Connect (OSTI)

To help meet the nation’s energy needs, recycling of partially used nuclear fuel is required to close the nuclear fuel cycle, but implementing this step will require considerable investment. This report evaluates financing scenarios for integrating recycling facilities into the nuclear fuel cycle. A range of options from fully government owned to fully private owned were evaluated using DPL (Decision Programming Language 6.0), which can systematically optimize outcomes based on user-defined criteria (e.g., lowest lifecycle cost, lowest unit cost). This evaluation concludes that the lowest unit costs and lifetime costs are found for a fully government-owned financing strategy, due to government forgiveness of debt as sunk costs. However, this does not mean that the facilities should necessarily be constructed and operated by the government. The costs for hybrid combinations of public and private (commercial) financed options can compete under some circumstances with the costs of the government option. This analysis shows that commercial operations have potential to be economical, but there is presently no incentive for private industry involvement. The Nuclear Waste Policy Act (NWPA) currently establishes government ownership of partially used commercial nuclear fuel. In addition, the recently announced Global Nuclear Energy Partnership (GNEP) suggests fuels from several countries will be recycled in the United States as part of an international governmental agreement; this also assumes government ownership. Overwhelmingly, uncertainty in annual facility capacity led to the greatest variations in unit costs necessary for recovery of operating and capital expenditures; the ability to determine annual capacity will be a driving factor in setting unit costs. For private ventures, the costs of capital, especially equity interest rates, dominate the balance sheet; and the annual operating costs, forgiveness of debt, and overnight costs dominate the costs computed for the government case. The uncertainty in operations, leading to lower than optimal processing rates (or annual plant throughput), is the most detrimental issue to achieving low unit costs. Conversely, lowering debt interest rates and the required return on investments can reduce costs for private industry.

David Shropshire; Sharon Chandler

2006-07-01T23:59:59.000Z

205

Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements  

SciTech Connect (OSTI)

In 1998, a major change in the technical strategy for managing Multi Canister Overpacks (MCO) while stored within the Canister Storage Building (CSB) occurred. The technical strategy is documented in Baseline Change Request (BCR) No. SNF-98-006, Simplified SNF Project Baseline (MCO Sealing) (FDH 1998). This BCR deleted the hot conditioning process initially adopted for the Spent Nuclear Fuel Project (SNF Project) as documented in WHC-SD-SNF-SP-005, Integrated Process Strategy for K Basins Spent Nuclear Fuel (WHC 199.5). In summary, MCOs containing Spent Nuclear Fuel (SNF) from K Basins would be placed in interim storage following processing through the Cold Vacuum Drying (CVD) facility. With this change, the needs for the Hot Conditioning System (HCS) and inerting/pressure retaining capabilities of the CSB storage tubes and the MCO Handling Machine (MHM) were eliminated. Mechanical seals will be used on the MCOs prior to transport to the CSB. Covers will be welded on the MCOs for the final seal at the CSB. Approval of BCR No. SNF-98-006, imposed the need to review and update the CSB functions and requirements baseline documented herein including changing the document title to ''Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements.'' This revision aligns the functions and requirements baseline with the CSB Simplified SNF Project Baseline (MCO Sealing). This document represents the Canister Storage Building (CSB) Subproject technical baseline. It establishes the functions and requirements baseline for the implementation of the CSB Subproject. The document is organized in eight sections. Sections 1.0 Introduction and 2.0 Overview provide brief introductions to the document and the CSB Subproject. Sections 3.0 Functions, 4.0 Requirements, 5.0 Architecture, and 6.0 Interfaces provide the data described by their titles. Section 7.0 Glossary lists the acronyms and defines the terms used in this document. Section 8.0 References lists the references used for this document.

KLEM, M.J.

2000-10-18T23:59:59.000Z

206

Letter Report: Looking Ahead at Nuclear Fuel Resources  

SciTech Connect (OSTI)

The future of nuclear energy and its ability to fulfill part of the world’s energy needs for centuries to come depend on a reliable input of nuclear fuel, either thorium or uranium. Obviously, the present nuclear fuel cycle is completely dependent on uranium. Future thorium cycles will also depend on 235U or fissile isotopes separated from used fuel to breed 232Th into fissile 233U. This letter report discusses several emerging areas of scientific understanding and technology development that will clarify and enable assured supplies of uranium and thorium well into the future. At the most fundamental level, the nuclear energy community needs to appreciate the origins of uranium and thorium and the processes of planetary accretion by which those materials have coalesced to form the earth and other planets. Secondly, the studies of geophysics and geochemistry are increasing understanding of the processes by which uranium and thorium are concentrated in various locations in the earth’s crust. Thirdly, the study of neutrinos and particularly geoneutrinos (neutrinos emitted by radioactive materials within the earth) has given an indication of the overall global inventories of uranium and thorium, though little indication for those materials’ locations. Crustal temperature measurements have also given hints of the vertical distribution of radioactive heat sources, primarily 238U and 232Th, within the continental crust. Finally, the evolving technologies for laser isotope separation are indicating methods for reducing the energy input to uranium enrichment but also for tailoring the isotopic vectors of fuels, burnable poisons and structural materials, thereby adding another tool for dealing with long-term waste management.

J. Stephen Herring

2013-09-01T23:59:59.000Z

207

Method for cleaning solution used in nuclear fuel reprocessing  

DOE Patents [OSTI]

Nuclear fuel processing solution consisting of tri-n-butyl phosphate and dodecane, with a complex of uranium, plutonium, or zirconium and with a solvent degradation product such as di-n-butyl phosphate therein, is contacted with an aqueous solution of a salt formed from hydrazine and either a dicarboxylic acid or a hydroxycarboxylic acid, thereby removing the aforesaid complex from the processing solution.

Tallent, O.K.; Crouse, D.J.; Mailen, J.C.

1980-12-17T23:59:59.000Z

208

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

209

Dry air oxidation kinetics of K-Basin spent nuclear fuel  

SciTech Connect (OSTI)

The safety and process analyses of the proposed Integrated Process Strategy (IPS) to move the N-Reactor spent nuclear fuel (SNF) stored at K-Basin to an interim storage facility require information about the oxidation behavior of the metallic uranium. Limited experiments have been performed on the oxidation reaction of SNF samples taken from an N-Reactor outer fuel element in various atmospheres. This report discusses studies on the oxidation behavior of SNF using two independent experimental systems: (1) a tube furnace with a flowing gas mixture of 2% oxygen/98% argon; and (2) a thermogravimetric system for dry air oxidation.

Abrefah, J.; Buchanan, H.C.; Gerry, W.M.; Gray, W.J.; Marschman, S.C.

1998-06-01T23:59:59.000Z

210

National briefing summaries: Nuclear fuel cycle and waste management  

SciTech Connect (OSTI)

Since 1976, the International Program Support Office (IPSO) at the Pacific Northwest Laboratory (PNL) has collected and compiled publicly available information concerning foreign and international radioactive waste management programs. This National Briefing Summaries is a printout of an electronic database that has been compiled and is maintained by the IPSO staff. The database contains current information concerning the radioactive waste management programs (with supporting information on nuclear power and the nuclear fuel cycle) of most of the nations (except eastern European countries) that now have or are contemplating nuclear power, and of the multinational agencies that are active in radioactive waste management. Information in this document is included for three additional countries (China, Mexico, and USSR) compared to the prior issue. The database and this document were developed in response to needs of the US Department of Energy.

Schneider, K.J.; Bradley, D.J.; Fletcher, J.F.; Konzek, G.J.; Lakey, L.T.; Mitchell, S.J.; Molton, P.M.; Nightingale, R.E.

1991-04-01T23:59:59.000Z

211

Effect of residual stress on the life prediction of dry storage canisters for used nuclear fuel  

E-Print Network [OSTI]

Used nuclear fuel dry storage canisters will likely be tasked with holding used nuclear fuel for a period longer than originally intended. Originally designed for 20 years, the storage time will likely approach 100 years. ...

Black, Bradley P. (Bradley Patrick)

2013-01-01T23:59:59.000Z

212

Technical strategy for the management of INEEL spent nuclear fuel  

SciTech Connect (OSTI)

This report presents evaluations, findings, and recommendations of the Idaho National Engineering and Environmental Laboratory (INEEL) Spent Nuclear Fuel Task Team. The technical strategy developed by the Task Team includes stabilization, near term storage, packaging, transport, and ultimate disposal. Key issues identified and discussed include waste characterization, criticality, packaging, waste form performance, and special fuels. Current plans focus on onsite needs, and include three central elements: (1) resolution of near-term vulnerabilities, (2) consolidation of storage locations, and (3) achieving dry storage in transportable packages. In addition to the Task Team report, appendices contain information on the INEEL spent fuel inventory; regulatory decisions and agreements; and analyses of criticality, packaging, storage, transportation, and system performance of a geological repository. 16 refs., 6 figs., 4 tabs.

NONE

1997-03-01T23:59:59.000Z

213

Categorization of Used Nuclear Fuel Inventory in Support of a Comprehensive National Nuclear Fuel Cycle Strategy - 13575  

SciTech Connect (OSTI)

A technical assessment of the current inventory [?70,150 metric tons of heavy metal (MTHM) as of 2011] of U.S.-discharged used nuclear fuel (UNF) has been performed to support decisions regarding fuel cycle strategies and research, development and demonstration (RD and D) needs. The assessment considered discharged UNF from commercial nuclear electricity generation and defense and research programs and determined that the current UNF inventory can be divided into the following three categories: 1. Disposal - excess material that is not needed for other purposes; 2. Research - material needed for RD and D purposes to support waste management (e.g., UNF storage, transportation, and disposal) and development of alternative fuel cycles (e.g., separations and advanced fuels/reactors); and 3. Recycle/Recovery - material with inherent and/or strategic value. A set of key assumptions and attributes relative to the various disposition options were used to categorize the current UNF inventory. Based on consideration of RD and D needs, time frames and material needs for deployment of alternative fuel cycles, characteristics of the current UNF inventory, and possible uses to support national security interests, it was determined that the vast majority of the current UNF inventory should be placed in the Disposal category, without the need to make fuel retrievable from disposal for reuse or research purposes. Access to the material in the Research and Recycle/Recovery categories should be retained to support RD and D needs and national security interests. This assessment does not assume any decision about future fuel cycle options or preclude any potential options, including those with potential recycling of commercial UNF. (authors)

Wagner, John C.; Peterson, Joshua L.; Mueller, Don E.; Gehin, Jess C.; Worrall, Andrew [Oak Ridge National Laboratory, P.O. Box 2008, Bldg. 5700, MS-6170, Oak Ridge, TN 37831 (United States)] [Oak Ridge National Laboratory, P.O. Box 2008, Bldg. 5700, MS-6170, Oak Ridge, TN 37831 (United States); Taiwo, Temitope; Nutt, Mark; Williamson, Mark A. [Argonne National Laboratory (United States)] [Argonne National Laboratory (United States); Todosow, Mike [Brookhaven National Laboratory (United States)] [Brookhaven National Laboratory (United States); Wigeland, Roald [Idaho National Laboratory (United States)] [Idaho National Laboratory (United States); Halsey, William G. [Lawrence Livermore National Laboratory (United States)] [Lawrence Livermore National Laboratory (United States); Omberg, Ronald P. [Pacific Northwest National Laboratory (United States)] [Pacific Northwest National Laboratory (United States); Swift, Peter N. [Sandia National Laboratories (United States)] [Sandia National Laboratories (United States); Carter, Joe [Savannah River National Laboratory (United States)] [Savannah River National Laboratory (United States)

2013-07-01T23:59:59.000Z

214

Nuclear fuel cycle assessment of India: a technical study for U.S.-India cooperation  

E-Print Network [OSTI]

The recent civil nuclear cooperation proposed by the Bush Administration and the Government of India has heightened the necessity of assessing India’s nuclear fuel cycle inclusive of nuclear materials and facilities. This agreement proposes...

Woddi, Taraknath Venkat Krishna

2009-05-15T23:59:59.000Z

215

Calculation Method for the Projection of Future Spent Nuclear Fuel Discharges  

SciTech Connect (OSTI)

This report describes the calculation method developed for the projection of future utility spent nuclear fuel (SNF) discharges in regard to their timing, quantity, burnup, and initial enrichment. This projection method complements the utility-supplied RW-859 data on historic discharges and short-term projections of SNF discharges by providing long-term projections that complete the total life cycle of discharges for each of the current U.S. nuclear power reactors. The method was initially developed in mid-1999 to update the SNF discharge projection associated with the 1995 RW-859 utility survey (CRWMS M&O 1996). and was further developed as described in Rev. 00 of this report (CRWMS M&O 2001a). Primary input to the projection of SNF discharges is the utility projection of the next five discharges from each nuclear unit, which is provided via the revised final version of the Energy Information Administration (EIA) 1998 RW-859 utility survey (EIA 2000a). The projection calculation method is implemented via a set of Excel 97 spreadsheets. These calculations provide the interface between receipt of the utility five-discharge projections that are provided in the RW-859 survey, and the delivery of projected life-cycle SNF discharge quantities and characteristics in the format requisite for performing logistics analysis to support design of the Civilian Radioactive Waste Management System (CRWMS). Calculation method improvements described in this report include the addition of a reactor-specific maximum enrichment-based discharge burnup limit. This limit is the consequence of the enrichment limit, currently 5 percent. which is imposed as a Nuclear Regulatory Commission (NRC) license condition on nuclear fuel fabrication plants. In addition, the calculation method now includes the capability for projecting future nuclear plant power upratings, consistent with many such recent plant uprates and the prospect of additional future uprates. Finally. this report summarizes the results of the 2002 Reference SNF Discharge Projection.

B. McLeod

2002-02-28T23:59:59.000Z

216

RIS-M-2575 REFERENCE NEUTRON RADIOGRAPHS OF NUCLEAR REACTOR FUEL  

E-Print Network [OSTI]

RISØ-M-2575 REFERENCE NEUTRON RADIOGRAPHS OF NUCLEAR REACTOR FUEL J. C. Domanus Abstract. Reference neutron radiographs of nuclear reactor fuel were produced by the Euraton Neutron Radiography Working Group-conpacted nuclear fuel pins. Those parts of the pins are shown where changes of ap- pearance differ from those

217

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

218

Synergistic Smart Fuel For In-pile Nuclear Reactor Measurements  

SciTech Connect (OSTI)

In March 2011, an earthquake of magnitude 9.0 on the Richter scale struck Japan with its epicenter on the northeast coast, near the Tohoku region. In addition to the immense physical destruction and casualties across the country, several nuclear power plants (NPP) were affected. It was the Fukushima Daiichi NPP that experienced the most severe and irreversible damage. The earthquake brought the reactors at Fukushima to an automatic shutdown and because the power transmission lines were damaged, emergency diesel generators (EDGs) were activated to ensure that there was continued cooling of the reactors and spent fuel pools. The situation was being successfully managed until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to the reactors.2 At this point, the situation became critical. There was a loss of the sensors and instrumentation within the reactor that could have provided valuable information to guide the operators to make informed decisions and avoid the unfortunate events that followed. In the light of these events, we have developed and tested a potential self-powered thermoacoustic system, which will have the ability to serve as a temperature sensor and can transmit data independently of electronic networks. Such a device is synergistic with the harsh environment of the nuclear reactor as it utilizes the heat from the nuclear fuel to provide the input power.

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

2013-10-01T23:59:59.000Z

219

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

220

EARTHQUAKE CAUSED RELEASES FROM A NUCLEAR FUEL CYCLE FACILITY  

SciTech Connect (OSTI)

The fuel cycle facility (FCF) at the Idaho National Laboratory is a nuclear facility which must be licensed in order to operate. A safety analysis is required for a license. This paper describes the analysis of the Design Basis Accident for this facility. This analysis involves a model of the transient behavior of the FCF inert atmosphere hot cell following an earthquake initiated breach of pipes passing through the cell boundary. The hot cell is used to process spent metallic nuclear fuel. Such breaches allow the introduction of air and subsequent burning of pyrophoric metals. The model predicts the pressure, temperature, volumetric releases, cell heat transfer, metal fuel combustion, heat generation rates, radiological releases and other quantities. The results show that releases from the cell are minimal and satisfactory for safety. This analysis method should be useful in other facilities that have potential for damage from an earthquake and could eliminate the need to back fit facilities with earthquake proof boundaries or lessen the cost of new facilities.

Charles W. Solbrig; Chad Pope; Jason Andrus

2014-08-01T23:59:59.000Z

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221

Nano-Scale Fission Product Phases in an Irradiated U-7Mo Alloy Nuclear Fuel  

SciTech Connect (OSTI)

Irradiated nuclear fuel is a very difficult material to characterize. Due to the large radiation fields associated with these materials, they are hard to handle and typically have to be contained in large hot cells. Even the equipment used for performing characterization is housed in hot cells or shielded glove boxes. The result is not only a limitation in the techniques that can be employed for characterization, but also a limitation in the size of features that can be resolved The most standard characterization techniques include light optical metallography (WM), scanning electron microscopy (SEM), and electron probe microanalysis (EPMA). These techniques are applied to samples that are typically prepared using grinding and polishing approaches that will always generate some mechanical damage on the sample surface. As a result, when performing SEM analysis, for example, the analysis is limited by the quality of the sample surface that can be prepared. However, a new approach for characterizing irradiated nuclear fuel has recently been developed at the Idaho National Laboratory (INL) in Idaho Falls, Idaho. It allows for a dramatic improvement in the quality of characterization that can be performed when using an instrument like an SEM. This new approach uses a dual-beam scanning microscope, where one of the beams isa focused ion beam (FIB), which can be used to generate specimens of irradiated fuel (-10µm x 10µm) for microstructural characterization, and the other beam is the electron beam of an SEM. One significant benefit of this approach is that the specimen surface being characterized has received much less damage (and smearing) than is caused by the more traditional approaches, which enables the imaging of nanometer­ sized microstructural features in the SEM. The process details are for an irradiated low-enriched uranium (LEU) U-Mo alloy fuel Another type of irradiated fuel that has been characterized using this technique is a mixed oxide fuel.

Dennis Keiser, Jr.; Brandon Miller; James Madden; Jan-Fong Jue; Jian Gan

2014-09-01T23:59:59.000Z

222

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

223

Overcoming Fuel Gas Containment Limitations to Energy Improvement  

E-Print Network [OSTI]

. For example, if the steam system were not fully understood, how would one know that an increase in waste heat steam production in one area of the refinery would not simply result in the venting of steam somewhere else? Or, without a detailed fuel balance... the development of the steam balance. One such package is Linnhoff March?s ProSteamTM software package (1). By including the power balance, the ProSteam model will mimic the behavior of the steam system and produce a realistic simulation of the site heat...

Davis, J.

2004-01-01T23:59:59.000Z

224

22.251 / 22.351 Systems Analysis of the Nuclear Fuel Cycle, Fall 2005  

E-Print Network [OSTI]

This course provides an in-depth technical and policy analysis of various options for the nuclear fuel cycle. Topics include uranium supply, enrichment fuel fabrication, in-core physics and fuel management of uranium, ...

Kazimi, Mujid S.

225

Corrosion of Spent Nuclear Fuel: The Long-Term Assessment  

SciTech Connect (OSTI)

Spent nuclear fuel, essentially U{sub 2}, accounts for over 95% of the total radioactivity of all of the radioactive wastes in the United States that require disposal, disposition or remediation. The UO{sub 2} in SNF is not stable under oxiding conditions and may also be altered under reducing conditions. The alteration of SNF results in the formation of new uranium phases that can cause the release or retardation of actinide and fission product radionuclides. Over the long term, and depending on the extent to which the secondary uranium phases incorporate fission products and actinides, these alteration phases become the near-field source term.

Rodney C. Ewing

2004-10-07T23:59:59.000Z

226

Welding fixture for nuclear fuel pin cladding assemblies  

DOE Patents [OSTI]

A welding fixture for locating a driver sleeve about the open end of a nuclear fuel pin cladding. The welding fixture includes a holder provided with an open cavity having shoulders for properly positioning the driver sleeve, the end cap, and a soft, high temperature resistant plastic protective sleeve that surrounds a portion of the end cap stem. Ejected contaminant particles spewed forth by closure of the cladding by pulsed magnetic welding techniques are captured within a contamination trap formed in the holder for ultimate removal and disposal of contaminating particles along with the holder.

Oakley, David J. (Richland, WA); Feld, Sam H. (West Richland, WA)

1986-01-01T23:59:59.000Z

227

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

228

Closure mechanism and method for spent nuclear fuel canisters  

DOE Patents [OSTI]

A canister is provided for storing, transporting, and/or disposing of spent nuclear fuel. The canister includes a canister shell, a top shield plug disposed within the canister, and a leak-tight closure arrangement. The closure arrangement includes a shear ring which forms a containment boundary of the canister, and which is welded to the canister shell and top shield plug. An outer seal plate, forming an outer seal, is disposed above the shear ring and is welded to the shield plug and the canister.

Doman, Marvin J. (Monroeville, PA)

2004-11-23T23:59:59.000Z

229

Process for recovery of palladium from nuclear fuel reprocessing wastes  

DOE Patents [OSTI]

Palladium is selectively removed from spent nuclear fuel reprocessing waste by adding sugar to a strong nitric acid solution of the waste to partially denitrate the solution and cause formation of an insoluble palladium compound. The process includes the steps of: (a) adjusting the nitric acid content of the starting solution to about 10 M; (b) adding 50% sucrose solution in an amount sufficient to effect the precipitation of the palladium compound; (c) heating the solution at reflux temperature until precipitation is complete; and (d) centrifuging the solution to separate the precipitated palladium compound from the supernatant liquid.

Campbell, D.O.; Buxton, S.R.

1980-06-16T23:59:59.000Z

230

Laser cutting apparatus for nuclear core fuel subassembly  

DOE Patents [OSTI]

The object of the invention is to provide a system and apparatus which employs laser cutting to disassemble a nuclear core fuel subassembly. The apparatus includes a gantry frame (C) which straddles the core fuel subassembly (14), an x-carriage (22) travelling longitudinally above the frame which carries a focus head assembly (D) having a vertically moving carriage (46) and a laterally moving carriage (52), a system of laser beam transferring and focusing mirrors carried by the x-carriage and focusing head assembly, and a shroud follower (F) and longitudinal follower (G) for following the shape of shroud (14) to maintain a beam focal point (44) fixed upon the shroud surface for accurate cutting.

