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Note: This page contains sample records for the topic "fuel cycle reactors" from the National Library of EnergyBeta (NLEBeta).
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We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


1

Rethinking the light water reactor fuel cycle  

E-Print Network (OSTI)

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

Shwageraus, Evgeni, 1973-

2004-01-01T23:59:59.000Z

2

Fuel cycle problems in fusion reactors  

SciTech Connect

Fuel cycle problems of fusion reactors evolve around the breeding, recovery, containment, and recycling of tritium. These processes are described, and their implications and alternatives are discussed. Technically, fuel cycle problems are solvable; economically, their feasibility is not yet known. (auth)

Hickman, R.G.

1976-01-13T23:59:59.000Z

3

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

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

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

4

Updated Uranium Fuel Cycle Environmental Impacts for Advanced Reactor Designs  

Science Conference Proceedings (OSTI)

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

Nitschke, R.

2004-10-03T23:59:59.000Z

5

Preparations for the Integral Fast Reactor fuel cycle demonstration  

Science Conference Proceedings (OSTI)

Modifications to the Hot Fuel Examination Facility-South (HFEF/S) have been in progress since mid-1988 to ready the facility for demonstration of the unique Integral Fast Reactor (IFR) pyroprocess fuel cycle. This paper updates the last report on this subject to the American Nuclear Society and describes the progress made in the modifications to the facility and in fabrication of the new process equipment. The IFR is a breeder reactor, which is central to the capability of any reactor concept to contribute to mitigation of environmental impacts of fossil fuel combustion. As a fast breeder, fuel of course must be recycled in order to have any chance of an economical fuel cycle. The pyroprocess fuel cycle, relying on a metal alloy reactor fuel rather than oxide, has the potential to be economical even at small-scale deployment. Establishing this quantitatively is one important goal of the IFR fuel cycle demonstration.

Lineberry, M.J.; Phipps, R.D.

1989-01-01T23:59:59.000Z

6

Reactor Physics and Fuel Cycle Analysis - Nuclear Engineering Division  

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

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

7

PROTEUS - Simulation Toolset for Reactor Physics and Fuel Cycle Analysis  

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

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

8

Fuel cycle design and analysis of SABR: subrcritical advanced burner reactor.  

E-Print Network (OSTI)

??Various fuel cycles for a sodium-cooled, subcritical, fast reactor with a fusion neutron source for the transmutation of light water reactor spent fuel have been… (more)

Sommer, Christopher

2008-01-01T23:59:59.000Z

9

Software: Reactor Physics and Fuel Cycle Analysis - Nuclear Engineering  

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

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

10

Strengthening the nuclear-reactor fuel cycle against proliferation  

SciTech Connect

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

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

1992-12-31T23:59:59.000Z

11

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

Science Conference Proceedings (OSTI)

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

D. E. Shropshire

2009-01-01T23:59:59.000Z

12

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

Science Conference Proceedings (OSTI)

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

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

13

Performance and fuel cycle cost study of the R2 reactor with HEU and LEU fuels  

SciTech Connect

A systematic study of the experiment performance and fuel cycle costs of the 50 MW R2 reactor operated by Studsvik Energiteknik AB has been performed using the current R2 HEU fuel, a variety of LEU fuel element designs, and two core-box/reflector configurations. The results include the relative performance of both in-core and ex-core experiments, control rod worths, and relative annual fuel cycle costs.

Pond, R.B.; Freese, K.E.; Matos, J.E.

1984-01-01T23:59:59.000Z

14

Economic Analyiss of "Symbiotic" Light Water Reactor/Fast Burner Reactor Fuel Cycles Proposed as Part of the U.S. Advanced Fuel Cycle Initiative (AFCI)  

Science Conference Proceedings (OSTI)

A spreadsheet-based 'static equilibrium' economic analysis was performed for three nuclear fuel cycle scenarios, each designed for 100 GWe-years of electrical generation annually: (1) a 'once-through' fuel cycle based on 100% LWRs fueled by standard UO2 fuel assemblies with all used fuel destined for geologic repository emplacement, (2) a 'single-tier recycle' scenario involving multiple fast burner reactors (37% of generation) accepting actinides (Pu,Np,Am,Cm) from the reprocessing of used fuel from the uranium-fueled LWR fleet (63% of generation), and (3) a 'two-tier' 'thermal+fast' recycle scenario where co-extracted U,Pu from the reprocessing of used fuel from the uranium-fueled part of the LWR fleet (66% of generation) is recycled once as full-core LWR MOX fuel (8% of generation), with the LWR MOX used fuel being reprocessed and all actinide products from both UO2 and MOX used fuel reprocessing being introduced into the closed fast burner reactor (26% of generation) fuel cycle. The latter two 'closed' fuel cycles, which involve symbiotic use of both thermal and fast reactors, have the advantages of lower natural uranium requirements per kilowatt-hour generated and less geologic repository space per kilowatt-hour as compared to the 'once-through' cycle. The overall fuel cycle cost in terms of $ per megawatt-hr of generation, however, for the closed cycles is 15% (single tier) to 29% (two-tier) higher than for the once-through cycle, based on 'expected values' from an uncertainty analysis using triangular distributions for the unit costs for each required step of the fuel cycle. (The fuel cycle cost does not include the levelized reactor life cycle costs.) Since fuel cycle costs are a relatively small percentage (10 to 20%) of the overall busbar cost (LUEC or 'levelized unit electricity cost') of nuclear power generation, this fuel cycle cost increase should not have a highly deleterious effect on the competitiveness of nuclear power. If the reactor life cycle costs are included in the analysis, with the fast reactors having a higher $/kw(e) capital cost than the LWRs, the overall busbar generation cost ($/MWh) for the closed cycles is approximately 12% higher than for the all-LWR once-through fuel cycle case, again based on the expected values from an uncertainty analysis. It should be noted that such a percentage increase in the cost of nuclear power is much smaller than that expected for fossil fuel electricity generation if CO2 is costed via a carbon tax, cap and trade regimes, or carbon capture and sequestration (CCS).