Walch, Allan P. (Manchester, CT); Caruolo, Antonio B. (Vernon, CT)

1982-02-23T23:59:59.000Z

231

THERMODYNAMIC MODEL FOR URANIUM DIOXIDE BASED NUCLEAR FUEL  

SciTech Connect (OSTI)

Many projects involving nuclear fuel rest on a quantitative understanding of the co-existing phases at various stages of burnup. Since the many fission products have considerably different abilities to chemically associate with oxygen, and the oxygen-to-metal molar ratio is slowly changing, the chemical potential of oxygen is a function of burnup. Concurrently, well-recognized small fractions of new phases such as inert gas, noble metals, zirconates, etc. also develop. To further complicate matters, the dominant UO2 fuel phase may be non-stoichiometric and most of the minor phases themselves have a variable composition dependent on temperature and possible contact with the coolant in the event of a sheathing breach. A thermodynamic fuel model to predict the phases in partially burned CANDU (CANada Deuterium Uranium) nuclear fuel containing many major fission products has been under development. The building blocks of the model are the standard Gibbs energies of formation of the many possible compounds expressed as a function of temperature. To these data are added mixing terms associated with the appearance of the component species in particular phases. In operational terms, the treatment rests on the ability to minimize the Gibbs energy in a multicomponent system, in our case using the algorithms developed by Eriksson. The model is capable of handling non-stoichiometry in the UO2 fluorite phase, dilute solution behaviour of significant solute oxides, noble metal inclusions, a second metal solid solution U(Pd-Rh-Ru)3, zirconate, molybdate, and uranate solutions as well as other minor solid phases, and volatile gaseous species.

Thompson, Dr. William T. [Royal Military College of Canada; Lewis, Dr. Brian J [Royal Military College of Canada; Corcoran, E. C. [Royal Military College of Canada; Kaye, Dr. Matthew H. [Royal Military College of Canada; White, S. J. [Royal Military College of Canada; Akbari, F. [Atomic Energy of Canada Limited, Chalk River Laboratories; Higgs, Jamie D. [Atomic Energy of Canada Limited, Point Lepreau; Thompson, D. M. [Praxair Inc.; Besmann, Theodore M [ORNL; Vogel, S. C. [Los Alamos National Laboratory (LANL)

2007-01-01T23:59:59.000Z

232

In-Pile Thermal Conductivity Measurement Method for Nuclear Fuels  

SciTech Connect (OSTI)

Thermophysical properties of advanced nuclear fuels and materials during irradiation must be known prior to their use in existing, advanced, or next generation reactors. Thermal conductivity is one of the most important properties for predicting fuel and material performance. A joint Utah State University (USU) / Idaho National Laboratory (INL) project, which is being conducted with assistance from the Institute for Energy Technology at the Norway Halden Reactor Project, is investigating in-pile fuel thermal conductivity measurement methods. This paper focuses on one of these methods – a multiple thermocouple method. This two-thermocouple method uses a surrogate fuel rod with Joule heating to simulate volumetric heat generation to gain insights about in-pile detection of thermal conductivity. Preliminary results indicated that this method can measure thermal conductivity over a specific temperature range. This paper reports the thermal conductivity values obtained by this technique and compares these values with thermal property data obtained from standard thermal property measurement techniques available at INL’s High Test Temperature Laboratory. Experimental results and material properties data are also compared to finite element analysis results.

Joy L. Rempe; Brandon Fox; Heng Ban; Joshua E. Daw; Darrell L. Knudson; Keith G. Condie

2009-08-01T23:59:59.000Z

233

Delayed Gamma-Ray Spectroscopy for Spent Nuclear Fuel Assay  

SciTech Connect (OSTI)

High-energy, beta-delayed gamma-ray spectroscopy is investigated as a non-destructive assay technique for the determination of plutonium mass in spent nuclear fuel. This approach exploits the unique isotope-specific signatures contained in the delayed gamma-ray emission spectra detected following active interrogation with an external neutron source. A high fidelity modeling approach is described that couples radiation transport, analytical decay/depletion, and a newly developed gamma-ray emission source reconstruction code. Initially simulated and analyzed was a “one-pass” delayed gamma-ray assay that focused on the long-lived signatures. Also presented are the results of an independent study that investigated “pulsed mode” measurements, to capture the more isotope-specific, short-lived signatures. Initial modeling results outlined in this paper suggest that delayed gamma-ray assay of spent nuclear fuel assemblies can be accomplished with a neutron generator of sufficient strength and currently available gamma-ray detectors.

Campbell, Luke W.; Hunt, Alan W.; Ludewigt, Bernhard A.; Mozin, Vladimir V.

2012-04-01T23:59:59.000Z

234

Spent nuclear fuel recycling with plasma reduction and etching  

DOE Patents [OSTI]

A method of extracting uranium from spent nuclear fuel (SNF) particles is disclosed. Spent nuclear fuel (SNF) (containing oxides of uranium, oxides of fission products (FP) and oxides of transuranic (TRU) elements (including plutonium)) are subjected to a hydrogen plasma and a fluorine plasma. The hydrogen plasma reduces the uranium and plutonium oxides from their oxide state. The fluorine plasma etches the SNF metals to form UF6 and PuF4. During subjection of the SNF particles to the fluorine plasma, the temperature is maintained in the range of 1200-2000 deg K to: a) allow any PuF6 (gas) that is formed to decompose back to PuF4 (solid), and b) to maintain stability of the UF6. Uranium (in the form of gaseous UF6) is easily extracted and separated from the plutonium (in the form of solid PuF4). The use of plasmas instead of high temperature reactors or flames mitigates the high temperature corrosive atmosphere and the production of PuF6 (as a final product). Use of plasmas provide faster reaction rates, greater control over the individual electron and ion temperatures, and allow the use of CF4 or NF3 as the fluorine sources instead of F2 or HF.

Kim, Yong Ho

2012-06-05T23:59:59.000Z

235

EDF Nuclear Power Plants Operating Experience with MOX fuel  

SciTech Connect (OSTI)

EDF started Plutonium recycling in PWR in 1987 and progressively all the 20 reactors, licensed in using MOX fuel, have been loaded with MOX assemblies. At the origin of MOX introduction, these plants operated at full power in base load and the core management limited the irradiation time of MOX fuel assemblies to 3 annual cycles. Since 1995 all these reactors can operate in load follow mode. Since that time, a large amount of experience has been accumulated. This experience is very positive considering: - Receipt, handling, in core behaviour, pool storage and shipment of MOX fuel; - Operation of the various systems of the plant; - Environment impact; - Radioprotection; - Safety file requirements; - Availability for the grid. In order to reduce the fuel cost and to reach a better adequacy between UO{sub 2} fuel reprocessing flow and plutonium consumption, EDF had decided to improve the core management of MOX plants. This new core management call 'MOX Parity' achieves parity for MOX and UO{sub 2} assemblies in term of discharge burn-up. Compared to the current MOX assembly the Plutonium content is increased from 7,08% to 8,65% (equivalent to natural uranium enriched to respectively 3,25% and 3,7%) and the maximum MOX assembly burn-up moves from 42 to 52 GWd/t. This amount of burn-up is obtained from loading MOX assemblies for one additional annual cycle. Some, but limited, adaptations of the plant are necessary. In addition a new MOX fuel assembly has been designed to comply with the safety criteria taking into account the core management performances. These design improvements are based on the results of an important R and D program including numerous experimental tests and post-irradiated fuel examinations. In particular, envelope conditions compared to MOX Parity neutronic solicitations has been extensively investigated in order to get a full knowledge of the in reactor fuel behavior. Moreover, the operating conditions of the plant have been evaluated in many details and finally no important impact is anticipated. The industrial maturity of plutonium recycling activities is fully demonstrated and a new progress can be done with a complete confidence. The licensing process of 'MOX Parity' core management is in progress and its implementation on the 20 PWR is now expected at mid 2007. (author)

Thibault, Xavier [EDF Generation, Tour EDF Part Dieu - 9 rue des Cuirassiers B.P.3181 - 69402 Lyon Cedex 03 (France)

2006-07-01T23:59:59.000Z

236

RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA Energy and Environment

Nero, A.V.

2010-01-01T23:59:59.000Z

237

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network [OSTI]

electricity generation capacity and operating efficiency of nuclear plants [Nuclear Plant Capacity Factor Nuclear Electricity Generationelectricity generation capacity and operating efficiency of nu- clear plants [

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

238

Spacer grid for reducing bowing in a nuclear fuel assembly  

SciTech Connect (OSTI)

A bi-metallic spacer grid having a zircaloy perimeter strip consisting of oppositely facing, thin walled metal plates for closely surrounding the array of fuel rods. A rigid, stainless steel cross member extends between internal surfaces of the oppositely facing perimeter plates. In the preferred embodiment, the perimeter plates have cantilevered portions extending above and below the main body of the perimeter strip. The cross members interact with the enlarged portion by urging them outward relative to the perimeter strip as the fuel assembly heats up during operation. The outwardly projecting interface surfaces of each assembly mechanically interact with the interface surfaces of adjacent assemblies providing a mechanical restraint which limits bowing of the assembly. The effectiveness of the spacer grids in limiting bowing is therefore not dependent upon controlling the mechanisms responsible for causing bow. When the reactor is in a cold condition such as during refueling , the exterior dimensions of the spacer grids are the same as those of the other zircaloy grids, which assures adequate clearance for insertion and withdrawal of individual fuel assemblies.

Wohlsen, W.D.

1982-04-20T23:59:59.000Z

239

Truck and rail charges for shipping spent fuel and nuclear waste  

SciTech Connect (OSTI)

The Pacific Northwest Laboratory developed techniques for calculating estimates of nuclear-waste shipping costs and compiled a listing of representative data that facilitate incorporation of reference shipping costs into varius logistics analyses. The formulas that were developed can be used to estimate costs that will be incurred for shipping spent fuel or nuclear waste by either legal-weight truck or general-freight rail. The basic data for this study were obtained from tariffs of a truck carrier licensed to serve the 48 contiguous states and from various rail freight tariff guides. Also, current transportation regulations as issued by the US Department of Transportation and the Nuclear Regulatory Commission were investigated. The costs that will be incurred for shipping spent fuel and/or nuclear waste, as addressed by the tariff guides, are based on a complex set of conditions involving the shipment origin, route, destination, weight, size, and volume and the frequency of shipments, existing competition, and the length of contracts. While the complexity of these conditions is an important factor in arriving at a ''correct'' cost, deregulation of the transportation industry means that costs are much more subject to negotiation and, thus, the actual fee that will be charged will not be determined until a shipping contract is actually signed. This study is designed to provide the baseline data necessary for making comparisons of the estimated costs of shipping spent fuel and/or nuclear wastes by truck and rail transportation modes. The scope of the work presented in this document is limited to the costs incurred for shipping, and does not include packaging, cask purchase/lease costs, or local fees placed on shipments of radioactive materials.

McNair, G.W.; Cole, B.M.; Cross, R.E.; Votaw, E.F.

1986-06-01T23:59:59.000Z

240

A Tomographic Method for Verification of the Integrity of Spent Nuclear Fuel  

E-Print Network [OSTI]

A Tomographic Method for Verification of the Integrity of Spent Nuclear Fuel S. Jacobsson.g. in medicine. This thesis describes a tomographic method developed for measurements on nuclear fuel assemblies of the integrity of the assemblies, i.e. for controlling that all fuel rods are present. The application has been

Haviland, David

Note: This page contains sample records for the topic "nuclear fuels limited" 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

Science based integrated approach to advanced nuclear fuel development - vision, approach, and overview  

SciTech Connect (OSTI)

Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Rcactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems is critical. In order to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. The purpose of this paper is to identify the modeling and simulation approach in order to deliver predictive tools for advanced fuels development. The coordination between experimental nuclear fuel design, development technical experts, and computational fuel modeling and simulation technical experts is a critical aspect of the approach and naturally leads to an integrated, goal-oriented science-based R & D approach and strengthens both the experimental and computational efforts. The Advanced Fuels Campaign (AFC) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Integrated Performance and Safety Code (IPSC) are working together to determine experimental data and modeling needs. The primary objective of the NEAMS fuels IPSC project is to deliver a coupled, three-dimensional, predictive computational platform for modeling the fabrication and both normal and abnormal operation of nuclear fuel pins and assemblies, applicable to both existing and future reactor fuel designs. The science based program is pursuing the development of an integrated multi-scale and multi-physics modeling and simulation platform for nuclear fuels. This overview paper discusses the vision, goals and approaches how to develop and implement the new approach.

Unal, Cetin [Los Alamos National Laboratory; Pasamehmetoglu, Kemal [IDAHO NATIONAL LAB; Carmack, Jon [IDAHO NATIONAL LAB

2010-01-01T23:59:59.000Z

242

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

243

EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs  

Broader source: Energy.gov [DOE]

Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs

244

Temperature limits for LMFBR fuel cladding under upset and emergency operating conditions  

SciTech Connect (OSTI)

LMFBR fuel pin cladding tube is subjected to high transient temperatures during incidents such as pump trip, pump to grid plate pipe rupture etc. It is required to know temperature limits under such transient operating conditions for components involved while analyzing such incidents. A methodology for deriving such limits for fuel clad tube is worked out in this paper by making use of the transient damage correlation proposed by W.F. Brizes et. al.

Govindarajan, S.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam Tamilnadu (India). Nuclear Systems Division

1996-07-01T23:59:59.000Z

245

National briefing summaries: Nuclear fuel cycle and waste management  

SciTech Connect (OSTI)

The National Briefing Summaries is a compilation of publicly available information concerning the nuclear fuel cycle and radioactive waste management strategies and programs of 21 nations, including the United States and three international agencies that have publicized their activities in this field. It presents available highlight information with references that may be used by the reader for additional information. The information in this document is compiled primarily for use by the US Department of Energy and other US federal agencies and their contractors to provide summary information on radioactive waste management activities in other countries. This document provides an awareness to managers and technical staff of what is occurring in other countries with regard to strategies, activities, and facilities. The information may be useful in program planning to improve and benefit United States' programs through foreign information exchange. Benefits to foreign exchange may be derived through a number of exchange activities.

Schneider, K.J.; Lakey, L.T.; Silviera, D.J.

1988-12-01T23:59:59.000Z

246

Investigation of Electrochemical Recovery of Zirconium from Spent Nuclear Fuels  

SciTech Connect (OSTI)

This project uses both modeling and experimental studies to design optimal electrochemical technology methods for recovery of zirconium from used nuclear fuel rods for more effective waste management. The objectives are to provide a means of efficiently separating zirconium into metallic high-level waste forms and to support development of a process for decontamination of zircaloy hulls to enable their disposal as low- and intermediate-level waste. Modeling work includes extension of a 3D model previously developed by Seoul National University for uranium electrorefining by adding the ability to predict zirconium behavior. Experimental validation activities include tests for recovery of zirconium from molten salt solutions and aqueous tests using surrogate materials. *This is a summary of the FY 2013 progress for I-NERI project # 2010-001-K provided to the I-NERI office.

Michael Simpson; II-Soon Hwang

2014-06-01T23:59:59.000Z

247

ULTRASONIC ARRAY TECHNIQUE FOR THE INSPECTION OF COPPER LINED CANISTERS FOR NUCLEAR WASTE FUEL  

E-Print Network [OSTI]

ULTRASONIC ARRAY TECHNIQUE FOR THE INSPECTION OF COPPER LINED CANISTERS FOR NUCLEAR WASTE FUEL and Waste Management Co.) for encapsulation of nuclear waste. Due to the radiation emitted by the nuclear, and characterization. The applicability of linear array technique for inspection of copper lined canisters for nuclear

248

Applications of nuclear data covariances to criticality safety and spent fuel characterization  

SciTech Connect (OSTI)

Covariance data computational methods and data used for sensitivity and uncertainty analysis within the SCALE nuclear analysis code system are presented. Applications in criticality safety calculations and used nuclear fuel analysis are discussed.

Williams, Mark L [ORNL] [ORNL; Ilas, Germina [ORNL] [ORNL; Marshall, William BJ J [ORNL] [ORNL; Rearden, Bradley T [ORNL] [ORNL

2014-01-01T23:59:59.000Z

249

On selection and operation of an international interim storage facility for spent nuclear fuel  

E-Print Network [OSTI]

Disposal of post-irradiation fuel from nuclear reactors has been an issue for the nuclear industry for many years. Most countries currently have no long-term disposal strategy in place. Therefore, the concept of an ...

Burns, Joe, 1966-

2004-01-01T23:59:59.000Z

250

Enrico Fermi Fast Reactor Spent Nuclear Fuel Criticality Calculations: Degraded Mode  

SciTech Connect (OSTI)

The objective of this calculation is to characterize the nuclear criticality safety concerns associated with the codisposal of the Department of Energy's (DOE) Enrico Fermi (EF) Spent Nuclear Fuel (SNF) in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP) and placed in a Monitored Geologic Repository (MGR). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (k{sub eff}) for the degraded mode internal configurations of the codisposal WP. The results of this calculation and those of Ref. 8 will be used to evaluate criticality issues and support the analysis that will be performed to demonstrate the viability of the codisposal concept for the Monitored Geologic Repository.

D.R. Moscalu; L. Angers; J. Monroe-Rammsey; H.R. Radulesca

2000-07-21T23:59:59.000Z

251

OECD/NEA Ongoing activities related to the nuclear fuel cycle  

SciTech Connect (OSTI)

As part of its role in encouraging international collaboration, the OECD Nuclear Energy Agency is coordinating a series of projects related to the Nuclear Fuel Cycle. The Nuclear Science Committee (NSC) Working Party on Scientific Issues of the Nuclear Fuel Cycle (WPFC) comprises five different expert groups covering all aspects of the fuel cycle from front to back-end. Activities related to fuels, materials, physics, separation chemistry, and fuel cycles scenarios are being undertaken. By publishing state-of-the-art reports and organizing workshops, the groups are able to disseminate recent research advancements to the international community. Current activities mainly focus on advanced nuclear systems, and experts are working on analyzing results and establishing challenges associated to the adoption of new materials and fuels. By comparing different codes, the Expert Group on Advanced Fuel Cycle Scenarios is aiming at gaining further understanding of the scientific issues and specific national needs associated with the implementation of advanced fuel cycles. At the back end of the fuel cycle, separation technologies (aqueous and pyrochemical processing) are being assessed. Current and future activities comprise studies on minor actinides separation and post Fukushima studies. Regular workshops are also organized to discuss recent developments on Partitioning and Transmutation. In addition, the Nuclear Development Committee (NDC) focuses on the analysis of the economics of nuclear power across the fuel cycle in the context of changes of electricity markets, social acceptance and technological advances and assesses the availability of the nuclear fuel and infrastructure required for the deployment of existing and future nuclear power. The Expert Group on the Economics of the Back End of the Nuclear Fuel Cycle (EBENFC), in particular, is looking at assessing economic and financial issues related to the long term management of spent nuclear fuel. (authors)

Cornet, S.M. [OECD Nuclear Energy Agency, 12 Boulevard des Iles, 92130 Issy-les-Moulineaux (France); McCarthy, K. [Idaho Nat. Lab. - P.O. Box 1625, Idaho Falls, ID 83415-3860 (United States); Chauvin, N. [CEA Saclay, Nuclear Energy Division, 91191 Gif/Yvette (France)

2013-07-01T23:59:59.000Z

252

Radioactive Iodine and Krypton Control for Nuclear Fuel Reprocessing Facilities  

SciTech Connect (OSTI)

The removal of volatile radionuclides generated during used nuclear fuel reprocessing in the US is almost certain to be necessary for the licensing of a reprocessing facility in the US. Various control technologies have been developed, tested, or used over the past 50 years for control of volatile radionuclide emissions from used fuel reprocessing plants. The US DOE has sponsored, since 2009, an Off-gas Sigma Team to perform research and development focused on the most pressing volatile radionuclide control and immobilization problems. In this paper, we focus on the control requirements and methodologies for 85Kr and 129I. Numerous candidate technologies have been studied and developed at laboratory and pilot-plant scales in an effort to meet the need for high iodine control efficiency and to advance alternatives to cryogenic separations for krypton control. Several of these show promising results. Iodine decontamination factors as high as 105, iodine loading capacities, and other adsorption parameters including adsorption rates have been demonstrated under some conditions for both silver zeolite (AgZ) and Ag-functionalized aerogel. Sorbents, including an engineered form of AgZ and selected metal organic framework materials (MOFs), have been successfully demonstrated to capture Kr and Xe without the need for separations at cryogenic temperatures.

N. R. Soelberg; J. D. Law; T. G. Garn; M. Greenhalgh; R. T. Jubin; P. Thallapally; D. M. Strachan

2013-08-01T23:59:59.000Z

253

Natural convection heat transfer within horizontal spent nuclear fuel assemblies  

SciTech Connect (OSTI)

Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array.

Canaan, R.E.

1995-12-01T23:59:59.000Z

254

Development of fission gas swelling and release models for metallic nuclear fuels  

E-Print Network [OSTI]

Fuel swelling and fission gas generation for fast reactor fuels are of high importance since they are among the main limiting factors in the development of metallic fast reactor fuel. Five new fission gas and swelling ...

Andrews, Nathan Christopher

2012-01-01T23:59:59.000Z

255

E-Print Network 3.0 - alloy nuclear fuels Sample Search Results  

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

...3 Ph... Waste disposal: Long term durability of UK spent nuclear fuel A 3.5-year PhD studentship is...

256

Sensitivity of economic performance of the nuclear fuel cycle to simulation modeling assumptions  

E-Print Network [OSTI]

Comparing different nuclear fuel cycles and assessing their implications require a fuel cycle simulation model as complete and realistic as possible. In this thesis, methodological implications of modeling choices are ...

Bonnet, Nicéphore

2007-01-01T23:59:59.000Z

257

Simulation of the nuclear fuel cycle with recycling : options and outcomes  

E-Print Network [OSTI]

A system dynamics simulation technique is applied to generate a new version of the CAFCA code to study the mass flow in the nuclear fuel cycle, and the impact of different options for advanced reactors and fuel recycling ...

Silva, Rodney Busquim e

2008-01-01T23:59:59.000Z

258

EIS-0306: Treatment and Management of Sodium-Bonded Spent Nuclear Fuel  

Broader source: Energy.gov [DOE]

This EIS evaluates the potential environmental impacts of the proposed electrometallurgical treatment of DOE-owned sodium bonded spent nuclear fuel in the Fuel Conditioning Facility at Argonne...

259

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

260

Used nuclear fuel storage options including implications of small modular reactors  

E-Print Network [OSTI]

This work addresses two aspects of the nuclear fuel cycle system with significant policy implications. The first is the preferred option for used fuel storage based on economics: local, regional or national storage. The ...

Brinton, Samuel O. (Samuel Otis)

2014-01-01T23:59:59.000Z

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


261

Nuclear envelope-limited chromatin sheets (ELCS) and heterochromatin higher order structure  

E-Print Network [OSTI]

REVIEW Nuclear envelope-limited chromatin sheets (ELCS) and heterochromatin higher order structure /Published online: 12 June 2009 # Springer-Verlag 2009 Abstract The interphase nucleus and nuclear envelope of these nuclear shapes and, consequently, a barrier to understanding the biochemical and biophysical causes

Olins, Ada L.

262

Full-Scale Cask Testing and Public Acceptance of Spent Nuclear Fuel Shipments - 12254  

SciTech Connect (OSTI)

Full-scale physical testing of spent fuel shipping casks has been proposed by the National Academy of Sciences (NAS) 2006 report on spent nuclear fuel transportation, and by the Presidential Blue Ribbon Commission (BRC) on America's Nuclear Future 2011 draft report. The U.S. Nuclear Regulatory Commission (NRC) in 2005 proposed full-scale testing of a rail cask, and considered 'regulatory limits' testing of both rail and truck casks (SRM SECY-05-0051). The recent U.S. Department of Energy (DOE) cancellation of the Yucca Mountain project, NRC evaluation of extended spent fuel storage (possibly beyond 60-120 years) before transportation, nuclear industry adoption of very large dual-purpose canisters for spent fuel storage and transport, and the deliberations of the BRC, will fundamentally change assumptions about the future spent fuel transportation system, and reopen the debate over shipping cask performance in severe accidents and acts of sabotage. This paper examines possible approaches to full-scale testing for enhancing public confidence in risk analyses, perception of risk, and acceptance of spent fuel shipments. The paper reviews the literature on public perception of spent nuclear fuel and nuclear waste transportation risks. We review and summarize opinion surveys sponsored by the State of Nevada over the past two decades, which show consistent patterns of concern among Nevada residents about health and safety impacts, and socioeconomic impacts such as reduced property values along likely transportation routes. We also review and summarize the large body of public opinion survey research on transportation concerns at regional and national levels. The paper reviews three past cask testing programs, the way in which these cask testing program results were portrayed in films and videos, and examines public and official responses to these three programs: the 1970's impact and fire testing of spent fuel truck casks at Sandia National Laboratories, the 1980's regulatory and demonstration testing of MAGNOX fuel flasks in the United Kingdom (the CEGB 'Operation Smash Hit' tests), and the 1980's regulatory drop and fire tests conducted on the TRUPACT II containers used for transuranic waste shipments to the Waste Isolation Pilot Plant in New Mexico. The primary focus of the paper is a detailed evaluation of the cask testing programs proposed by the NRC in its decision implementing staff recommendations based on the Package Performance Study, and by the State of Nevada recommendations based on previous work by Audin, Resnikoff, Dilger, Halstead, and Greiner. The NRC approach is based on demonstration impact testing (locomotive strike) of a large rail cask, either the TAD cask proposed by DOE for spent fuel shipments to Yucca Mountain, or a similar currently licensed dual-purpose cask. The NRC program might also be expanded to include fire testing of a legal-weight truck cask. The Nevada approach calls for a minimum of two tests: regulatory testing (impact, fire, puncture, immersion) of a rail cask, and extra-regulatory fire testing of a legal-weight truck cask, based on the cask performance modeling work by Greiner. The paper concludes with a discussion of key procedural elements - test costs and funding sources, development of testing protocols, selection of testing facilities, and test peer review - and various methods of communicating the test results to a broad range of stakeholder audiences. (authors)

Dilger, Fred [Black Mountain Research, Henderson, NV 81012 (United States); Halstead, Robert J. [State of Nevada Agency for Nuclear Projects Carson City, NV 80906 (United States); Ballard, James D. [Department of Sociology, California State University, Northridge Northridge, CA 91330 (United States)

2012-07-01T23:59:59.000Z

263

Limited Test Ban Treaty Signed | National Nuclear Security Administration  

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

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264

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

265

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

266

Application of Neutron-Absorbing Structural-Amorphous metal (SAM) Coatings for Spent Nuclear Fuel (SNF) Container to Enhance Criticality Safety Controls  

E-Print Network [OSTI]

241 Pu, etc. ). To prevent nuclear criticality in spent fuelto enhance criticality safety for spent nuclear fuel inSpent Nuclear Fuel (SNF) Container to Enhance Criticality

2006-01-01T23:59:59.000Z

267

SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES  

SciTech Connect (OSTI)

Waste packages are loaded with commercial spent nuclear fuel (SNF) that satisfies the minimum burnup requirements of a criticality loading curve. The burnup value assigned by the originating nuclear utility to each SNF assembly (assigned burnup) is used to load waste packages in compliance with a criticality loading curve. The burnup provided by a nuclear utility has uncertainties, so conservative calculation methods are used to characterize those uncertainties for incorporation into the criticality loading curves. Procedural safety controls ensure that the correct assembly is loaded into each waste package to prevent a misload that could create a condition affecting the safety margins. Probabilistic analyses show that procedural safety controls can minimize the chance of a misload but can not completely eliminate the possibility. Physical measurements of burnup with instrumentation in the surface facility are not necessary due to the conservative calculation methods used to produce the criticality loading curves. The reactor records assigned burnup of a commercial SNF assembly contains about two percent uncertainty, which is increased to five-percent to ensure conservatism. This five-percent uncertainty is accommodated by adjusting the criticality loading curve. Also, the record keeping methods of nuclear utilities are not uniform and the level of detail required by the NRC has varied over the last several decades. Thus, some SNF assemblies may have assigned burnups that are averages for a batch of assemblies with similar characteristics. Utilities typically have access to more detailed core-follow records that allow the batch average burnup to be changed to an assembly specific burnup. Alternatively, an additional safety margin is incorporated into the criticality loading curve to accommodate SNF assemblies with batch average burnups or greater uncertainties due to the methodology used by the nuclear utility. The utility records provide the assembly identifier, initial {sup 235}U enrichment, and time of discharge from the reactor as well as the assigned burnup, but the distribution. of burnup axially along the assembly length is not provided. The axial burnup profile is maintained within acceptable bounds by the operating conditions of the nuclear reactor and is calculated during preparations to reload a reactor, but the actual burnup profile is not measured. The axial burnup profile is important to the determination of the reactivity of a waste package, so a conservative evaluation of the calculated axial profiles for a large database of SNF has been performed. The product of the axial profile evaluation is a profile that is conservative. Thus, there is no need for physical measurement of the axial profile. The assembly identifier is legible on each SNF assembly and the utility records provide the associated characteristics of the assembly. The conservative methodologies used to determine the criticality loading curve for a waste package provide sufficient margin so that criticality safety is assured for preclosure operations even in the event of a misload. Consideration of misload effects for postclosure time periods is provided by the criticality Features, Events, and Processes (FEPs) analysis. The conservative approaches used to develop and apply the criticality loading curve are thus sufficiently robust that the utility assigned burnup is an adequate source of burnup values, and additional means of verification of assigned burnup through physical measurements are not needed.

BSC

2004-12-01T23:59:59.000Z

268

Radioactive Iodine and Krypton Control for Nuclear Fuel Reprocessing Facilities  

SciTech Connect (OSTI)

Nuclear fission results in the production of fission products and activation products, some of which tend to be volatile during used fuel reprocessing. These can evolve in volatile species in the reprocessing facility off-gas streams, depending on the separations and reprocessing technologies that are used. Radionuclides that have been identified as “volatile radionuclides” are noble gases (most notably isotopes of Kr and Xe); 3H; 14C; and 129I. Radionuclides that tend to form volatile species that evolve into reprocessing facility off-gas systems are more challenging to efficiently control compared to radionuclides that tend to stay in solid or liquid phases. Future used fuel reprocessing facilities in the United States can require efficient capture of some volatile radionuclides in their off-gas streams to meet regulatory emission requirements. In aqueous reprocessing, these radionuclides are most commonly expected to evolve into off-gas streams in tritiated water [3H2O (T2O) and 3HHO (THO)], radioactive CO2, noble gases, and gaseous HI, I2, or volatile organic iodides. The fate and speciation of these radionuclides from a non-aqueous fuel reprocessing facility is less well known at this time, but active investigations are in progress. An Off-Gas Sigma Team was formed in late FY 2009 to integrate and coordinate the Fuel Cycle Research and Development (FCR&D) activities directed towards the capture and sequestration of the these volatile radionuclides (Jubin 2012a). The Sigma Team concept was envisioned to bring together multidisciplinary teams from across the DOE complex that would work collaboratively to solve the technical challenges and to develop the scientific basis for the capture and immobilization technologies such that the sum of the efforts was greater than the individual parts. The Laboratories currently participating in this effort are Argonne National Laboratory (ANL), Idaho National Laboratory (INL), Oak Ridge National Laboratory (ORNL), Pacific Northwest National Laboratory (PNNL), and Sandia National Laboratories (SNL). This article focuses on control of volatile radionuclides that evolve during aqueous reprocessing of UNF. In particular, most of the work by the Off-gas Sigma Team has focused on the capture and sequestration of 129I and 85Kr, mainly because, as discussed below, control of 129I can require high efficiencies to meet regulatory requirements, and control of 85Kr using cryogenic processing, which has been the technology demonstrated and used commercially to date, can add considerable cost to a reprocessing facility.

Soelberg, Nicolas R. [Idaho National Laboratory, Idaho Falls, ID (United States); Garn, Troy [Idaho National Laboratory, Idaho Falls, ID (United States); Greenhalgh, Mitchell [Idaho National Laboratory, Idaho Falls, ID (United States); Law, Jack [Idaho National Laboratory, Idaho Falls, ID (United States); Jubin, Robert T. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Strachan, Denis M. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Thallapally, Praveen K. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

2013-07-22T23:59:59.000Z

269

Overview of reductants utilized in nuclear fuel reprocessing/recycling  

SciTech Connect (OSTI)

Most of the aqueous processes developed, or under consideration worldwide for the recycling of used nuclear fuel (UNF) utilize the oxido-reduction properties of actinides to separate them from other radionuclides. Generally, after acid dissolution of the UNF, (essentially in nitric acid solution), actinides are separated from the raffinate by liquid-liquid extraction using specific solvents, associated along the process, with a particular reductant that will allow the separation to occur. For example, the industrial PUREX process utilizes hydroxylamine as a plutonium reductant. Hydroxylamine has numerous advantages: not only does it have the proper attributes to reduce Pu(IV) to Pu(III), but it is also a non-metallic chemical that is readily decomposed to innocuous products by heating. However, it has been observed that the presence of high nitric acid concentrations or impurities (such as metal ions) in hydroxylamine solutions increase the likelihood of the initiation of an autocatalytic reaction. Recently there has been some interest in the application of simple hydrophilic hydroxamic ligands such as acetohydroxamic acid (AHA) for the stripping of tetravalent actinides in the UREX process flowsheet. This approach is based on the high coordinating ability of hydroxamic acids with tetravalent actinides (Np and Pu) compared with hexavalent uranium. Thus, the use of AHA offers a route for controlling neptunium and plutonium in the UREX process by complexant based stripping of Np(IV) and Pu(IV) from the TBP solvent phase, while U(VI) ions are not affected by AHA and remain solvated in the TBP phase. In the European GANEX process, AHA is also used to form hydrophilic complexes with actinides and strip them from the organic phase into nitric acid. However, AHA does not decompose completely when treated with nitric acid and hampers nitric acid recycling. In lieu of using AHA in the UREX + process, formohydroxamic acid (FHA), although not commercially available, hold promises as a replacement for AHA. FHA undergoes hydrolysis to formic acid which is volatile, thus allowing the recycling of nitric acid. Unfortunately, FHA powder was not stable in the experiments we ran in our laboratory. In addition, AHA and FHA also decompose to hydroxylamine which may undergo an autocatalytic reaction. Other reductants are available and could be extremely useful for actinides separation. The review presents the current plutonium reductants used in used nuclear fuel reprocessing and will introduce innovative and novel reductants that could become reducers for future research on UNF separation.

Patricia Paviet-Hartmann; Catherine Riddle; Keri Campbell; Edward Mausolf

2013-10-01T23:59:59.000Z

270

World nuclear fuel market: proceedings of the international conference on nuclear energy  

SciTech Connect (OSTI)

Thirteen papers, along with discussion and comments, are divided into four conference sessions covering: the prospect for primary markets for enriched uranium; secondary trading markets for enriched uranium; the management of irradiatied fuel and economics of reprocessing; and an evaluation of plutonium recycling in thermal reactors. The speakers address technical, economic, and political issues relating to both front-end and back-end management of the fuel cycle. The papers were presented at the 9th International Conference on Nuclear Energy in Nice, France during October, 1982. A separate abstract was prepared for each of the 13 papers selected for the Energy Data Base (EDB), Energy Research Abstracts (ERA), and Energy Abstracts for Policy Analysis (EAPA). (DCK)

Not Available

1982-01-01T23:59:59.000Z

271

Spent Nuclear Fuel Dry Transfer System Cold Demonstration Project Final Report  

SciTech Connect (OSTI)

The spent nuclear fuel dry transfer system (DTS) provides an interface between large and small casks and between storage-only and transportation casks. It permits decommissioning of reactor pools after shutdown and allows the use of large storage-only casks for temporary onsite storage of spent nuclear fuel irrespective of reactor or fuel handling limitations at a reactor site. A cold demonstration of the DTS prototype was initiated in August 1996 at the Idaho National Engineering and Environmental Laboratory (INEEL). The major components demonstrated included the fuel assembly handling subsystem, the shield plug/lid handling subsystem, the cask interface subsystem, the demonstration control subsystem, a support frame, and a closed circuit television and lighting system. The demonstration included a complete series of DTS operations from source cask receipt and opening through fuel transfer and closure of the receiving cask. The demonstration included both normal operations and recovery from off-normal events. It was designed to challenge the system to determine whether there were any activities that could be made to jeopardize the activities of another function or its safety. All known interlocks were challenged. The equipment ran smoothly and functioned as designed. A few "bugs" were corrected. Prior to completion of the demonstration testing, a number of DTS prototype systems were modified to apply lessons learned to date. Additional testing was performed to validate the modifications. In general, all the equipment worked exceptionally well. The demonstration also helped confirm cost estimates that had been made at several points in the development of the system.

Christensen, Max R; McKinnon, M. A.

1999-12-01T23:59:59.000Z

272

Nuclear power generation and fuel cycle report 1997  

SciTech Connect (OSTI)

Nuclear power is an important source of electric energy and the amount of nuclear-generated electricity continued to grow as the performance of nuclear power plants improved. In 1996, nuclear power plants supplied 23 percent of the electricity production for countries with nuclear units, and 17 percent of the total electricity generated worldwide. However, the likelihood of nuclear power assuming a much larger role or even retaining its current share of electricity generation production is uncertain. The industry faces a complex set of issues including economic competitiveness, social acceptance, and the handling of nuclear waste, all of which contribute to the uncertain future of nuclear power. Nevertheless, for some countries the installed nuclear generating capacity is projected to continue to grow. Insufficient indigenous energy resources and concerns over energy independence make nuclear electric generation a viable option, especially for the countries of the Far East.

NONE

1997-09-01T23:59:59.000Z

273

Characterization of a Stochastic Procedure for the Generation and Transport of Fission Fragments within Nuclear Fuels  

E-Print Network [OSTI]

With the ever-increasing demands of the nuclear power community to extend fuel cycles and overall core-lifetimes in a safe and economic manner, it is becoming more necessary to extend the working knowledge of nuclear fuel performance. From...

Hackemack, Michael Wayne

2013-04-15T23:59:59.000Z

274

Origin of Low Thermal Conductivity in Nuclear Fuels Quan Yin and Sergey Y. Savrasov  

E-Print Network [OSTI]

, the thermal conductivity of UO2 is very low, and the search for alternative materials continuesOrigin of Low Thermal Conductivity in Nuclear Fuels Quan Yin and Sergey Y. Savrasov Department in a very low thermal conductivity of modern nuclear fuels. Consider semiconducting UO2 which is a main

Savrasov, Sergej Y.

275

Status of radioiodine control for nuclear fuel reprocessing plants  

SciTech Connect (OSTI)

This report summarizes the status of radioiodine control in a nuclear fuel reprocessing plant with respect to capture, fixation, and disposal. Where possible, we refer the reader to a number of survey documents which have been published in the last four years. We provide updates where necessary. Also discussed are factors which must be considered in developing criteria for iodine control. For capture from gas streams, silver mordenite and a silver nitrate impregnated silica (AC-6120) are considered state-of-the-art and are recommended. Three aqueous scrubbing processes have been demonstrated: Caustic scrubbing is simple but probably will not give an adequate iodine retention by itself. Mercurex (mercuric nitrate-nitric acid scrubbing) has a number of disadvantages including the use of toxic mercury. Iodox (hyperazeotropic nitric acid scrubbing) is effective but employs a very corrosive and hazardous material. Other technologies have been tested but require extensive development. The waste forms recommended for long-term storage or disposal are silver iodide, the iodates of barium, strontium, or calcium, and silver loaded sorbents, all fixed in cement. Copper iodide in bitumen (asphalt) is a possibility but requires testing. The selection of a specific form will be influenced by the capture process used.

Burger, L.L.; Scheele, R.D.

1983-07-01T23:59:59.000Z

276

High Efficiency Generation of Hydrogen Fuels using Nuclear Power Annual Report August, 2000 - July 2001  

SciTech Connect (OSTI)

OAK B188 High Efficiency Generation of Hydrogen Fuels using Nuclear Power Annual Report August 2000 - July 2001. Currently no large scale, cost-effective, environmentally attractive hydrogen production process is available for commercialization nor has such a process been identified. Hydrogen is a promising energy carrier, which potentially could replace the fossil fuels used in the transportation sector of our economy. Carbon dioxide emissions from fossil fuel combustion are thought to be responsible for global warming. The purpose of this work is to determine the potential for efficient, cost-effective, large-scale production of hydrogen utilizing high temperature heat from an advanced nuclear power station. The benefits of this work will include the generation of a low-polluting transportable energy feedstock in an efficient method that has little or no implication for greenhouse gas emissions from a primary energy source whose availability and sources are domestically controlled. This will help to ensure energy for a future transportation/energy infrastructure that is not influenced/controlled by foreign governments. This report describes work accomplished during the second year (Phase 2) of a three year project whose objective is to ''define an economically feasible concept for production of hydrogen, by nuclear means, using an advanced high temperature nuclear reactor as the energy source.'' The emphasis of the first year (Phase 1) was to evaluate thermochemical processes which offer the potential for efficient, cost-effective, large-scale production of hydrogen from water, in which the primary energy input is high temperature heat from an advanced nuclear reactor and to select one (or, at most, three) for further detailed consideration. Phase 1 met its goals and did select one process, the sulfur-iodine process, for investigation in Phases 2 and 3. The combined goals of Phases 2 and 3 were to select the advanced nuclear reactor best suited to driving the selected thermochemical process and to define the selected reactor and process to the point that capital costs, operating costs and the resultant cost of hydrogen can be estimated. During original contract negotiation, it was necessary to reduce work scope to meet funding limits. As a result, the reactor interface and process will not be iterated to the point that only hydrogen is produced. Rather, hydrogen and electricity will be co-generated and the hydrogen cost will be stated as a function of the electricity sales price.

Brown, L.C.

2002-11-01T23:59:59.000Z

277

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

278

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

279

Savannah River Site, Spent Nuclear Fuel Management, Draft Environmental Impact Statement  

SciTech Connect (OSTI)

The proposed DOE action described in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets assigned to the Savannah River Site (SRS), including placing these materials in forms suitable for ultimate disposition. Options to treat, package, and store this material are discussed. The material included in this EIS consists of approximately 68 metric tons heavy metal (MTHM) of spent nuclear fuel (20 MTHM of aluminum-based spent nuclear fuel at SRS, as much as 28 MTHM of aluminum-clad spent nuclear fuel from foreign and domestic research reactors to be shipped to SRS through 2035, and 20 MTHM of stainless-steel or zirconium-clad spent nuclear fuel and some programmatic material stored at SRS for repackaging and dry storage pending shipment offsite).

N /A

1998-12-24T23:59:59.000Z

280

Technical data summary supporting the spent nuclear fuel environment impact statement, March 1994  

SciTech Connect (OSTI)

This report has been compiled by the WSRC Nuclear Materials Processing Division`s Planning Section at the request of the Office of Spent Fuel Management and Special Projects (EM-37) to support issuance of the Spent Nuclear Fuel Environmental Impact Statement. Savannah River Site input data evaluates five programmatic options (including {open_quotes}No Action{close_quotes}) ranging up to transfer of all DOE responsibility spent fuel to the SRS. For each option, a range of management/disposition scenarios has been examined. Each case summary provides information relative to the technical proposal, technical issues, environmental impacts, and projected costs for a forty year period (FY-35) when it is assumed that the material will be dispositioned from the SRS. The original issue of the report which was prepared under severe time constraints contained many simplifications and assumptions. Although the revisions have corrected some of the shortcomings of the original report, it is still highly recommended that significant additional study be performed before basing key decisions upon the data contained in this report. The data represents the best effort by a significant group of technical personnel familiar with nuclear materials processing, handling, and storage; but it is likely that careful scrutiny will reveal numerous discrepancies, inconsistencies and omissions. Nor does this report attempt to analyze every potential disposal pathway, but probably establishes the bounds for the most of the viable pathways. The bulk of the effort went into defining the engineering approaches necessary to execute the various mission scenarios which were changed since the last revision. The decision to limit reprocessing to only SRS aluminum clad required a major alteration of the TDS. Collection and/or calculation of much of the various waste, emission, and utility consumption data, so important to an EIS, has been updated since the last revision, but not thoroughly completed.

Geddes, R.L.; Claxton, R.E.; Lengel, J.D. [and others

1994-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear fuels limited" 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

Neutronics of accelerator-driven subcritical fission for burning transuranics in used nuclear fuel  

SciTech Connect (OSTI)

We report the development of a conceptual design for accelerator-driven subcritical fission in a molten salt core (ADSMS). ADSMS is capable of destroying all of the transuranics at the same rate and proportion as they are produced in a conventional nuclear power plant. The ADSMS core is fueled solely by transuranics extracted from used nuclear fuel and reduces its radiotoxicity by a factor 10,000. ADSMS offers a way to close the nuclear fuel cycle so that the full energy potential in the fertile fuels uranium and thorium can be recovered.

Sattarov, A.; Assadi, S.; Badgley, K.; Baty, A.; Comeaux, J.; Gerity, J.; Kellams, J.; Mcintyre, P.; Pogue, N.; Sooby, E.; Tsvetkov, P.; Rosaire, G. [Texas A and M University, College Station, TX 77845 (United States); Mann, T. [Argone National Laboratory, Argone, IL (United States)

2013-04-19T23:59:59.000Z

282

Range of Applicability and Bias Determination for Postclosure Criticality of Commercial Spent Nuclear Fuel  

SciTech Connect (OSTI)

The purpose of this calculation report, Range of Applicability and Bias Determination for Postclosure Criticality of Commercial Spent Nuclear Fuel, is to validate the computational method used to perform postclosure criticality calculations. The validation process applies the criticality analysis methodology approach documented in Section 3.5 of the Disposal Criticality Analysis Methodology Topical Report. The application systems for this validation consist of waste packages containing transport, aging, and disposal canisters (TAD) loaded with commercial spent nuclear fuel (CSNF) of varying assembly types, initial enrichments, and burnup values that are expected from the waste stream and of varying degree of internal component degradation that may occur over the 10,000-year regulatory time period. The criticality computational tool being evaluated is the general-purpose Monte Carlo N-Particle (MCNP) transport code. The nuclear cross-section data distributed with MCNP 5.1.40 and used to model the various physical processes are based primarily on the Evaluated Nuclear Data File/B Version VI (ENDF/B-VI) library. Criticality calculation bias and bias uncertainty and lower bound tolerance limit (LBTL) functions for CSNF waste packages are determined based on the guidance in ANSI/ANS 8.1-1998 (Ref. 4) and ANSI/ANS 8.17-2004 (Ref. 5), as described in Section 3.5.3 of Ref. 1. The development of this report is consistent with Test Plan for: Range of Applicability and Bias Determination for Postclosure Criticality. This calculation report has been developed in support of licensing activities for the proposed repository at Yucca Mountain, Nevada, and the results of the calculation may be used in the criticality evaluation for CSNF waste packages based on a conceptual TAD canister.

Radulescu, Georgeta [ORNL; Mueller, Don [ORNL; Goluoglu, Sedat [ORNL; Hollenbach, Daniel F [ORNL; Fox, Patricia B [ORNL

2007-10-01T23:59:59.000Z

283

Nuclear fuel cycle assessment of India: a technical study for U.S.-India cooperation  

E-Print Network [OSTI]

........................................................16 II.G. Current State of Indian Nuclear Program.........................................................17 III INDIAN NUCLEAR FACILITIES .................................................................18 IV FUEL CYCLE... pattern for TBR-1 .........................................100 Fig. 16. India’s proposed nuclear power production strategy........................................103 Fig. 17. Comparison for uranium utilization in electricity generation...

Woddi, Taraknath Venkat Krishna

2008-10-10T23:59:59.000Z

284

International Source Book: Nuclear Fuel Cycle Research and Development Vol 1 Volume 1  

SciTech Connect (OSTI)

This document starts with an overview that summarizes nuclear power policies and waste management activities for nations with significant commercial nuclear fuel cycle activities either under way or planned. A more detailed program summary is then included for each country or international agency conducting nuclear fuel cycle and waste management research and development. This first volume includes the overview and the program summaries of those countries listed alphabetically from Argentina to Italy.

Harmon,, K. M.; Lakey,, L. T.

1983-07-01T23:59:59.000Z

285

Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel  

E-Print Network [OSTI]

and Nuclear Recoil . . . . . . . . . . . . . . . . . . . . .2 Quantitative Measurements using NRF 2.1 Nuclear ResonanceFuture Work A Transmission Nuclear Resonance Fluorescence

Ludewigt, Bernhard A

2011-01-01T23:59:59.000Z

286

Knock limitations of methane-air mixtures in a turbocharged dual-fuel engine  

SciTech Connect (OSTI)

Knock limitations are investigated using natural gas, with diesel pilot ignition, as a fuel for the 3406 DI-TA Caterpillar diesel engine. Thermodynamic properties at TDC are generated by computer and compared with experimental results. Exhaust emissions are analyzed. A comparison is made of dual-fuel operation relative to diesel. Observations are made to determine the onset of knock. The onset of knock is characterized as a function of engine speed, load, inlet manifold temperature, and air-fuel ratio (A/F). The conditions at the inset of knock are determined using cylinder pressure data. The most efficient operating range is determined with knock avoidance as a prime parameter.

Song, Y.K.; Acker, G.H.; Schaetzle, W.J.; Brett, C.E.

1987-01-01T23:59:59.000Z

287

A Critical Step Toward Sustainable Nuclear Fuel Disposal | Department of  

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustion |Energy UsageAUDITVehiclesTankless orA BRIEF HISTORY OF THE DEPARTMENT OF ENERGY AA

288

EA-1977: Acceptance and Disposition of Used Nuclear Fuel Containing  

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector General Office of Audit ServicesMirant PotomacFinal1935:Department of Energy Notice

289

Potential safety-related incidents with possible applicability to a nuclear fuel reprocessing plant  

SciTech Connect (OSTI)

The occurrence of certain potential events in nuclear fuel reprocessing plants could lead to significant consequences involving risk to operating personnel or to the general public. This document is a compilation of such potential initiating events in nuclear fuel reprocessing plants. Possible general incidents and incidents specific to key operations in fuel reprocessing are considered, including possible causes, consequences, and safety features designed to prevent, detect, or mitigate such incidents.

Perkins, W.C.; Durant, W.S.; Dexter, A.H.

1980-12-01T23:59:59.000Z

290

Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations.  

SciTech Connect (OSTI)

Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k{sub eff} of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data.

Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

1999-02-17T23:59:59.000Z

291

THERMODYNAMIC AND KINETIC MODELING OF ADVANCED NUCLEAR FUELS - FINAL LDRD-ER REPORT  

SciTech Connect (OSTI)

This project enhanced our theoretical capabilities geared towards establishing the basic science of a high-throughput protocol for the development of advanced nuclear fuel that should couple modern computational materials modeling and simulation tools, fabrication and characterization capabilities, and targeted high throughput performance testing experiments. The successful conclusion of this ER project allowed us to upgrade state-of-the-art modeling codes, and apply these modeling tools to ab initio energetics and thermodynamic assessments of phase diagrams of various mixtures of actinide alloys, propose a tool for optimizing composition of complex alloys for specific properties, predict diffusion behavior in diffusion couples made of actinide and transition metals, include one new equation in the LLNL phase-field AMPE code, and predict microstructure evolution during alloy coring. In FY11, despite limited funding, the team also initiated an experimental activity, with collaboration from Texas A&M University by preparing samples of nuclear fuels in bulk forms and for diffusion couple studies and metallic matrices, and performing preliminary characterization.

Turchi, P

2011-11-28T23:59:59.000Z

292

Discussing spent nuclear fuel in high school classrooms: addressing public fears through early education  

SciTech Connect (OSTI)

The Inreach program combines the Deep River Science Academy (DRSA) 'learning through research' approach with state of the art communication technology to bring scientific research to high school classrooms. The Inreach program follows the DRSA teaching model where a university student tutor works on a research project with scientific staff at AECL's Chalk River Laboratories. Participating high school classes are located across Canada. The high school students learn about the ongoing research activities via weekly web conferences. In order to engage the students and encourage participation in the conferences, themed exercises linked to the research project are provided to the students. The DRSA's Inreach program uses a cost-effective internet technology to reach a wide audience, in an interactive setting, without anyone leaving their desks or offices. An example Inreach research project is presented here: an investigation of the potential of the Canadian supercritical water cooled reactor (SCWR) concept to burn transuranic elements (Np, Pu, Am, Cm) to reduce the impact of used nuclear fuel. During this project a university student worked with AECL (Atomic Energy of Canada Limited) researchers on technical aspects of the project, and high school students followed their progress and learned about the composition, hazards, and disposition options for used nuclear fuel. Previous projects included the effects of tritium on cellular viability and neutron diffraction measurement of residual stresses in automobile engines.

Winkel, S. [Deep River Science Academy, 20 Forest Ave. P.O. Box 600, Deep River, Ontario K0J 1P0 (Canada); Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada); Sullivan, J.; Jones, S.; Sullivan, K. [Deep River Science Academy, 20 Forest Ave. P.O. Box 600, Deep River, Ontario K0J 1P0 (Canada); Hyland, B.; Pencer, J.; Colton, A. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

2013-07-01T23:59:59.000Z

293

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

294

Spent nuclear fuel characterization for a bounding reference assembly for the receiving basin for off-site fuel  

SciTech Connect (OSTI)

The Basis for Interim Operation (BIO) for the Receiving Basin for Off-Site Fuel (RBOF) facility at the Department of Energy (DOE) Savannah River Site (SRS) nuclear materials production complex, developed in accordance with draft DOE-STD-0019-93, required a hazard categorization for the safety analysis section as outlined in DOE-STD-1027-92. The RBOF facility was thus established as a Category-2 facility (having potential for significant on-site consequences from a radiological release) as defined in DOE 5480.23. Given the wide diversity of spent nuclear fuel stored in the RBOF facility, which made a detailed assessment of the total nuclear inventory virtually impossible, the categorization required a conservative calculation based on the concept of a hypothetical, bounding reference fuel assembly integrated over the total capacity of the facility. This scheme not only was simple but also precluded a potential delay in the completion of the BIO.

Kahook, S.D.; Garrett, R.L.; Canas, L.R.; Beckum, M.J. [Westinghouse Savannah River, Aiken, SC (United States)

1995-07-01T23:59:59.000Z

295

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

296

SULFUR HEXAFLUORIDE TREATMENT OF USED NUCLEAR FUEL TO ENHANCE SEPARATIONS  

SciTech Connect (OSTI)

Reactive Gas Recycling (RGR) technology development has been initiated at Savannah River National Laboratory (SRNL), with a stretch-goal to develop a fully dry recycling technology for Used Nuclear Fuel (UNF). This approach is attractive due to the potential of targeted gas-phase treatment steps to reduce footprint and secondary waste volumes associated with separations relying primarily on traditional technologies, so long as the fluorinators employed in the reaction are recycled for use in the reactors or are optimized for conversion of fluorinator reactant. The developed fluorination via SF{sub 6}, similar to the case for other fluorinators such as NF{sub 3}, can be used to address multiple fuel forms and downstream cycles including continued processing for LWR via fluorination or incorporation into a aqueous process (e.g. modified FLUOREX) or for subsequent pyro treatment to be used in advanced gas reactor designs such metal- or gas-cooled reactors. This report details the most recent experimental results on the reaction of SF{sub 6} with various fission product surrogate materials in the form of oxides and metals, including uranium oxides using a high-temperature DTA apparatus capable of temperatures in excess of 1000{deg}C . The experimental results indicate that the majority of the fission products form stable solid fluorides and sulfides, while a subset of the fission products form volatile fluorides such as molybdenum fluoride and niobium fluoride, as predicted thermodynamically. Additional kinetic analysis has been performed on additional fission products. A key result is the verification that SF{sub 6} requires high temperatures for direct fluorination and subsequent volatilization of uranium oxides to UF{sub 6}, and thus is well positioned as a head-end treatment for other separations technologies, such as the volatilization of uranium oxide by NF{sub 3} as reported by colleagues at PNNL, advanced pyrochemical separations or traditional full recycle approaches. Based on current results of the research at SRNL on SF{sub 6} fluoride volatility for UNF separations, SF{sub 6} treatment renders all anticipated volatile fluorides studied to be volatile, and all non-volatile fluorides studied to be non-volatile, with the notable exception of uranium oxides. This offers an excellent opportunity to use this as a head-end separations treatment process because: 1. SF{sub 6} can be used to remove volatile fluorides from a UNF matrix while leaving behind uranium oxides. Therefore an agent such as NF{sub 3} should be able to very cleanly separate a pure UF{sub 6} stream, leaving compounds in the bottoms such as PuF{sub 4}, SrF{sub 2} and CsF after the UNF matrix has been pre-treated with SF{sub 6}. 2. Due to the fact that the uranium oxide is not separated in the volatilization step upon direct contact with SF{sub 6} at moderately high temperatures (? 1000{deg}C), this fluoride approach may be wellsuited for head-end processing for Gen IV reactor designs where the LWR is treated as a fuel stock, and it is not desired to separate the uranium from plutonium, but it is desired to separate many of the volatile fission products. 3. It is likely that removal of the volatile fission products from the uranium oxide should simplify both traditional and next generation pyroprocessing techniques. 4. SF{sub 6} treatment to remove volatile fission products, with or without treatment with additional fluorinators, could be used to simplify the separations of traditional aqueous processes in similar fashion to the FLUOREX process. Further research should be conducted to determine the separations efficiency of a combined SF{sub 6}/NF{sub 3} separations approach which could be used as a stand-alone separations technology or a head-end process.

Gray, J.; Torres, R.; Korinko, P.; Martinez-Rodriguez, M.; Becnel, J.; Garcia-Diaz, B.; Adams, T.

2012-09-25T23:59:59.000Z

297

Results from NNWSI [Nevada Nuclear Waste Storage Investigations] Series 2 bare fuel dissolution tests  

SciTech Connect (OSTI)

The dissolution and radionuclide release behavior of spent fuel in groundwater is being studied by the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. Two bare spent fuel specimens plus the empty cladding hulls were tested in NNWSI J-13 well water in unsealed fused silica vessels under ambient hot cell air conditions (25{degree}C) in the currently reported tests. One of the specimens was prepared from a rod irradiated in the H. B. Robinson Unit 2 reactor and the other from a rod irradiated in the Turkey Point Unit 3 reactor. Results indicate that most radionuclides of interest fall into three groups for release modeling. The first group principally includes the actinides (U, Np, Pu, Am, and Cm), all of which reached solubility-limited concentrations that were orders of magnitude below those necessary to meet the NRC 10 CFR 60.113 release limits for any realistic water flux predicted for the Yucca Mountain repository site. The second group is nuclides of soluble elements such as Cs, Tc, and I, for which release rates do not appear to be solubility-limited and may depend on the dissolution rate of fuel. In later test cycles, {sup 137}Cs, {sup 90}Sr, {sup 99}Tc, and {sup 129}I were continuously released at rates between about 5 {times} 10{sup {minus}5} and 1 {times} 10{sup {minus}4} of inventory per year. The third group is radionuclides that may be transported in the vapor phase, of which {sup 14}C is of primary concern. Detailed test results are presented and discussed. 17 refs., 15 figs., 21 tabs.

Wilson, C.N.

1990-09-01T23:59:59.000Z

298

Nuclear Fuel Cycle Options Evaluation to Inform R&D Planning  

SciTech Connect (OSTI)

An Evaluation and Screening (E&S) of nuclear fuel cycle options has been conducted in fulfilment of a Charter specified for the study by the U.S. Department of Energy (DOE) Office of Nuclear Energy. The E&S study used an objective and independently reviewed evaluation process to provide information about the potential benefits and challenges that could strengthen the basis and provide guidance for the research and development(R&D) activities undertaken by the DOE Fuel Cycle Technologies Program Office. Using the nine evaluation criteria specified in the Charter and associated evaluation metrics and processes developed during the E&S study, a screening was conducted of 40 nuclear fuel cycle evaluation groups to provide answers to the questions: (1) Which nuclear fuel cycle system options have the potential for substantial beneficial improvements in nuclear fuel cycle performance, and what aspects of the options make these improvements possible? (2)Which nuclear material management approaches can favorably impact the performance of fuel cycle options? (3)Where would R&D investment be needed to support the set of promising fuel cycle system options and nuclear material management approaches identified above, and what are the technical objectives of associated technologies?

R. Wigeland; T. Taiwo; M. Todosow; H. Ludewig; W. Halsey; J. Gehin; R. Jubin; J. Buelt; S. Stockinger; K. Jenni; B. Oakley

2014-04-01T23:59:59.000Z

299

Identification and Analysis of Critical Gaps in Nuclear Fuel Cycle Codes Required by the SINEMA Program  

SciTech Connect (OSTI)

The current state of the art in nuclear fuel cycle (NFC) modeling is an eclectic mixture of codes with various levels of applicability, flexibility, and availability. In support of the advanced fuel cycle systems analyses, especially those by the Advanced Fuel Cycle Initiative (AFCI), Unviery of Cincinnati in collaboration with Idaho State University carried out a detailed review of the existing codes describing various aspects of the nuclear fuel cycle and identified the research and development needs required for a comprehensive model of the global nuclear energy infrastructure and the associated nuclear fuel cycles. Relevant information obtained on the NFC codes was compiled into a relational database that allows easy access to various codes' properties. Additionally, the research analyzed the gaps in the NFC computer codes with respect to their potential integration into programs that perform comprehensive NFC analysis.

Adrian Miron; Joshua Valentine; John Christenson; Majd Hawwari; Santosh Bhatt; Mary Lou Dunzik-Gougar: Michael Lineberry

2009-10-01T23:59:59.000Z

300

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

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301

Redundancy of Supply in the International Nuclear Fuel Fabrication Market: Are Fabrication Services Assured?  

SciTech Connect (OSTI)

For several years, Pacific Northwest National Laboratory (PNNL) has been assessing the reliability of nuclear fuel supply in support of the U.S. Department of Energy/National Nuclear Security Administration. Three international low enriched uranium reserves, which are intended back up the existing and well-functioning nuclear fuel market, are currently moving toward implementation. These backup reserves are intended to provide countries credible assurance that of the uninterrupted supply of nuclear fuel to operate their nuclear power reactors in the event that their primary fuel supply is disrupted, whether for political or other reasons. The efficacy of these backup reserves, however, may be constrained without redundant fabrication services. This report presents the findings of a recent PNNL study that simulated outages of varying durations at specific nuclear fuel fabrication plants. The modeling specifically enabled prediction and visualization of the reactors affected and the degree of fuel delivery delay. The results thus provide insight on the extent of vulnerability to nuclear fuel supply disruption at the level of individual fabrication plants, reactors, and countries. The simulation studies demonstrate that, when a reasonable set of qualification criteria are applied, existing fabrication plants are technically qualified to provide backup fabrication services to the majority of the world's power reactors. The report concludes with an assessment of the redundancy of fuel supply in the nuclear fuel market, and a description of potential extra-market mechanisms to enhance the security of fuel supply in cases where it may be warranted. This report is an assessment of the ability of the existing market to respond to supply disruptions that occur for technical reasons. A forthcoming report will address political disruption scenarios.

Seward, Amy M.; Toomey, Christopher; Ford, Benjamin E.; Wood, Thomas W.; Perkins, Casey J.

2011-11-14T23:59:59.000Z

302

AIR SHIPMENT OF SPENT NUCLEAR FUEL FROM THE BUDAPEST RESEARCH REACTOR  

SciTech Connect (OSTI)

The shipment of spent nuclear fuel is usually done by a combination of rail, road or sea, as the high activity of the SNF needs heavy shielding. Air shipment has advantages, e.g. it is much faster than any other shipment and therefore minimizes the transit time as well as attention of the public. Up to now only very few and very special SNF shipments were done by air, as the available container (TUK6) had a very limited capacity. Recently Sosny developed a Type C overpack, the TUK-145/C, compliant with IAEA Standard TS-R-1 for the VPVR/M type Skoda container. The TUK-145/C was first used in Vietnam in July 2013 for a single cask. In October and November 2013 a total of six casks were successfully shipped from Hungary in three air shipments using the TUK-145/C. The present paper describes the details of these shipments and formulates the lessons learned.

Dewes, J.

2014-02-24T23:59:59.000Z

303

Radiation Imaging of Dry-Storage Casks for Spent Nuclear Fuel  

SciTech Connect (OSTI)

The authors report the results of a measurement campaign conducted on six dry-storage, spent-nuclear-fuel storage casks at the Idaho National Laboratory. a gamma-ray imager, a thermal-neutron imager and a Ge-spectrometer were used to collect data. The campaign was conducted to examine the feasibility of using cask radiation signatures as unique identifiers for individual casks as part of a safeguards regime. The results clearly show different morphologies for the various cask types although the signatures are deemed insufficient to uniquely identify individual casks of the same type. Based on results with the Ge-spectrometer and differences between thermal neutron images and neutron-dose meters, this result is attributed to the limitations of the extant imagers used, rather than of the basic concept.

Ziock, K; Caffrey, G; Lebrun, A; Forman, L; Vanier, P; Wharton, J

2005-11-08T23:59:59.000Z

304

Ukraine Fuel Removal: Fact Sheet | National Nuclear Security Administration  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLasDelivered energy consumption by sectorlong version) The0 -ITER's centralNewUSUS0,Ukraine Fuel

305

EIS-0279: Spent Nuclear Fuel Management, Aiken, South Carolina | Department  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-UpHeat Pump Models |Conduct,Final9:DepartmentExtensionRecordRecord of Decision

306

Review of Used Nuclear Fuel Storage and Transportation Technical Gap  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustion | Department ofT ib l LPROJECTS IN RENEWABLEOperated inFebruary 26, 2009atEnergy

307

Review of Used Nuclear Fuel Storage and Transportation Technical Gap  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustion | Department ofT ib l LPROJECTS IN RENEWABLEOperated inFebruary 26, 2009atEnergyAnalysis |

308

Used Nuclear Fuel Loading and Structural Performance Under Normal  

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector GeneralDepartment of EnergyofProject is on Track|SolarDepartment of Energy Use ofPlan

309

Used Nuclear Fuel Loading and Structural Performance Under Normal  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustion |Energy Usage »of EnergyTheTwo New12.'6/0.2 ......Uranium LeaseThroughAugust

310

Management of Salt Waste from Electrochemical Processing of Used Nuclear Fuel  

SciTech Connect (OSTI)

Electrochemical processing of used nuclear fuel involves operation of one or more cells containing molten salt electrolyte. Processing of the fuel results in contamination of the salt via accumulation of fission products and transuranic (TRU) actinides. Upon reaching contamination limits, the salt must be removed and either disposed or treated to remove the contaminants and recycled back to the process. During development of the Experimental Breeder Reactor-II spent fuel treatment process, waste salt from the electrorefiner was to be stabilized in a ceramic waste form and disposed of in a high-level waste repository. With the cancellation of the Yucca Mountain high-level waste repository, other options are now being considered. One approach that involves direct disposal of the salt in a geologic salt formation has been evaluated. While waste forms such as the ceramic provide near-term resistance to corrosion, they may not be necessary to ensure adequate performance of the repository. To improve the feasibility of direct disposal, recycling a substantial fraction of the useful salt back to the process equipment could minimize the volume of the waste. Experiments have been run in which a cold finger is used for this purpose to crystallize LiCl from LiCl/CsCl. If it is found to be unsuitable for transportation, the salt waste could also be immobilized in zeolite without conversion to the ceramic waste form.

Michael F. Simpson; Michael N. Patterson; Joon Lee; Yifeng Wang; Joshua Versey; Ammon Williams; Supathorn Phongikaroon; James Allensworth; Man-Sung Yim

2013-10-01T23:59:59.000Z

311

Management of salt waste from electrochemical processing of used nuclear fuel  

SciTech Connect (OSTI)

Electrochemical processing of used nuclear fuel involves operation of one or more cells containing molten salt electrolyte. Processing of the fuel results in contamination of the salt via accumulation of fission products and transuranic (TRU) actinides. Upon reaching contamination limits, the salt must be removed and either disposed or treated to remove the contaminants and recycled back to the process. During development of the Experimental Breeder Reactor-II spent fuel treatment process, waste salt from the electro-refiner was to be stabilized in a ceramic waste form and disposed of in a high-level waste repository. With the cancellation of the Yucca Mountain high-level waste repository, other options are now being considered. One approach that involves direct disposal of the salt in a geologic salt formation has been evaluated. While waste forms such as the ceramic provide near-term resistance to corrosion, they may not be necessary to ensure adequate performance of the repository. To improve the feasibility of direct disposal, recycling a substantial fraction of the useful salt back to the process equipment could minimize the volume of the waste. Experiments have been run in which a cold finger is used for this purpose to crystallize LiCl from LiCl/CsCl. If it is found to be unsuitable for transportation, the salt waste could also be immobilized in zeolite without conversion to the ceramic waste form. (authors)

Simpson, M.F.; Patterson, M.N. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, Idaho 83415 (United States); Lee, J.; Wang, Y. [Sandia National Laboratory, Albuquerque, NM (United States); Versey, J.; Phongikaroon, S. [University of Idaho, Idaho Falls, ID (United States)

2013-07-01T23:59:59.000Z

312

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

313

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

314

Thermal Analysis of Surrogate Simulated Molten Salts with Metal Chloride Impurities for Electrorefining Used Nuclear Fuel  

SciTech Connect (OSTI)

This project is a fundamental study to measure thermal properties (liquidus, solidus, phase transformation, and enthalpy) of molten salt systems of interest to electrorefining operations, which are used in both the fuel cycle research & development mission and the spent fuel treatment mission of the Department of Energy. During electrorefining operations the electrolyte accumulates elements more active than uranium (transuranics, fission products and bond sodium). The accumulation needs to be closely monitored because the thermal properties of the electrolyte will change as the concentration of the impurities increases. During electrorefining (processing techniques used at the Idaho National Laboratory to separate uranium from spent nuclear fuel) it is important for the electrolyte to remain in a homogeneous liquid phase for operational safeguard and criticality reasons. The phase stability of molten salts in an electrorefiner may be adversely affected by the buildup of fission products in the electrolyte. Potential situations that need to be avoided are: (i) build up of fissile elements in the salt approaching the criticality limits specified for the vessel (ii) freezing of the salts due to change in the liquidus temperature and (iii) phase separation (non-homogenous solution) of elements. The stability (and homogeneity) of the phases can potentially be monitored through the thermal characterization of the salts, which can be a function of impurity concentration. This work describes the experimental results of typical salts compositions, consisting of chlorides of strontium, samarium, praseodymium, lanthanum, barium, cerium, cesium, neodymium, sodium and gadolinium (as a surrogate for both uranium and plutonium), used in the processing of used nuclear fuels. Differential scanning calorimetry was used to analyze numerous salt samples providing results on the thermal properties. The property of most interest to pyroprocessing is the liquidus temperature. It was previously known the liquidus temperature of the molten salt would change as spent fuel is processed through the Mk-IV electrorefiner. However, the extent of the increase in liquidus temperature was not known. This work is first of its kind in determining thermodynamic properties of a molten salt electrolyte containing transuranics, fission products and bond sodium. Experimental data concluded that the melting temperature of the electrolyte will become greater than the operating temperature of the Mk-IV ER during current fuel processing campaigns. Collected data also helps predict when the molten salt electrolyte will no longer be able to support electrorefining operations.

Toni Y. Gutknecht; Guy L. Fredrickson; Vivek Utgikar

2012-04-01T23:59:59.000Z

315

Final Environmental Impact Statement for the Treatment and Management of Sodium-Bonded Spent Nuclear Fuel  

SciTech Connect (OSTI)

DOE is responsible for the safe and efficient management of its sodium-bonded spent nuclear fuel. This fuel contains metallic sodium, a highly reactive material; metallic uranium, which is also reactive; and in some cases, highly enriched uranium. The presence of reactive materials could complicate the process of qualifying and licensing DOE's sodium-bonded spent nuclear fuel inventory for disposal in a geologic repository. Currently, more than 98 percent of this inventory is located at the Idaho National Engineering and Environmental Laboratory (INEEL), near Idaho Falls, Idaho. In addition, in a 1995 agreement with the State of Idaho, DOE committed to remove all spent nuclear fuel from Idaho by 2035. This EIS evaluates the potential environmental impacts associated with the treatment and management of sodium-bonded spent nuclear fuel in one or more facilities located at Argonne National Laboratory-West (ANL-W) at INEEL and either the F-Canyon or Building 105-L at the Savannah River Site (SRS) near Aiken, South Carolina. DOE has identified and assessed six proposed action alternatives in this EIS. These are: (1) electrometallurgical treatment of all fuel at ANL-W, (2) direct disposal of blanket fuel in high-integrity cans with the sodium removed at ANL-W, (3) plutonium-uranium extraction (PUREX) processing of blanket fuel at SRS, (4) melt and dilute processing of blanket fuel at ANL-W, (5) melt and dilute processing of blanket fuel at SRS, and (6) melt and dilute processing of all fuel at ANL-W. In addition, Alternatives 2 through 5 include the electrometallurgical treatment of driver fuel at ANL-W. Under the No Action Alternative, the EIS evaluates both the continued storage of sodium-bonded spent nuclear fuel until the development of a new treatment technology or direct disposal without treatment. Under all of the alternatives, the affected environment is primarily within 80 kilometers (50 miles) of spent nuclear fuel treatment facilities. Analyses indicate little difference in the environmental impacts among alternatives. DOE has identified electrometallurgical treatment as its Preferred Alternative for the treatment and management of all sodium-bonded spent nuclear fuel, except for the Fermi-1 blanket fuel. The No Action Alternative is preferred for the Fermi-1 blanket spent nuclear fuel.

N /A

2000-08-04T23:59:59.000Z

316

Preliminary Evaluation of Removing Used Nuclear Fuel From Nine Shutdown Sites  

SciTech Connect (OSTI)

The Blue Ribbon Commission on America’s Nuclear Future 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. In this report, a preliminary evaluation of removing used nuclear fuel from nine shutdown sites was conducted. The shutdown sites included Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion. At these sites a total of 7649 used nuclear fuel assemblies and a total of 2813.2 metric tons heavy metal (MTHM) of used nuclear fuel are contained in 248 storage canisters. In addition, 11 canisters containing greater-than-Class C (GTCC) low-level radioactive waste are stored at these sites. The evaluation was divided in four components: • characterization of the used nuclear fuel and GTCC low-level radioactive waste inventory at the shutdown sites • an evaluation of the onsite transportation conditions at the shutdown sites • an evaluation of the near-site transportation infrastructure and experience relevant to the shipping of transportation casks containing used nuclear fuel from the shutdown sites • an evaluation of the actions necessary to prepare for and remove used nuclear fuel and GTCC low-level radioactive waste from the shutdown sites. Using these evaluations the authors developed time sequences of activities and time durations for removing the used nuclear fuel and GTCC low-level radioactive waste from a single shutdown site, from three shutdown sites located close to each other, and from all nine shutdown sites.

Maheras, Steven J. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Best, Ralph [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Ross, Steven B. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Buxton, Kenneth A. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); England, Jeffery L. [Savannah River National Laboratory, Aiken, SC (United States); McConnell, Paul [Sandia National Laboratories, Albuquerque, NM (United States)

2013-04-30T23:59:59.000Z

317

Deployment evaluation methodology for the electrometallurgical treatment of DOE-EM spent nuclear fuel  

SciTech Connect (OSTI)

Part of the Department of Energy (DOE) spent nuclear fuel (SNF) inventory may require some type of treatment to meet acceptance criteria at various disposition sites. The current focus for much of this spent nuclear fuel is the electrometallurgical treatment process under development at Argonne National Laboratory. Potential flowsheets for this treatment process are presented. Deployment of the process for the treatment of the spent nuclear fuel requires evaluation to determine the spent nuclear fuel program need for treatment and compatibility of the spent nuclear fuel with the process. The evaluation of need includes considerations of cost, technical feasibility, process material disposition, and schedule to treat a proposed fuel. A siting evaluation methodology has been developed to account for these variables. A work breakdown structure is proposed to gather life-cycle cost information to allow evaluation of alternative siting strategies on a similar basis. The evaluation methodology, while created specifically for the electrometallurgical evaluation, has been written such that it could be applied to any potential treatment process that is a disposition option for spent nuclear fuel. Future work to complete the evaluation of the process for electrometallurgical treatment is discussed.

Dahl, C.A.; Adams, J.P.; Ramer, R.J.

1998-07-01T23:59:59.000Z

318

Nuclear Fuel Storage and Transportation Planning Project Overview |  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Careerlumens_placard-green.epsEnergy Second Quarter4, 2014 Dr.7446AugustJuneElectricityFacility

319

Nuclear Regulatory Commission's Integrated Strategy for Spent Fuel Management  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGY TAXBalanced ScorecardReactor Technology Subcommittee of NEAC Mujid KazimiNRC's

320

Preliminary Evaluation of Removing Used Nuclear Fuel From Shutdown Sites |  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNG |September 15, 2010Energy6DepartmentOutages

Note: This page contains sample records for the topic "nuclear fuels limited" 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

Pilot Application to Nuclear Fuel Cycle Options | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Careerlumens_placard-green.epsEnergy1.pdfMarket | DepartmentPhotoelectrochemicalInan<aPilgrimA

322

Nuclear Fuels Storage & Transportation Planning Project | Department of  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGYWomen OwnedofDepartment ofJaredOakscience-based, applied engineering nationalEnergy

323

Behavior of Spent Nuclear Fuel in Water Pool Storage  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth (AOD)ProductssondeadjustsondeadjustAboutScience ProgramBackground8.0.1Vulture SpatialBECOMEBehaviorBehavior of

324

Used Nuclear Fuels Storage, Transportation, and Disposal Analysis Resource  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLasDelivered energy consumption by sectorlongUpdates by DianeDemographics UsageUsage by Job

325

Energy Department Announces New Investment in Nuclear Fuel Storage Research  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Career Scientists'Montana.Program -Department of EnergyStakeholdersEnergyDevelop Clean Energy|

326

Fuel Cycle Science & Technology | Nuclear Science | ORNL  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation ProposedUsing ZirconiaPolicyFeasibilityFieldMinds"OfficeTourFrom3, 2015authors Advanced

327

Integrated Used Nuclear Fuel Storage, Transportation, and Disposal Canister  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation ProposedUsingFunInfrared Land SurfaceVirus-Infected Macaques throughBiomass IntegratedSystem -

328

Sizing up nuclear fuel | Y-12 National Security Complex  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLas ConchasPassiveSubmitted forHighlightsSeminarsSiliconSite Map TUNL pdf'sFundSizeSizing a

329

Nuclear Regulatory Commission's Integrated Strategy for Spent Fuel  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNG |September 15, 2010 PrintingNeed| Department ofDC.Navy UnitedManagement |

330

Advanced Nuclear Fuel | Y-12 National Security Complex  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth (AOD)ProductssondeadjustsondeadjustAbout the BuildingInnovation Portal AdvancedMethods forDevelopment

331

Spent Fuel Analyses for Nuclear Safeguards | ornl.gov  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLas ConchasPassiveSubmittedStatus Tom Fletcher,Future |CarlosSpeakersSpectroscopySpencer R.Spent

332

Constraints, limits and extensions for nuclear energy functionals  

E-Print Network [OSTI]

In the present contribution, we discuss the behavior of Skyrme forces when they are employed to study both neutron stars and giant resonance states in 208Pb within the fully self-consistent Random Phase Approximation (RPA). We point out that clear correlations exist between the results for the isoscalar monopole and isovector dipole resonances (ISGMR and IVGDR), and definite quantities which characterize the equation of state (EOS) of uniform matter. We propose that the RPA results or, to some extent, the mentioned EOS parameters, are used as constraints when a force is fitted. This suggestion can be valid also when the fit of a more general energy density functional is envisaged. We use our considerations to select a limited number of Skyrme forces (10) out of a large sample of 78 interactions.

G. Colo'

2009-02-21T23:59:59.000Z

333

GAS PUFF FUELED H-MODE DISCHARGES WITH GOOD ENERGY CONFINEMENT ABOVE THE GREENWALD DENSITY LIMIT ON DIII-D  

SciTech Connect (OSTI)

A261. GAS PUFF FUELED H-MODE DISCHARGES WITH GOOD ENERGY CONFINEMENT ABOVE THE GREENWALD DENSITY LIMIT ON DIII-D

T.H. OSBORNE; A.W. LEONARD; M.A. MAHDAVI; M.S. CHU; M.E. FENSTERMACHER; R.J. LA HAYE; G.R. McKEE; T.W. PETRIE; E.J. DOYLE; G.M. STAEBLER; M.R. WADE

2000-11-01T23:59:59.000Z

334

An improved structural mechanics model for the FRAPCON nuclear fuel performance code  

E-Print Network [OSTI]

In order to provide improved predictions of Pellet Cladding Mechanical Interaction (PCMI) for the FRAPCON nuclear fuel performance code, a new model, the FRAPCON Radial-Axial Soft Pellet (FRASP) model, was developed. This ...

Mieloszyk, Alexander James

2012-01-01T23:59:59.000Z

335

Sensitivity analysis and optimization of the nuclear fuel cycle : a systematic approach  

E-Print Network [OSTI]

For decades, nuclear energy development was based on the expectation that recycling of the fissionable materials in the used fuel from today's light water reactors into advanced (fast) reactors would be implemented as soon ...

Passerini, Stefano

2012-01-01T23:59:59.000Z

336

Concrete Shield Performance of the VSC-17 Spent Nuclear Fuel Cask  

SciTech Connect (OSTI)

The VSC-17 Spent Nuclear Fuel Storage Cask was surveyed for degradation of the concrete shield by radiation measurement, temperature measurement, and ultrasonic testing. No general loss of shielding function was identified.

Koji Shirai

2006-04-01T23:59:59.000Z

337

EA-1117: Management of Spent Nuclear Fuel on the Oak Ridge Reservation, Oak Ridge, Tennessee  

Broader source: Energy.gov [DOE]

This EA evaluates the environmental impacts of the proposal for the management of spent nuclear fuel on the U.S. Department of Energy's Oak Ridge Reservation to implement the preferred alternative...

338

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

339

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

340

Environmental Aspects of Advanced Nuclear Fuel Cycles: Parametric Modeling and Preliminary Analysis  

E-Print Network [OSTI]

Nuclear power has the potential to help reduce rising carbon emissions, but to be considered sustainable, it must also demonstrate the availability of an indefinite fuel supply as well as not produce any significant negative environmental effects...

Yancey, Kristina D.

2010-07-14T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear fuels limited" 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

Development of the fundamental attributes and inputs for proliferation resistance assessments of nuclear fuel cycles  

E-Print Network [OSTI]

Robust and reliable quantitative proliferation resistance assessment tools are critical to a strengthened nonproliferation regime and to the future deployment of nuclear fuel cycle technologies. Efforts to quantify proliferation resistance have thus...

Giannangeli, Donald D. J., III

2007-09-17T23:59:59.000Z

342

Development of Gd-Enriched Alloys for Spent Nuclear Fuel Applications--Part 1: Preliminary Characterization  

E-Print Network [OSTI]

of Gd-containing alloys for storage, transport, and disposal of spent nuclear fuel. However, unlike, the basket materials must be corrosion resistant under the projected stor- age conditions. Recent research

DuPont, John N.

343

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

344

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

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGY TAX POLICIES7.pdfFuel Cell Vehicle Basics Fuel Cell Vehicle Basics AugustSiCNEAC

345

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

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGY TAX POLICIES7.pdfFuel Cell Vehicle Basics Fuel Cell Vehicle Basics

346

Application of ALARA principles to shipment of spent nuclear fuel  

SciTech Connect (OSTI)

The public exposure from spent fuel shipment is very low. In view of this low exposure and the perfect safety record for spent fuel shipment, existing systems can be considered satisfactory. On the other hand, occupational exposure reduction merits consideration and technology improvement to decrease dose should concentrate on this exposure. Practices that affect the age of spent fuel in shipment and the number of times the fuel must be shipped prior to disposal have the largest impact. A policy to encourage a 5-year spent fuel cooling period prior to shipment coupled with appropriate cask redesign to accommodate larger loads would be consistent with ALARA and economic principles. And finally, bypassing high population density areas will not in general reduce shipment dose.

Greenborg, J.; Brackenbush, L.W.; Murphy, D.W. Burnett, R.A.; Lewis, J.R.

1980-05-01T23:59:59.000Z

347

Mass-spectrometric determination of americium and curium in spent nuclear fuel  

SciTech Connect (OSTI)

It was shown that it is possible to carry out isotopic analysis of americium and curium isolated in one fraction from a sample of nuclear fuel without chemical separation of the elements. This became possible as a result of the use in the mass-spectrometric measurements of a highly efficient ion source with surface ionization in a partially closed cavity. With this ion source, the content of americium and curium in nuclear fuel was determined by isotope dilution.

Kalygin, V.V.; Gabeskiriya, V.Ya.

1987-01-01T23:59:59.000Z

348

Potential opportunities for nano materials to help enable enhanced nuclear fuel performance  

SciTech Connect (OSTI)

This presentation is an overview of the technical challenges for development of nuclear fuels with enhanced performance and accident tolerance. Key specific aspects of improved fuel performance are noted. Examples of existing nanonuclear projects and concepts are presented and areas of potential focus are suggested. The audience for this presentation includes representatives from: DOE-NE, other national laboratories, industry and academia. This audience is a mixture of nanotechnology experts and nuclear energy researchers and managers.

McClellan, Kenneth J. [Los Alamos National Laboratory

2012-06-06T23:59:59.000Z

349

Method for forming nuclear fuel containers of a composite construction and the product thereof  

DOE Patents [OSTI]

An improved method for producing nuclear fuel containers of a composite construction having components providing therein a barrier system for resisting destructive action by volatile fission products or impurities and also interdiffusion of metal constituents, and the product thereof. The composite nuclear fuel containers of the method comprise a casing of zirconium or alloy thereof with a layer of copper overlying an oxidized surface portion of the zirconium or alloy thereof.

Cheng, Bo-Ching (Fremont, CA); Rosenbaum, Herman S. (Fremont, CA); Armijo, Joseph S. (Saratoga, CA)

1984-01-01T23:59:59.000Z

350

Performance of solid oxide fuel cells approaching the two-dimensional limit  

SciTech Connect (OSTI)

We model electrochemical kinetics and physical conduction mechanisms for carrier transport in electrolyte membranes to determine the limits of dimensionality reduction (down to 1?nm) on maximum power output of solid oxide fuel cells with symmetric Pt electrodes. Using Y-doped ZrO{sub 2}, we find a minimum thickness of ?6?nm to realize near ideal chemical potential in such fuel cells, which is limited by electronic breakdown when approaching the dielectric breakdown strength. For larger electrolyte thicknesses, the greatest source of electronic leakage influencing power loss is from Ohmic transport of minority carriers and emission of trapped carriers. For porous metallic electrodes, an ideal microstructure with the particle size comparable to particle spacing dimensions is found to accurately model experimental results. The role of electronic trap states in the electrolyte band gap on power density characteristics is highlighted.

Kerman, K., E-mail: kkerman@fas.harvard.edu; Ramanathan, S. [School of Engineering and Applied Sciences, Harvard University, Cambridge, Massachusetts 02138 (United States)

2014-05-07T23:59:59.000Z

351

Analyzing Nuclear Fuel Cycles from Isotopic Ratios of Waste Products Applicable to Measurement by Accelerator Mass Spectrometry  

SciTech Connect (OSTI)

An extensive study was conducted to determine isotopic ratios of nuclides in spent fuel that may be utilized to reveal historical characteristics of a nuclear reactor cycle. This forensic information is important to determine the origin of unknown nuclear waste. The distribution of isotopes in waste products provides information about a nuclear fuel cycle, even when the isotopes of uranium and plutonium are removed through chemical processing. Several different reactor cycles of the PWR, BWR, CANDU, and LMFBR were simulated for this work with the ORIGEN-ARP and ORIGEN 2.2 codes. The spent fuel nuclide concentrations of these reactors were analyzed to find the most informative isotopic ratios indicative of irradiation cycle length and reactor design. Special focus was given to long-lived and stable fission products that would be present many years after their creation. For such nuclides, mass spectrometry analysis methods often have better detection limits than classic gamma-ray spectroscopy. The isotopic ratios {sup 151}Sm/{sup 146}Sm, {sup 149}Sm/{sup 146}Sm, and {sup 244}Cm/{sup 246}Cm were found to be good indicators of fuel cycle length and are well suited for analysis by accelerator mass spectroscopy.

Biegalski, S R; Whitney, S M; Buchholz, B

2005-08-24T23:59:59.000Z

352

A Spouted Bed Reactor Monitoring System for Particulate Nuclear Fuel  

SciTech Connect (OSTI)

Conversion and coating of particle nuclear fuel is performed in spouted (fluidized) bed reactors. The reactor must be capable of operating at temperatures up to 2000°C in inert, flammable, and coating gas environments. The spouted bed reactor geometry is defined by a graphite retort with a 2.5 inch inside diameter, conical section with a 60° included angle, and a 4 mm gas inlet orifice diameter through which particles are removed from the reactor at the completion of each run. The particles may range from 200 µm to 2 mm in diameter. Maintaining optimal gas flow rates slightly above the minimum spouting velocity throughout the duration of each run is complicated by the variation of particle size and density as conversion and/or coating reactions proceed in addition to gas composition and temperature variations. In order to achieve uniform particle coating, prevent agglomeration of the particle bed, and monitor the reaction progress, a spouted bed monitoring system was developed. The monitoring system includes a high-sensitivity, low-response time differential pressure transducer paired with a signal processing, data acquisition, and process control unit which allows for real-time monitoring and control of the spouted bed reactor. The pressure transducer is mounted upstream of the spouted bed reactor gas inlet. The gas flow into the reactor induces motion of the particles in the bed and prevents the particles from draining from the reactor due to gravitational forces. Pressure fluctuations in the gas inlet stream are generated as the particles in the bed interact with the entering gas stream. The pressure fluctuations are produced by bulk movement of the bed, generation and movement of gas bubbles through the bed, and the individual motion of particles and particle subsets in the bed. The pressure fluctuations propagate upstream to the pressure transducer where they can be monitored. Pressure fluctuation, mean differential pressure, gas flow rate, reactor operating temperature data from the spouted bed monitoring system are used to determine the bed operating regime and monitor the particle characteristics. Tests have been conducted to determine the sensitivity of the monitoring system to the different operating regimes of the spouted particle bed. The pressure transducer signal response was monitored over a range of particle sizes and gas flow rates while holding bed height constant. During initial testing, the bed monitoring system successfully identified the spouting regime as well as when particles became interlocked and spouting ceased. The particle characterization capabilities of the bed monitoring system are currently being tested and refined. A feedback control module for the bed monitoring system is currently under development. The feedback control module will correlate changes in the bed response to changes in the particle characteristics and bed spouting regime resulting from the coating and/or conversion process. The feedback control module will then adjust the gas composition, gas flow rate, and run duration accordingly to maintain the bed in the desired spouting regime and produce optimally coated/converted particles.

D. S. Wendt; R. L. Bewley; W. E. Windes

2007-06-01T23:59:59.000Z

353

Measures of the environmental footprint of the front end of the nuclear fuel cycle  

SciTech Connect (OSTI)

Previous estimates of environmental impacts associated with the front end of the nuclear fuel cycle (FEFC) have focused primarily on energy consumption and CO2 emissions. Results have varied widely. This work builds upon reports from operating facilities and other primary data sources to build a database of front end environmental impacts. This work also addresses land transformation and water withdrawals associated with the processes of the FEFC. These processes include uranium extraction, conversion, enrichment, fuel fabrication, depleted uranium disposition, and transportation. To allow summing the impacts across processes, all impacts were normalized per tonne of natural uranium mined as well as per MWh(e) of electricity produced, a more conventional unit for measuring environmental impacts that facilitates comparison with other studies. This conversion was based on mass balances and process efficiencies associated with the current once-through LWR fuel cycle. Total energy input is calculated at 8.7 x 10- 3 GJ(e)/MWh(e) of electricity and 5.9 x 10- 3 GJ(t)/MWh(e) of thermal energy. It is dominated by the energy required for uranium extraction, conversion to fluoride compound for subsequent enrichment, and enrichment. An estimate of the carbon footprint is made from the direct energy consumption at 1.7 kg CO2/MWh(e). Water use is likewise dominated by requirements of uranium extraction, totaling 154 L/MWh(e). Land use is calculated at 8 x 10- 3 m2/MWh(e), over 90% of which is due to uranium extraction. Quantified impacts are limited to those resulting from activities performed within the FEFC process facilities (i.e. within the plant gates). Energy embodied in material inputs such as process chemicals and fuel cladding is identified but not explicitly quantified in this study. Inclusion of indirect energy associated with embodied energy as well as construction and decommissioning of facilities could increase the FEFC energy intensity estimate by a factor of up to 2.

E. Schneider; B. Carlsen; E. Tavrides; C. van der Hoeven; U. Phathanapirom

2013-11-01T23:59:59.000Z

354

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

355

Nuclear tanker producing liquid fuels from air and water  

E-Print Network [OSTI]

Emerging technologies in CO? air capture, high temperature electrolysis, microchannel catalytic conversion, and Generation IV reactor plant systems have the potential to create a shipboard liquid fuel production system ...

Galle-Bishop, John Michael

2011-01-01T23:59:59.000Z

356

Study of feasible and sustainable multilateral approach on nuclear fuel cycle  

SciTech Connect (OSTI)

Despite the Fukushima accident it is undeniable that nuclear power remains one of the most important methods to handle global growth of economic/energy consumption and issues with greenhouse gases. If the demand for nuclear power increases, the demand for not only the generation of power but also for refining uranium (U), conversion, enrichment, re-conversion, and fuel manufacturing should increase. In addition, concerns for the proliferation of 'Sensitive Nuclear Technologies' (SNT) should also increase. We propose a demand-side approach, where nuclear fuel cycle (NFC) activities would be implemented among multiple states. With this approach, NFC services, in particular those using SNTs, are multilaterally executed and controlled, thereby preventing unnecessary proliferation of SNTs, and enabling safe and appropriate control of nuclear technologies and nuclear materials. This proposal would implement nuclear safety and security at an international level and solve transport issues for nuclear fuels. This proposal is based on 3 types of cooperation for each element of NFC: type A: cooperation for 3S only, services received; Type B: cooperation for 3S, MNA (Multilateral Nuclear Activities) without transfer of ownership to MNA; and Type C cooperation for 3S, MNA holding ownership rights. States involved in the 3 types of activity should be referred to as partner states, host states, and site states respectively. The feasibility of the proposal is discussed for the Asian region.

Kuno, Y.; Tazaki, M. [University of Tokyo, Tokyo (Japan); Japan Atomic Energy Agency - JAEA, 4-49 Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319-1184 (Japan); Akiba, M.; Takashima, R.; Izumi, Y.; Tanaka, S. [University of Tokyo, Tokyo (Japan)

2013-07-01T23:59:59.000Z

357

THE ATTRACTIVENESS OF MATERIAS ASSOCIATED WITH THORIUM-BASED NUCLEAR FUEL CYCLES FOR PHWRS  

SciTech Connect (OSTI)

This paper reports the continued evaluation of the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with thorium based nuclear fuel cycles. Specifically, this paper examines a thorium fuel cycle in which a pressurized heavy water reactor (PHWR) is fueled with mixtures of natural uranium/233U/thorium. This paper uses a PHWR fueled with natural uranium as a base fuel cycle, and then compares material attractiveness of fuel cycles that use 233U/thorium salted with natural uranium. The results include the material attractiveness of fuel at beginning of life (BoL), end of life (EoL), and the number of fuel assemblies required to collect a bare critical mass of plutonium or uranium. This study indicates what is required to render the uranium as having low utility for use in nuclear weapons; in addition, this study estimates the increased number of assemblies required to accumulate a bare critical mass of plutonium that has a higher utility for use in nuclear weapons. This approach identifies that some fuel cycles may be easier to implement the International Atomic Energy Agency (IAEA) safeguards approach and have a more effective safeguards by design outcome. For this study, approximately one year of fuel is required to be reprocessed to obtain one bare critical mass of plutonium. Nevertheless, the result of this paper suggests that all spent fuel needs to be rigorously safeguarded and provided with high levels of physical protection. This study was performed at the request of the United States Department of Energy /National Nuclear Security Administration (DOE/NNSA). The methodology and key findings will be presented.

Prichard, Andrew W.; Niehus, Mark T.; Collins, Brian A.; Bathke, Charles G.; Ebbinghaus, Bartley B.; Hase, Kevin R.; Sleaford, Brad W.; Robel, Martin; Smith, Brian W.

2011-07-17T23:59:59.000Z

358

Atomic Diffusion in the Uranium-50wt% Zirconium Nuclear Fuel System  

E-Print Network [OSTI]

Atomic diffusion phenomena were examined in a metal-alloy nuclear fuel system composed of ?-phase U-50wt%Zr fuel in contact with either Zr-10wt%Gd or Zr-10wt%Er. Each alloy was fabricated from elemental feed material via melt-casting, and diffusion...

Eichel, Daniel

2013-06-17T23:59:59.000Z

359

Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel  

SciTech Connect (OSTI)

The Next Generation Safeguard Initiative (NGSI) of the U.S Department of Energy is supporting a multi-lab/university collaboration to quantify the plutonium (Pu) mass in spent nuclear fuel (SNF) assemblies and to detect the diversion of pins with non-destructive assay (NDA) methods. The following 14 NDA techniques are being studied: Delayed Neutrons, Differential Die-Away, Differential Die-Away Self-Interrogation, Lead Slowing Down Spectrometer, Neutron Multiplicity, Passive Neutron Albedo Reactivity, Total Neutron (Gross Neutron), X-Ray Fluorescence, {sup 252}Cf Interrogation with Prompt Neutron Detection, Delayed Gamma, Nuclear Resonance Fluorescence, Passive Prompt Gamma, Self-integration Neutron Resonance Densitometry, and Neutron Resonance Transmission Analysis. Understanding and maturity of the techniques vary greatly, ranging from decades old, well-understood methods to new approaches. Nuclear Resonance Fluorescence (NRF) is a technique that had not previously been studied for SNF assay or similar applications. Since NRF generates isotope-specific signals, the promise and appeal of the technique lies in its potential to directly measure the amount of a specific isotope in an SNF assay target. The objectives of this study were to design and model suitable NRF measurement methods, to quantify capabilities and corresponding instrumentation requirements, and to evaluate prospects and the potential of NRF for SNF assay. The main challenge of the technique is to achieve the sensitivity and precision, i.e., to accumulate sufficient counting statistics, required for quantifying the mass of Pu isotopes in SNF assemblies. Systematic errors, considered a lesser problem for a direct measurement and only briefly discussed in this report, need to be evaluated for specific instrument designs in the future. Also, since the technical capability of using NRF to measure Pu in SNF has not been established, this report does not directly address issues such as cost, size, development time, nor concerns related to the use of Pu in measurement systems. This report discusses basic NRF measurement concepts, i.e., backscatter and transmission methods, and photon source and {gamma}-ray detector options in Section 2. An analytical model for calculating NRF signal strengths is presented in Section 3 together with enhancements to the MCNPX code and descriptions of modeling techniques that were drawn upon in the following sections. Making extensive use of the model and MCNPX simulations, the capabilities of the backscatter and transmission methods based on bremsstrahlung or quasi-monoenergetic photon sources were analyzed as described in Sections 4 and 5. A recent transmission experiment is reported on in Appendix A. While this experiment was not directly part of this project, its results provide an important reference point for our analytical estimates and MCNPX simulations. Used fuel radioactivity calculations, the enhancements to the MCNPX code, and details of the MCNPX simulations are documented in the other appendices.

Ludewigt, Bernhard A; Quiter, Brian J.; Ambers, Scott D.

2011-01-14T23:59:59.000Z

360

Overview of the International R&D Recycling Activities of the Nuclear Fuel Cycle  

SciTech Connect (OSTI)

Nuclear power has demonstrated over the last 30 years its capacity to produce base-load electricity at a low, predictable and stable cost due to the very low economic dependence on the price of uranium. However the management of used nuclear fuel remains the “Achilles’ Heel” of this energy source since the storage of used nuclear fuel is increasing as evidenced by the following number with 2,000 tons of UNF produced each year by the 104 US nuclear reactor units which equates to a total of 62,000 spent fuel assemblies stored in dry cask and 88,000 stored in pools. Two options adopted by several countries will be presented. The first one adopted by Europe, Japan and Russia consists of recycling the used nuclear fuel after irradiation in a nuclear reactor. Ninety six percent of uranium and plutonium contained in the spent fuel could be reused to produce electricity and are worth recycling. The separation of uranium and plutonium from the wastes is realized through the industrial PUREX process so that they can be recycled for re-use in a nuclear reactor as a mixed oxide (MOX) fuel. The second option undertaken by Finland, Sweden and the United States implies the direct disposal of used nuclear fuel into a geologic formation. One has to remind that only 30% of the worldwide used nuclear fuel are currently recycled, the larger part being stored (90% in pool) waiting for scientific or political decisions. A third option is emerging with a closed fuel cycle which will improve the global sustainability of nuclear energy. This option will not only decrease the volume amount of nuclear waste but also the long-term radiotoxicity of the final waste, as well as improving the long-term safety and the heat-loading of the final repository. At the present time, numerous countries are focusing on the R&D recycling activities of the ultimate waste composed of fission products and minor actinides (americium and curium). Several new chemical extraction processes, such as TRUSPEAK, EXAM, or LUCA processes are pursued worldwide and their approaches will be highlighted.

Patricia Paviet-Hartmann

2012-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear fuels limited" from the National Library of EnergyBeta (NLEBeta).
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361

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

362

FLOWSHEET EVALUATION FOR THE DISSOLVING AND NEUTRALIZATION OF SODIUM REACTOR EXPERIMENT USED NUCLEAR FUEL  

SciTech Connect (OSTI)

This report includes the literature review, hydrogen off-gas calculations, and hydrogen generation tests to determine that H-Canyon can safely dissolve the Sodium Reactor Experiment (SRE; thorium fuel), Ford Nuclear Reactor (FNR; aluminum alloy fuel), and Denmark Reactor (DR-3; silicide fuel, aluminum alloy fuel, and aluminum oxide fuel) assemblies in the L-Bundles with respect to the hydrogen levels in the projected peak off-gas rates. This is provided that the number of L-Bundles charged to the dissolver is controlled. Examination of SRE dissolution for potential issues has aided in predicting the optimal batching scenario. The calculations detailed in this report demonstrate that the FNR, SRE, and DR-3 used nuclear fuel (UNF) are bounded by MURR UNF and may be charged using the controls outlined for MURR dissolution in a prior report.

Daniel, W. E.; Hansen, E. K.; Shehee, T. C.

2012-10-30T23:59:59.000Z

363

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

364

Ukraine Loads U.S. Nuclear Fuel into Power Plant as Part of DOE-Ukraine  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Career Scientists' Research Petroleum ReserveDepartment of EnergyDepartment ofRemarksNuclear Fuel

365

Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides  

DOE Patents [OSTI]

Method for direct coprocessing of nuclear fuels derived from a product stream of a fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

Lloyd, Milton H. (Oak Ridge, TN)

1983-01-01T23:59:59.000Z

366

Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides  

DOE Patents [OSTI]

Method for direct coprocessing of nuclear fuels derived from a product stream of fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

Lloyd, M.H.

1981-01-09T23:59:59.000Z

367

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

368

Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 1, Appendix D, Part B: Naval spent nuclear fuel management  

SciTech Connect (OSTI)

This volume contains the following attachments: transportation of Naval spent nuclear fuel; description of Naval spent nuclear receipt and handling at the Expended Core Facility at the Idaho National Engineering Laboratory; comparison of storage in new water pools versus dry container storage; description of storage of Naval spent nuclear fuel at servicing locations; description of receipt, handling, and examination of Naval spent nuclear fuel at alternate DOE facilities; analysis of normal operations and accident conditions; and comparison of the Naval spent nuclear fuel storage environmental assessment and this environmental impact statement.

Not Available

1994-06-01T23:59:59.000Z

369

Savannah River Site Spent Nuclear Fuel Management Final Environmental Impact Statement  

SciTech Connect (OSTI)

The proposed DOE action considered in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets at the Savannah River Site (SRS) in Aiken County, South Carolina, including placing these materials in forms suitable for ultimate disposition. Options to treat, package, and store this material are discussed. The material included in this EIS consists of approximately 68 metric tons heavy metal (MTHM) of spent nuclear fuel 20 MTHM of aluminum-based spent nuclear fuel at SRS, as much as 28 MTHM of aluminum-clad spent nuclear fuel from foreign and domestic research reactors to be shipped to SRS through 2035, and 20 MTHM of stainless-steel or zirconium-clad spent nuclear fuel and some Americium/Curium Targets stored at SRS. Alternatives considered in this EIS encompass a range of new packaging, new processing, and conventional processing technologies, as well as the No Action Alternative. A preferred alternative is identified in which DOE would prepare about 97% by volume (about 60% by mass) of the aluminum-based fuel for disposition using a melt and dilute treatment process. The remaining 3% by volume (about 40% by mass) would be managed using chemical separation. Impacts are assessed primarily in the areas of water resources, air resources, public and worker health, waste management, socioeconomic, and cumulative impacts.

N /A

2000-04-14T23:59:59.000Z

370

LWR NUCLEAR FUEL BUNDLE DATA FOR USE IN FUEL BUNDLE HANDLING  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: VegetationEquipment Surfaces and Interfaces Sample6, 2011 LOS ALAMOS,Transition and Timeline AboutCoreLWR

371

Nuclear spin structure in dark matter search: The finite momentum transfer limit  

E-Print Network [OSTI]

Spin-dependent elastic scattering of weakly interacting massive dark matter particles (WIMP) off nuclei is reviewed. All available, within different nuclear models, structure functions S(q) for finite momentum transfer (q>0) are presented. These functions describe the recoil energy dependence of the differential event rate due to the spin-dependent WIMP-nucleon interactions. This paper, together with the previous paper ``Nuclear spin structure in dark matter search: The zero momentum transfer limit'', completes our review of the nuclear spin structure calculations involved in the problem of direct dark matter search.

V. A. Bednyakov; F. Simkovic

2006-08-09T23:59:59.000Z

372

Sputter-Deposited Pt/CrN Nanoparticle PEM Fuel Cell Cathodes: Limited Proton Conductivity Through Electrode  

E-Print Network [OSTI]

Sputter-Deposited Pt/CrN Nanoparticle PEM Fuel Cell Cathodes: Limited Proton Conductivity Through for proton exchange membrane PEM fuel cells. X-ray diffraction and scanning electron microscopy show manuscript received September 17, 2009. Published November 13, 2009. Proton exchange membrane PEM fuel cells

Gall, Daniel

373

Development of a techno-economic model to optimization DOE spent nuclear fuel disposition  

SciTech Connect (OSTI)

The purpose of the National Spent Nuclear Fuel (NSNF) Program conducted by Lockheed Martin Idaho Technology Co. (LMITCO) at the Idaho National Engineering and Environmental Laboratory (INEEL) is to evaluate what to do with the spent nuclear fuel (SNF) in the Department of Energy (DOE) complex. Final disposition of the SNF may require that the fuel be treated to minimize material concerns. The treatments may range from electrometallurgical treatment and chemical dissolution to engineering controls. Treatment options and treatment locations will depend on the fuel type and the current locations of the fuel. One of the first steps associated with selecting one or more sites for treating the SNF in the DOE complex is to determine the cost of each option. An economic analysis will assist in determining which fuel treatment alternative attains the optimum disposition of SNF at the lowest possible cost to the government and the public. For this study, a set of questions was developed for the electrometallurgical treatment process for fuels at several locations. The set of questions addresses all issues associated with the design, construction, and operation of a production facility. A matrix table was developed to determine questions applicable to various fuel treatment options. A work breakdown structure (WBS) was developed to identify a treatment process and costs from initial design to shipment of treatment products to final disposition. Costs will be applied to determine the life-cycle cost of each option. This technique can also be applied to other treatment techniques for treating spent nuclear fuel.

Ramer, R.J.; Plum, M.M.; Adams, J.P.; Dahl, C.A.

1997-11-01T23:59:59.000Z

374

Spent nuclear fuel storage -- Performance tests and demonstrations  

SciTech Connect (OSTI)

This report summarizes the results of heat transfer and shielding performance tests and demonstrations conducted from 1983 through 1992 by or in cooperation with the US Department of Energy (DOE), Office of Commercial Radioactive Waste Management (OCRWM). The performance tests consisted of 6 to 14 runs involving one or two loadings, usually three backfill environments (helium, nitrogen, and vacuum backfills), and one or two storage system orientations. A description of the test plan, spent fuel load patterns, results from temperature and dose rate measurements, and fuel integrity evaluations are contained within the report.

McKinnon, M.A.; DeLoach, V.A.

1993-04-01T23:59:59.000Z

375

Accelerator-driven subcritical fission in molten salt core: Closing the nuclear fuel cycle for green nuclear energy  

SciTech Connect (OSTI)

A technology for accelerator-driven subcritical fission in a molten salt core (ADSMS) is being developed as a basis for the destruction of the transuranics in used nuclear fuel. The molten salt fuel is a eutectic mixture of NaCl and the chlorides of the transuranics and fission products. The core is driven by proton beams from a strong-focusing cyclotron stack. This approach uniquely provides an intrinsically safe means to drive a core fueled only with transuranics, thereby eliminating competing breeding terms.

McIntyre, Peter; Assadi, Saeed; Badgley, Karie; Baker, William; Comeaux, Justin; Gerity, James; Kellams, Joshua; McInturff, Al; Pogue, Nathaniel; Sattarov, Akhdiyor; Sooby, Elizabeth; Tsvetkov, Pavel [Dept. of Physics, Texas A and M University, College Station, TX 77843 and Dept. of Nuclear Engineering, Texas A and M University, College Station, TX 77843 (United States); Phongikaroon, Supathorn; Simpson, Michael [Dept. of Chemical Engineering, University of Idaho, Idaho Falls ID 83402 (United States)

2013-04-19T23:59:59.000Z

376

Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel  

E-Print Network [OSTI]

Level Scheme Library Depleted Uranium Evaluated Nuclear Datafuel and the 238 U in depleted uranium (DU) was used as a

Ludewigt, Bernhard A

2011-01-01T23:59:59.000Z

377

Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process  

SciTech Connect (OSTI)

The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the “UREX+3c fuel cycle” and the “Alternative Fuel Cycle” (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount of the resulting MOX. The study considered two sub-cases within each of the two fuel cycles in which the uranium and plutonium from the first generation of MOX spent fuel (i) would not be recycled to produce a second generation of MOX for use in LWRs or (ii) would be recycled to produce a second generation of MOX fuel for use in LWRs. The study also investigated the effects of recycling MOX spent fuel multiple times in LWRs. The study assumed that both fuel cycles would store and then reprocess spent MOX fuel that is not recycled to produce a next generation of LWR MOX fuel and would use the recovered products to produce FR fuel. The study further assumed that FRs would begin to be brought on-line in 2043, eleven years after recycle begins in LWRs, when products from 5-year cooled spent MOX fuel would be available. Fuel for the FRs would be made using the uranium, plutonium, and minor actinides recovered from MOX. For the cases where LWR fuel was assumed to be recycled one time, the 1st generation of MOX spent fuel was used to provide nuclear materials for production of FR fuel. For the cases where the LWR fuel was assumed to be recycled two times, the 2nd generation of MOX spent fuel was used to provide nuclear materials for production of FR fuel. The number of FRs in operation was assumed to increase in successive years until the rate that actinides were recovered from permanently discharged spent MOX fuel equaled the rate the actinides were consumed by the operating fleet of FRs. To compare the two fuel cycles, the study analyzed recycle of nuclear fuel in LWRs and FRs and determined the radiological characteristics of irradiated nuclear fuel, nuclear waste products, and recycle nuclear fuels. It also developed a model to simulate the flows of nuclear materials that could occur in the two advanced nuclear fuel cycles over 81 years beginning in 2020 and ending in 2100. Simulations projected the flows of uranium, plutonium, and minor actinides as these nuclear fuel materials were produced and consumed in a fleet of 100 1,000 MWe LWRs and in FRs. The model als

E. R. Johnson; R. E. Best

2009-12-28T23:59:59.000Z

378

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

379

Meeting U.S. Liquid Transport Fuel Needs with a Nuclear Hydrogen Biomass System  

SciTech Connect (OSTI)

The two major energy challenges for the United States are replacing crude oil in our transportation system and eliminating greenhouse gas emissions. A domestic-source greenhouse-gas-neutral nuclear hydrogen biomass system to replace oil in the transportation sector is described. Some parts of the transportation system can be electrified with electricity supplied by nuclear energy sources that do not emit significant quantities of greenhouse gases. Other components of the transportation system require liquid fuels. Biomass can be converted to greenhouse-gas-neutral liquid fuels; however, the conversion of biomass-to-liquid fuels is energy intensive. There is insufficient biomass to meet U.S. liquid fuel demands and provide the energy required to process the biomass-to-liquid fuels. With the use of nuclear energy to provide heat, electricity, and hydrogen for the processing of biomass-to-liquid fuels, the liquid fuel production per unit of biomass is dramatically increased, and the available biomass could meet U.S. liquid fuel requirements.

Forsberg, Charles W [ORNL

2007-01-01T23:59:59.000Z

380

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

Note: This page contains sample records for the topic "nuclear fuels limited" 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

Room at the Mountain: Estimated Maximum Amounts of Commercial Spent Nuclear Fuel Capable of Disposal in a Yucca Mountain Repository  

SciTech Connect (OSTI)

The purpose of this paper is to present an initial analysis of the maximum amount of commercial spent nuclear fuel (CSNF) that could be emplaced into a geological repository at Yucca Mountain. This analysis identifies and uses programmatic, material, and geological constraints and factors that affect this estimation of maximum amount of CSNF for disposal. The conclusion of this initial analysis is that the current legislative limit on Yucca Mountain disposal capacity, 63,000 MTHM of CSNF, is a small fraction of the available physical capacity of the Yucca Mountain system assuming the current high-temperature operating mode (HTOM) design. EPRI is confident that at least four times the legislative limit for CSNF ({approx}260,000 MTHM) can be emplaced in the Yucca Mountain system. It is possible that with additional site characterization, upwards of nine times the legislative limit ({approx}570,000 MTHM) could be emplaced. (authors)

Kessler, John H. [Electric Power Research Institute - EPRI, 3420 Hillview Avenue, Palo Alto, California 94304 (United States); Kemeny, John [University of Arizona, Tucson AZ 85721 (United States); King, Fraser [Integrity Corrosion Consulting, Ltd., 6732 Silverview Drive NW, Calgary, Alberta (Canada); Ross, Alan M. [Alan M. Ross and Associates, 1061 Gray Fox Circle Pleasanton, CA 94566 (Canada); Ross, Benjamen [Disposal Safety, Inc., Bethesda, MD 20814 (United States)

2006-07-01T23:59:59.000Z

382

Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing  

SciTech Connect (OSTI)

Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training.

D.R. Jackson; G.R. Kiebel

1999-08-24T23:59:59.000Z

383

TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES  

SciTech Connect (OSTI)

A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection prior to loading. A measurement of the average assembly burnup is required and that measurement must be within 10% of the utility burnup record for the assembly to be accepted. The measurement device must be accurate to within 10%. Each step is described in detail for use with any computer code system and is then demonstrated with the SCALE 4.2 computer code package using 27BURNUPLIB cross sections.

DOE

1997-04-01T23:59:59.000Z

384

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

385

Nuclear structure aspects of neutrinoless {beta}{beta} decay: limits on the electron neutrino mass  

SciTech Connect (OSTI)

We discuss some features of the nuclear structure elements participant in the calculation of the mass sector of the half-life of the neutrinoless double beta decay, and the consequences upon the adopted limits of the electron-neutrino mass.

Civitarese, O. [Department of Physics. University of La Plata, 1900 La Plata (Argentina); Suhonen, J. [Department of Physics. University of Jyvaeskyla, (Finland)

2007-02-12T23:59:59.000Z

386

Safeguarding Nuclear Fuel Processing | U.S. DOE Office of Science (SC)  

Office of Science (SC) Website

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear SecurityTensile Strain Switched5 IndustrialIsadore Perlman,BiosScience (SC) Regional &ReviewSLACSafeguarding

387

Public Acceptability of and Preferences for Used Nuclear Fuel Management in the U.S.  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 - SeptemberMicroneedlesAdvanced PhotonPseudo-Single-BunchPseudogaps,Pu Qian Pu45d

388

Separation of Nuclear Fuel Surrogates from Silicon Carbide Inert Matrix  

SciTech Connect (OSTI)

The objective of this project has been to identify a process for separating transuranic species from silicon carbide (SiC). Silicon carbide has become one of the prime candidates for the matrix in inert matrix fuels, (IMF) being designed to reduce plutonium inventories and the long half-lives actinides through transmutation since complete reaction is not practical it become necessary to separate the non-transmuted materials from the silicon carbide matrix for ultimate reprocessing. This work reports a method for that required process.l

Dr. Ronald Baney

2008-12-15T23:59:59.000Z

389

The Impact of Microbially Influenced Corrosion on Spent Nuclear Fuel and Storage Life  

SciTech Connect (OSTI)

A study was performed to evaluate if microbial activity could be considered a threat to spent nuclear fuel integrity. The existing data regarding the impact of microbial influenced corrosion (MIC) on spent nuclear fuel storage does not allow a clear assessment to be made. In order to identify what further data are needed, a literature survey on MIC was accomplished with emphasis on materials used in nuclear fuel fabrication, e.g., A1, 304 SS, and zirconium. In addition, a survey was done at Savannah River, Oak Ridge, Hanford, and the INEL on the condition of their wet storage facilities. The topics discussed were the SNF path forward, the types of fuel, ramifications of damaged fuel, involvement of microbial processes, dry storage scenarios, ability to identify microbial activity, definitions of water quality, and the use of biocides. Information was also obtained at international meetings in the area of biological mediated problems in spent fuel and high level wastes. Topics dis cussed included receiving foreign reactor research fuels into existing pools, synergism between different microbes and other forms of corrosion, and cross contamination.

J. H. Wolfram; R. E. Mizia; R. Jex; L. Nelson; K. M. Garcia

1996-10-01T23:59:59.000Z

390

Criticality safety considerations in the geologic disposal of spent nuclear fuel assemblies  

SciTech Connect (OSTI)

Features of geologic disposal which hamper the demonstration that criticality cannot occur therein include possible changes of shape and form, intrusion of water as a neutron moderator, and selective leaching of spent fuel constituents. If the criticality safety of spent fuel disposal depends on burnup, independent measurements verifying the burnup should be performed prior to disposal. The status of nondestructive analysis method which might provide such verification is discussed. Calculations were performed to assess the potential for increasing the allowed size of a spent fuel disposal canister if potential water intrusion were limited by close-packing the enclosed rods. Several factors were identified which severely limited the potential of this application. The theoretical limit of hexagonal close-packing cannot be achieved due to fuel rod bowing. It is concluded that disposal canisters should be sized on the basis of assumed optimum moderation. Several topics for additional research were identified during this limited study.

Gore, B.F.; McNair, G.W.; Heaberlin, S.W.

1980-05-01T23:59:59.000Z

391

Development of High Fidelity, Fuel-Like Thermal Simulators for Non-Nuclear Testing  

SciTech Connect (OSTI)

Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Work at the NASA Marshall Space Flight Center seeks to develop high fidelity thermal simulators that not only match the static power profile that would be observed in an operating, fueled nuclear reactor, but also match the dynamic fuel pin performance during feasible transients. Comparison between the fuel pins and thermal simulators is made at the outer fuel clad surface, which corresponds to the outer sheath surface in the thermal simulator. The thermal simulators that are currently being developed are designed to meet the geometric and power requirements of a proposed surface power reactor design, accommodate testing of various axial power profiles, and incorporate imbedded instrumentation. Static and dynamic fuel pin performances for a proposed reactor design have been determined using SINDA/FLUINT thermal analysis software, and initial comparison has been made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analysis, a conceptual high fidelity design will be developed, followed by engineering design, fabrication, and testing to validate the overall design process. Although the resulting thermal simulator will be designed for a specific reactor concept, establishing this rigorous design process will assist in streamlining the thermal simulator development for other reactor concepts. This paper presents the current status of high fidelity thermal simulator design relative to a SNAP derivative reactor design that could be applied for Lunar surface power.

Bragg-Sitton, Shannon M.; Dickens, Ricky; Adams, Mike; Davis, Joe [NASA Marshall Space Flight Center, Nuclear Systems Branch/ER24, MSFC, AL 25812 (United States); Dixon, David [Los Alamos National Laboratory, Decision Applications Division, Los Alamos, NM 87545 (United States); North Carolina State University, Raleigh, NC (United States); Kapernick, Richard [Los Alamos National Laboratory, Decision Applications Division, Los Alamos, NM 87545 (United States)

2007-01-30T23:59:59.000Z

392

Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs draft environmental impact statement. Volume 1, Appendix B: Idaho National Engineering Laboratory Spent Nuclear Fuel Management Program  

SciTech Connect (OSTI)

The US Department of Energy (DOE) has prepared this report to assist its management in making two decisions. The first decision, which is programmatic, is to determine the management program for DOE spent nuclear fuel. The second decision is on the future direction of environmental restoration, waste management, and spent nuclear fuel management activities at the Idaho National Engineering Laboratory. Volume 1 of the EIS, which supports the programmatic decision, considers the effects of spent nuclear fuel management on the quality of the human and natural environment for planning years 1995 through 2035. DOE has derived the information and analysis results in Volume 1 from several site-specific appendixes. Volume 2 of the EIS, which supports the INEL-specific decision, describes environmental impacts for various environmental restoration, waste management, and spent nuclear fuel management alternatives for planning years 1995 through 2005. This Appendix B to Volume 1 considers the impacts on the INEL environment of the implementation of various DOE-wide spent nuclear fuel management alternatives. The Naval Nuclear Propulsion Program, which is a joint Navy/DOE program, is responsible for spent naval nuclear fuel examination at the INEL. For this appendix, naval fuel that has been examined at the Naval Reactors Facility and turned over to DOE for storage is termed naval-type fuel. This appendix evaluates the management of DOE spent nuclear fuel including naval-type fuel.

Not Available

1994-06-01T23:59:59.000Z

393

An evaluation of thermal modeling techniques utilized for nuclear fuel rods  

E-Print Network [OSTI]

like to thank my graduate advisor, Dr. K. L. Peddicord, for his technical advice and guidance throughout this project and my studies in Nuclear Engineering at Texas AgiM University. Thanks are also extended to Dr. Hassan and Dr. Caton for reviewing...AN EVALUATION OF THERMAL MODELING TECHNIQUES UTILIZED FOR NUCLEAR FUEL RODS A Thesis by JEFFREY WARREN SIMMONS Submitted to the Office of Graduate Studies of Texas Asr M University in partial fulfillment of the requirements for the degree...

Simmons, Jeffrey Warren

1989-01-01T23:59:59.000Z

395

CONTROL OF POPULATION DENSITIES SURROUNDING NUCLEAR POWER PLANTS. VOLUME 5 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA Energy and Environment

Nero, jA.V.

2010-01-01T23:59:59.000Z

396

Overcoming the Range Limitation of Medium-Duty Battery Electric Vehicles through the use of Hydrogen Fuel-Cells  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 - September 2006 The 2002Optics Group (X-rayLSDPreciousM206 Mechanical Convection

397

Draft Environmental Impact Statement for the Treatment and Management of Sodium-Bonded Spent Nuclear Fuel  

SciTech Connect (OSTI)

This document summarizes the U.S. Department of Energy's ''Draft Environmental Impact Statement for the Treatment and Management of Sodium-Bonded Spent Nuclear Fuel''. In addition to information concerning the background, purpose and need for the proposed action, and the National Environmental Policy Act process, this summary describes the characteristics of sodium-bonded spent nuclear fuel, the proposed treatment methods, the proposed facilities, the alternatives considered, and the environmental consequences of these alternatives. A glossary is included at the end to assist the reader with some of the technical terms used in this document.

N /A

1999-07-30T23:59:59.000Z

398

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

399

A Preliminary Evaluation of Using Fill Materials to Stabilize Used Nuclear Fuel During Storage and Transportation  

SciTech Connect (OSTI)

This report contains a preliminary evaluation of potential fill materials that could be used to fill void spaces in and around used nuclear fuel contained in dry storage canisters in order to stabilize the geometry and mechanical structure of the used nuclear fuel during extended storage and transportation after extended storage. Previous work is summarized, conceptual descriptions of how canisters might be filled were developed, and requirements for potential fill materials were developed. Elements of the requirements included criticality avoidance, heat transfer or thermodynamic properties, homogeneity and rheological properties, retrievability, material availability and cost, weight and radiation shielding, and operational considerations. Potential fill materials were grouped into 5 categories and their properties, advantages, disadvantages, and requirements for future testing were discussed. The categories were molten materials, which included molten metals and paraffin; particulates and beads; resins; foams; and grout. Based on this analysis, further development of fill materials to stabilize used nuclear fuel during storage and transportation is not recommended unless options such as showing that the fuel remains intact or canning of used nuclear fuel do not prove to be feasible.

Maheras, Steven J.; Best, Ralph; Ross, Steven B.; Lahti, Erik A.; Richmond, David J.

2012-08-01T23:59:59.000Z

400

Nuclear Industry Input to the Development of Concepts for the Consolidated Storage of Used Nuclear Fuel - 13411  

SciTech Connect (OSTI)

EnergySolutions and its team partners, NAC International, Exelon Nuclear Partners, Talisman International, TerranearPMC, Booz Allen Hamilton and Sargent and Lundy, have carried out a study to develop concepts for a Consolidated Storage Facility (CSF) for the USA's stocks of commercial Used Nuclear Fuel (UNF), and the packaging and transport provisions required to move the UNF to the CSF. The UNF is currently stored at all 65 operating nuclear reactor sites in the US, and at 10 shutdown sites. The study was funded by the US Department of Energy and followed the recommendations of the Blue Ribbon Commission on America's Nuclear Future (BRC), one of which was that the US should make prompt efforts to develop one or more consolidated storage facilities for commercial UNF. The study showed that viable schemes can be devised to move all UNF and store it at a CSF, but that a range of schemes is required to accommodate the present widely varying UNF storage arrangements. Although most UNF that is currently stored at operating reactor sites is in water-filled pools, a significant amount is now dry stored in concrete casks. At the shutdown sites, the UNF is dry stored at all but two of the ten sites. Various types of UNF dry storage configurations are used at the operating sites and shutdown sites that include vertical storage casks that are also licensed for transportation, vertical casks that are licensed for storage only, and horizontally orientated storage modules. The shutdown sites have limited to nonexistent UNF handling infrastructure and several no longer have railroad connections, complicating UNF handling and transport off the site. However four methods were identified that will satisfactorily retrieve the UNF canisters within the storage casks and transport them to the CSF. The study showed that all of the issues associated with the transportation and storage of UNF from all sites in the US can be accommodated by adopting a staged approach to the construction of the CSF. Stage 1 requires only a cask storage pad and railroad interface to be constructed, and the CSF can then receive the UNF that is in transportable storage casks. Stage 2 adds a canister handling facility, a storage cask fabrication facility and an expanded storage pad, and enables the receipt of all canistered UNF from both operating and shutdown sites. Stage 3 provides a repackaging facility with a water-filled pool that provides flexibility for a range of repackaging scenarios. This includes receiving and repackaging 'bare' UNF into suitable canisters that can be placed into interim storage at the CSF, and enables UNF that is being received, or already in storage onsite, to be repackaged into canisters that are suitable for disposal at a geologic repository. The study used the 'Total System Model' (TSM) to analyze a range of CSF capacities and operating scenarios with differing parameters covering UNF pickup orders, one or more CSF sites, CSF start dates, CSF receipt rates and geologic repository start dates. The TSM was originally developed to model movement of UNF to the Yucca Mountain repository and was modified for this study to enable the CSF to become the 'gateway' to a future geologic repository. The TSM analysis enabled costs to be estimated for each scenario and showed how these are influenced by each of the parameters. This information will provide essential underpinning for a future Conceptual Design preparation. (authors)

Phillips, Chris; Thomas, Ivan; McNiven, Steven [EnergySolutions Federal EPC., 2345 Stevens Drive, Richland, WA, 99354 (United States)] [EnergySolutions Federal EPC., 2345 Stevens Drive, Richland, WA, 99354 (United States); Lanthrum, Gary [NAC International, 3930 East Jones Bridge Road, Norcross, GA, 30092 (United States)] [NAC International, 3930 East Jones Bridge Road, Norcross, GA, 30092 (United States)

2013-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear fuels limited" 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

New generation nuclear fuel structures: dense particles in selectively soluble matrix  

SciTech Connect (OSTI)

We have developed a technology for dispersing sub-millimeter sized fuel particles within a bulk matrix that can be selectively dissolved. This may enable the generation of advanced nuclear fuels with easy separation of actinides and fission products. The large kinetic energy of the fission products results in most of them escaping from the sub-millimeter sized fuel particles and depositing in the matrix during burning of the fuel in the reactor. After the fuel is used and allowed to cool for a period of time, the matrix can be dissolved and the fission products removed for disposal while the fuel particles are collected by filtration for recycle. The success of such an approach would meet a major goal of the GNEP program to provide advanced recycle technology for nuclear energy production. The benefits of such an approach include (1) greatly reduced cost of the actinide/fission product separation process, (2) ease of recycle of the fuel particles, and (3) a radiation barrier to prevent theft or diversion of the recycled fuel particles during the time they are re-fabricated into new fuel. In this study we describe a method to make surrogate nuclear fuels of micrometer scale W (shell)/Mo (core) or HfO2 particles embedded in an MgO matrix that allows easy separation of the fission products and their embedded particles. In brief, the method consists of physically mixing W-Mo or hafnia particles with an MgO precursor. Heating the mixture, in air or argon, without agitation, to a temperature is required for complete decomposition of the precursor. The resulting material was examined using chemical analysis, scanning electron microscopy, X-ray diffraction and micro X-ray computed tomography and found to consist of evenly dispersed particles in an MgO + matrix. We believe this methodology can be extended to actinides and other matrix materials.

Sickafus, Kurt E [Los Alamos National Laboratory; Devlin, David J [Los Alamos National Laboratory; Jarvinen, Gordon D [Los Alamos National Laboratory; Patterson, Brian M [Los Alamos National Laboratory; Pattillo, Steve G [Los Alamos National Laboratory; Valdez, James [Los Alamos National Laboratory; Phillips, Jonathan [Los Alamos National Laboratory

2009-01-01T23:59:59.000Z

402

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

403

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

SciTech Connect (OSTI)

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

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

1997-12-01T23:59:59.000Z

404

Conceptual design report for the ICPP spent nuclear fuel dry storage project  

SciTech Connect (OSTI)

The conceptual design is presented for a facility to transfer spent nuclear fuel from shipping casks to dry storage containers, and to safely store those containers at ICPP at INEL. The spent fuels to be handled at the new facility are identified and overall design and operating criteria established. Physical configuration of the facility and the systems used to handle the SNF are described. Detailed cost estimate for design and construction of the facility is presented.

NONE

1996-07-01T23:59:59.000Z

405

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

406

THE ECONOMICS OF REPROCESSING vs DIRECT DISPOSAL OF SPENT NUCLEAR FUEL  

SciTech Connect (OSTI)

This report assesses the economics of reprocessing versus direct disposal of spent nuclear fuel. The breakeven uranium price at which reprocessing spent nuclear fuel from existing light-water reactors (LWRs) and recycling the resulting plutonium and uranium in LWRs would become economic is assessed, using central estimates of the costs of different elements of the nuclear fuel cycle (and other fuel cycle input parameters), for a wide range of range of potential reprocessing prices. Sensitivity analysis is performed, showing that the conclusions reached are robust across a wide range of input parameters. The contribution of direct disposal or reprocessing and recycling to electricity cost is also assessed. The choice of particular central estimates and ranges for the input parameters of the fuel cycle model is justified through a review of the relevant literature. The impact of different fuel cycle approaches on the volume needed for geologic repositories is briefly discussed, as are the issues surrounding the possibility of performing separations and transmutation on spent nuclear fuel to reduce the need for additional repositories. A similar analysis is then performed of the breakeven uranium price at which deploying fast neutron breeder reactors would become competitive compared with a once-through fuel cycle in LWRs, for a range of possible differences in capital cost between LWRs and fast neutron reactors. Sensitivity analysis is again provided, as are an analysis of the contribution to electricity cost, and a justification of the choices of central estimates and ranges for the input parameters. The equations used in the economic model are derived and explained in an appendix. Another appendix assesses the quantities of uranium likely to be recoverable worldwide in the future at a range of different possible future prices.

Matthew Bunn; Steve Fetter; John P. Holdren; Bob van der Zwaan

2003-07-01T23:59:59.000Z

407

Development of a Techno-Economic Model to Optimize DOE Spent Nuclear Fuel Disposition  

SciTech Connect (OSTI)

The National Spent Nuclear Fuel (NSNF) Program is evaluating final disposition of spent nuclear fuel (SNE) in the Department of Energy (DOE) complex. Final disposition of SNF may require that the fuel be treated to minimize material concerns. The treatments may range from electrometallurgical treatment (EMT) and chemical dissolution to engineering controls. Treatment options and treatment locations will depend on fuel type and location of the fuel. One of the first steps associated with selecting one or more sites for treating SNF in the DOE complex is to determine the cost of each option. An economic analysis will assist in determining which fuel treatment alternative attains the optimum disposition of SNF at the lowest possible cost to the government and the public. For this study, a set of questions was developed for the EMT process for fuels at several locations. The set of questions addresses all issues associated with design, construction, and operation of a production facility. A matrix table was developed to determine questions applicable to various fuel treatment options. A work breakdown structure (WBS) was developed to identify a treatment process and costs from initial design to shipment of treatment products to final disposition. Costs can be applied to determine the life cycle cost of each option. This technique can also be applied to other treatment techniques for treating SNF.

Ramer, R. J.; Plum, M. M.; Adams, J. P.; Dahl, C. A.

1998-02-01T23:59:59.000Z

408

Limiter  

DOE Patents [OSTI]

A limiter with a specially contoured front face is provided. The front face of the limiter (the plasma-side face) is flat with a central indentation. In addition, the limiter shape is cylindrically symmetric so that the limiter can be rotated for greater heat distribution. This limiter shape accommodates the various power scrape-off distances lambda p, which depend on the parallel velocity, V/sub parallel/, of the impacting particles.

Cohen, S.A.; Hosea, J.C.; Timberlake, J.R.

1984-10-19T23:59:59.000Z

409

Functionalized ultra-porous titania nanofiber membranes as nuclear waste separation and sequestration scaffolds for nuclear fuels recycle.  

SciTech Connect (OSTI)

Advanced nuclear fuel cycle concept is interested in reducing separations to a simplified, one-step process if possible. This will benefit from the development of a one-step universal getter and sequestration material so as a simplified, universal waste form was proposed in this project. We have developed a technique combining a modified sol-gel chemistry and electrospinning for producing ultra-porous ceramic nanofiber membranes with controllable diameters and porous structures as the separation/sequestration materials. These ceramic nanofiber materials have been determined to have high porosity, permeability, loading capacity, and stability in extreme conditions. These porous fiber membranes were functionalized with silver nanoparticles and nanocrystal metal organic frameworks (MOFs) to introduce specific sites to capture gas species that are released during spent nuclear fuel reprocessing. Encapsulation into a durable waste form of ceramic composition was also demonstrated.

Liu, Haiqing; Bell, Nelson Simmons; Cipiti, Benjamin B.; Lewis, Tom Goslee,; Sava, Dorina Florentina; Nenoff, Tina Maria

2012-09-01T23:59:59.000Z

410

Enterprise SRS: leveraging ongoing operations to advance nuclear fuel cycles research and development programs  

SciTech Connect (OSTI)

The Savannah River Site (SRS) is re-purposing its vast array of assets (including H Canyon - a nuclear chemical separation plant) to solve issues regarding advanced nuclear fuel cycle technologies, nuclear materials processing, packaging, storage and disposition. The vehicle for this transformation is Enterprise SRS which presents a new, radical view of SRS as a united endeavor for 'all things nuclear' as opposed to a group of distinct and separate entities with individual missions and organizations. Key among the Enterprise SRS strategic initiatives is the integration of research into SRS facilities but also in other facilities in conjunction with on-going missions to provide researchers from other national laboratories, academic institutions, and commercial entities the opportunity to demonstrate their technologies in a relevant environment and scale prior to deployment. To manage that integration of research demonstrations into site facilities, a center for applied nuclear materials processing and engineering research has been established in SRS.

Murray, A.M.; Marra, J.E.; Wilmarth, W.R. [Savannah River National Laboratory, Aiken, SC 29808 (United States); McGuire, P.W.; Wheeler, V.B. [Department of Energy-Savannah River Operations Office, Aiken SC 29808 (United States)

2013-07-01T23:59:59.000Z

411

DEMONSTRATION OF LONG-TERM STORAGE CAPABILITY FOR SPENT NUCLEAR FUEL IN L BASIN  

SciTech Connect (OSTI)

The U.S. Department of Energy decisions for the ultimate disposition of its inventory of used nuclear fuel presently in, and to be received and stored in, the L Basin at the Savannah River Site, and schedule for project execution have not been established. A logical decision timeframe for the DOE is following the review of the overall options for fuel management and disposition by the Blue Ribbon Commission on America's Nuclear Future (BRC). The focus of the BRC review is commercial fuel; however, the BRC has included the DOE fuel inventory in their review. Even though the final report by the BRC to the U.S. Department of Energy is expected in January 2012, no timetable has been established for decisions by the U.S. Department of Energy on alternatives selection. Furthermore, with the imminent lay-up and potential closure of H-canyon, no ready path for fuel disposition would be available, and new technologies and/or facilities would need to be established. The fuel inventory in wet storage in the 3.375 million gallon L Basin is primarily aluminum-clad, aluminum-based fuel of the Materials Test Reactor equivalent design. An inventory of non-aluminum-clad fuel of various designs is also stored in L Basin. Safe storage of fuel in wet storage mandates several high-level 'safety functions' that would be provided by the Structures, Systems, and Components (SSCs) of the storage system. A large inventory of aluminum-clad, aluminum-based spent nuclear fuel, and other nonaluminum fuel owned by the U.S. Department of Energy is in wet storage in L Basin at the Savannah River Site. An evaluation of the present condition of the fuel, and the Structures, Systems, or Components (SSCs) necessary for its wet storage, and the present programs and storage practices for fuel management have been performed. Activities necessary to validate the technical bases for, and verify the condition of the fuel and the SSCs under long-term wet storage have also been identified. The overall conclusion is that the fuel can be stored in L Basin, meeting general safety functions for fuel storage, for an additional 50 years and possibly beyond contingent upon continuation of existing fuel management activities and several augmented program activities. It is concluded that the technical bases and well-founded technologies have been established to store spent nuclear fuel in the L Basin. Methodologies to evaluate the fuel condition and characteristics, and systems to prepare fuel, isolate damaged fuel, and maintain water quality storage conditions have been established. Basin structural analyses have been performed against present NPH criteria. The aluminum fuel storage experience to date, supported by the understanding of the effects of environmental variables on materials performance, demonstrates that storage systems that minimize degradation and provide full retrievability of the fuel up to and greater than 50 additional years will require maintaining the present management programs, and with the recommended augmented/additional activities in this report.

Sindelar, R.; Deible, R.

2011-04-27T23:59:59.000Z

412

Extended Dry Storage of Used Nuclear Fuel: Technical Issues: A USA Perspective  

SciTech Connect (OSTI)

Used nuclear fuel will likely be stored dry for extended periods of time in the USA. Until a final disposition pathway is chosen, the storage periods will almost definitely be longer than were originally intended. The ability of the important-to-safety structures, systems and components (SSC's) continue to meet storage and transport safety functions over extended times must be determined. It must be assured that there is no significant degradation of the fuel or dry cask storage systems. Also, it is projected that the maximum discharge burnups of the used nuclear fuel will increase. Thus, it is necessary to obtain data on high burnup fuel to demonstrate that the used nuclear fuel remains intact after extended storage. An evaluation was performed to determine the conditions that may lead to failure of dry storage SSC's. This paper documents the initial technical gap analysis performed to identify data and modeling needs to develop the desired technical bases to ensure the safety functions of dry stored fuel.

McConnell, Paul; Hanson, Brady D.; Lee, Moo; Sorenson, Ken B.

2011-10-28T23:59:59.000Z

413

Nuclear Materials Characterization in the Materials and Fuels Complex Analytical Hot Cells  

SciTech Connect (OSTI)

As energy prices skyrocket and interest in alternative, clean energy sources builds, interest in nuclear energy has increased. This increased interest in nuclear energy has been termed the “Nuclear Renaissance”. The performance of nuclear fuels, fuels and reactor materials and waste products are becoming a more important issue as the potential for designing new nuclear reactors is more immediate. The Idaho National Laboratory (INL) Materials and Fuels Complex (MFC) Analytical Laboratory Hot Cells (ALHC) are rising to the challenge of characterizing new reactor materials, byproducts and performance. The ALHC is a facility located near Idaho Falls, Idaho at the INL Site. It was built in 1958 as part of the former Argonne National Laboratory West Complex to support the operation of the second Experimental Breeder Reactor (EBR-II). It is part of a larger analytical laboratory structure that includes wet chemistry, instrumentation and radiochemistry laboratories. The purpose of the ALHC is to perform analytical chemistry work on highly radioactive materials. The primary work in the ALHC has traditionally been dissolution of nuclear materials so that less radioactive subsamples (aliquots) could be transferred to other sections of the laboratory for analysis. Over the last 50 years though, the capabilities within the ALHC have also become independent of other laboratory sections in a number of ways. While dissolution, digestion and subdividing samples are still a vitally important role, the ALHC has stand alone capabilities in the area of immersion density, gamma scanning and combustion gas analysis. Recent use of the ALHC for immersion density shows that extremely fine and delicate operations can be performed with the master-slave manipulators by qualified operators. Twenty milligram samples were tested for immersion density to determine the expansion of uranium dioxide after irradiation in a nuclear reactor. The data collected confirmed modeling analysis with very tight precision. The gamma scanning equipment in the ALHC has taken on a new role also as a micro-gamma scanning system and has been put into service; allowing the linear and radial counting of a spent fuel segment to determine reaction characteristics within a small section of nuclear fuel. The nitrogen, oxygen and carbon analysis allows the identification of these impurities in spent nuclear fuel and also most oxides, nitrides, carbides, C-14 and tritium.

Michael Rodriquez

2009-03-01T23:59:59.000Z

414

Roadmap to an Engineering-Scale Nuclear Fuel Performance & Safety Code  

SciTech Connect (OSTI)

Developing new fuels and qualifying them for large-scale deployment in power reactors is a lengthy and expensive process, typically spanning a period of two decades from concept to licensing. Nuclear fuel designers serve an indispensable role in the process, at the initial exploratory phase as well as in analysis of the testing results. In recent years fuel performance capabilities based on first principles have been playing more of a role in what has traditionally been an empirically dominated process. Nonetheless, nuclear fuel behavior is based on the interaction of multiple complex phenomena, and recent evolutionary approaches are being applied more on a phenomenon-by-phenomenon basis, targeting localized problems, as opposed to a systematic approach based on a fundamental understanding of all interacting parameters. Advanced nuclear fuels are generally more complex, and less understood, than the traditional fuels used in existing reactors (ceramic UO{sub 2} with burnable poisons and other minor additives). The added challenges are primarily caused by a less complete empirical database and, in the case of recycled fuel, the inherent variability in fuel compositions. It is clear that using the traditional approach to develop and qualify fuels over the entire range of variables pertinent to the U.S. Department of Energy (DOE) Office of Nuclear Energy on a timely basis with available funds would be very challenging, if not impossible. As a result the DOE Office of Nuclear Energy has launched the Nuclear Energy Advanced Modeling and Simulation (NEAMS) approach to revolutionize fuel development. This new approach is predicated upon transferring the recent advances in computational sciences and computer technologies into the fuel development program. The effort will couple computational science with recent advances in the fundamental understanding of physical phenomena through ab initio modeling and targeted phenomenological testing to leapfrog many fuel-development activities. Realizing the full benefits of this approach will likely take some time. However, it is important that the developmental activities for modeling and simulation be tightly coupled with the experimental activities to maximize feedback effects and accelerate both the experimental and analytical elements of the program toward a common objective. The close integration of modeling and simulation and experimental activities is key to developing a useful fuel performance simulation capability, providing a validated design and analysis tool, and understanding the uncertainties within the models and design process. The efforts of this project are integrally connected to the Transmutation Fuels Campaign (TFC), which maintains as a primary objective to formulate, fabricate, and qualify a transuranic-based fuel with added minor actinides for use in future fast reactors. Additional details of the TFC scope can be found in the Transmutation Fuels Campaign Execution Plan. This project is an integral component of the TFC modeling and simulation effort, and this multiyear plan borrowed liberally from the Transmutation Fuels Campaign Modeling and Simulation Roadmap. This document provides the multiyear staged development plan to develop a continuum-level Integrated Performance and Safety Code (IPSC) to predict the behavior of the fuel and cladding during normal reactor operations and anticipated transients up to the point of clad breach.

Turner, John A [ORNL; Clarno, Kevin T [ORNL; Hansen, Glen A [ORNL

2009-09-01T23:59:59.000Z

415

Application of Diagnostic/Prognostic Methods to Critical Equipment for the Spent Nuclear Fuel Cleanup Program  

SciTech Connect (OSTI)

The management of the Spent Nuclear Fuel (SNF) project at the Hanford K-Basin in the 100 N Area has successfully restructured the preventive maintenance, spare parts inventory requirements, and the operator rounds data requirements. In this investigation, they continue to examine the different facets of the operations and maintenance (O&M) of the K-Basin cleanup project in search of additional reliability and cost savings. This report focuses on the initial findings of a team of PNNL engineers engaged to identify potential opportunities for reducing the cost of O&M through the application of advanced diagnostics (fault determination) and prognostics (residual life/reliability determination). The objective is to introduce predictive technologies to eliminate or reduce high impact equipment failures. The PNNL team in conjunction with the SNF engineers found the following major opportunities for cost reduction and/or enhancing reliability: (1) Provide data routing and automated analysis from existing detection systems to a display center that will engage the operations and engineering team. This display will be operator intuitive with system alarms and integrated diagnostic capability. (2) Change operating methods to reduce major transients induced in critical equipment. This would reduce stress levels on critical equipment. (3) Install a limited sensor set on failure prone critical equipment to allow degradation or stressor levels to be monitored and alarmed. This would provide operators and engineers with advance guidance and warning of failure events. Specific methods for implementation of the above improvement opportunities are provided in the recommendations. They include an Integrated Water Treatment System (IWTS) decision support system, introduction of variable frequency drives on certain pump motors, and the addition of limited diagnostic instrumentation on specified critical equipment.

Casazza, Lawrence O.; Jarrell, Donald B.; Koehler, Theresa M.; Meador, Richard J.; Wallace, Dale E.

2002-02-28T23:59:59.000Z

416

Nuclear Dynamics Consequence Analysis (NDCA) for the Disposal of Spent Nuclear Fuel in an Underground Geologic Repository - Volume 3: Appendices  

SciTech Connect (OSTI)

The United States Department of Energy Office of Environmental Management's (DOE/EM's) National Spent Nuclear Fuel Program (NSNFP), through a collaboration between Sandia National Laboratories (SNL) and Idaho National Engineering and Environmental Laboratory (INEEL), is conducting a systematic Nuclear Dynamics Consequence Analysis (NDCA) of the disposal of SNFs in an underground geologic repository sited in unsaturated tuff. This analysis is intended to provide interim guidance to the DOE for the management of the SNF while they prepare for final compliance evaluation. This report presents results from a Nuclear Dynamics Consequence Analysis (NDCA) that examined the potential consequences and risks of criticality during the long-term disposal of spent nuclear fuel owned by DOE-EM. This analysis investigated the potential of post-closure criticality, the consequences of a criticality excursion, and the probability frequency for post-closure criticality. The results of the NDCA are intended to provide the DOE-EM with a technical basis for measuring risk which can be used for screening arguments to eliminate post-closure criticality FEPs (features, events and processes) from consideration in the compliance assessment because of either low probability or low consequences. This report is composed of an executive summary (Volume 1), the methodology and results of the NDCA (Volume 2), and the applicable appendices (Volume 3).

Taylor, L.L.; Wilson, J.R. (INEEL); Sanchez, L.C.; Aguilar, R.; Trellue, H.R.; Cochrane, K. (SNL); Rath, J.S. (New Mexico Engineering Research Institute)

1998-10-01T23:59:59.000Z

417

Performance Assessment Analyses Unique to Department of Energy Spent Nuclear Fuel  

SciTech Connect (OSTI)

This paper describes the iterative process of grouping and performance assessment that has led to the current grouping of the U.S. Department of Energy (DOE) spent nuclear fuel (SNF). The unique sensitivity analyses that form the basis for incorporating DOE fuel into the total system performance assessment (TSPA) base case model are described. In addition, the chemistry that results from dissolution of DOE fuel and high level waste (HLW) glass in a failed co-disposal package, and the effects of disposal of selected DOE SNF in high integrity cans are presented.

Loo, Henry Hung Yiu; Duguid, J. O.

2000-06-01T23:59:59.000Z

418

Ab Initio Enhanced calphad Modeling of Actinide-Rich Nuclear Fuels  

SciTech Connect (OSTI)

The process of fuel recycling is central to the Advanced Fuel Cycle Initiative (AFCI), where plutonium and the minor actinides (MA) Am, Np, and Cm are extracted from spent fuel and fabricated into new fuel for a fast reactor. Metallic alloys of U-Pu-Zr-MA are leading candidates for fast reactor fuels and are the current basis for fast spectrum metal fuels in a fully recycled closed fuel cycle. Safe and optimal use of these fuels will require knowledge of their multicomponent phase stability and thermodynamics (Gibbs free energies). In additional to their use as nuclear fuels, U-Pu-Zr-MA contain elements and alloy phases that pose fundamental questions about electronic structure and energetics at the forefront of modern many-body electron theory. This project will validate state-of-the-art electronic structure approaches for these alloys and use the resulting energetics to model U-Pu-Zr-MA phase stability. In order to keep the work scope practical, researchers will focus on only U-Pu-Zr-{Np,Am}, leaving Cm for later study. The overall objectives of this project are to: Provide a thermodynamic model for U-Pu-Zr-MA for improving and controlling reactor fuels; and, Develop and validate an ab initio approach for predicting actinide alloy energetics for thermodynamic modeling.

Morgan, Dane [Univ. of Wisconsin, Madison, WI (United States); Yang, Yong Austin [Univ. of Wisconsin, Madison, WI (United States)

2013-10-28T23:59:59.000Z

419

Characterization program management plan for Hanford K basin spent nuclear fuel  

SciTech Connect (OSTI)

The program management plan for characterization of the K Basin spent nuclear fuel was revised to incorporate corrective actions in response to SNF Project QA surveillance 1K-FY-99-060. This revision of the SNF Characterization PMP replaces Duke Eng.

TRIMBLE, D.J.

1999-07-19T23:59:59.000Z

420

Method of removing Pu(IV) polymer from nuclear fuel reclaiming liquid  

DOE Patents [OSTI]

A Pu(IV) polymer not extractable from a nuclear fuel reclaiming solution by conventional processes is electrolytically converted to Pu.sup.3+ and PuO.sub.2.sup.2+ ions which are subsequently converted to Pu.sup.4+ ions extractable by the conventional processes.

Tallent, Othar K. (Oak Ridge, TN); Mailen, James C. (Oak Ridge, TN); Bell, Jimmy T. (Kingston, TN); Arwood, Phillip C. (Harriman, TN)

1982-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear fuels limited" from the National Library of EnergyBeta (NLEBeta).
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421

Energy Frontier Research Center, Center for Materials Science of Nuclear Fuels  

SciTech Connect (OSTI)

This is a document required by Basic Energy Sciences as part of a mid-term review, in the third year of the five-year award period and is intended to provide a critical assessment of the Center for Materials Science of Nuclear Fuels (strategic vision, scientific plans and progress, and technical accomplishments).

Todd R. Allen

2011-12-01T23:59:59.000Z

422

Integrated data base report - 1994: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics  

SciTech Connect (OSTI)

The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel and commercial and U.S. government-owned radioactive wastes. Except for transuranic wastes, inventories of these materials are reported as of December 31, 1994. Transuranic waste inventories are reported as of December 31, 1993. All spent nuclear fuel and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions.

NONE

1995-09-01T23:59:59.000Z

423

Spent nuclear fuel integrity during dry storage - performance tests and demonstrations  

SciTech Connect (OSTI)

This report summarizes the results of fuel integrity surveillance determined from gas sampling during