Williams, Kent Alan [ORNL; Shropshire, David E. [Idaho National Laboratory (INL)

2009-01-01T23:59:59.000Z

15

FUEL CYCLE COSTS IN A GRAPHITE MODERATED SLIGHTLY ENRICHED FUSED SALT REACTOR  

SciTech Connect

A fuel cycle economic study has been made for a 315Mwe graphite- moderated slightly enriched fused-salt reactor. Fuel cycle costs of less than 1.5 mills may be possible for such reactors operating on a ten-year cycle even when the fuel is discarded at the end of the cycle. Recovery of the uranium and plutonium at the end of the cycle reduces the fuel cycle costs to approximates 1 mill/kwh. Changes in the waste storage cost, reprocessing cost or salt inventory have a relatively minor effect on fuel cycle costs. (auth)

Guthrie, C.E.

1959-01-01T23:59:59.000Z

16

Dr. Hussein Khalil at Reactor and Fuel Cycle Technologies Subcommittee  

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

Blue Blue ribbon presentation by Dr. Hussein Khalil Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library About Nuclear Energy Nuclear Reactors Designed by Argonne Argonne's Nuclear Science and Technology Legacy Opportunities within NE Division Visit Argonne Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Celebrating the 70th Anniversary of Chicago Pile 1 (CP-1) Argonne OutLoud on Nuclear Energy Argonne Energy Showcase 2012 Highlights Bookmark and Share Blue ribbon presentation by Hussein Khalil Hussein Khalil Dr. Hussein Khalil during the panel discussion Oct. 21, 2010 On October 12 Hussein Khalil, director of Argonne's Nuclear Engineering Division, participated in a Reactor and Fuel Cycle Technologies

17

Effect of reduced enrichment on the fuel cycle for research reactors  

SciTech Connect

The new fuels developed by the RERTR Program and by other international programs for application in research reactors with reduced uranium enrichment (<20% EU) are discussed. It is shown that these fuels, combined with proper fuel-element design and fuel-management strategies, can provide at least the same core residence time as high-enrichment fuels in current use, and can frequently significantly extend it. The effect of enrichment reduction on other components of the research reactor fuel cycle, such as uranium and enrichment requirements, fuel fabrication, fuel shipment, and reprocessing are also briefly discussed with their economic implications. From a systematic comparison of HEU and LEU cores for the same reference research reactor, it is concluded that the new fuels have a potential for reducing the research reactor fuel cycle costs while reducing, at the same time, the uranium enrichment of the fuel.

Travelli, A.

1982-01-01T23:59:59.000Z

18

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

SciTech Connect

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

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

2001-05-15T23:59:59.000Z

19

THORIUM BREEDER REACTOR EVALUATION. PART 1. FUEL YIELD AND FUEL CYCLE COSTS IN FIVE THERMAL BREEDERS  

SciTech Connect

The performances of aqueous-homogeneous (AHBR), molten-salt (MSBR), liquid-bismuth (LBBR), gas-cooled graphite-moderated (GGBR), and deuterium- moderated gas-cooled (DGBR) breeder reactors were evaluated in respect to fuel yield, fuel cycle costs, and development status. A net electrical plant capability of 1000 Mwe was selected, and the fuel and fertile streams were processed continuously on-site. The maximum annual fuel yields were 1.5 mills/ kwhr. The minimum estimated fuel cycle costs were 0.9, 0.6, 1.0, 1.2, and 1.3 mills/kwhr at fuel yields of were 0.9, 0.9, 1.5, 1.5, and 1.3 mills/kwhr. Only the AHBR and the MSBR are capable of achieving fuel yields substantially in excess of 4%/yr, and therefore, in view of the uncertainties in nuclear data and efficiencies of processing methods, only these two can be listed with confidence as being able to satisfy the main criterion of the AEC longrange thorium breeder program, viz. a doubling time of 25 years or less. The development effort required to bring the various concepts to the stage where a prototype station could be designed was estimated to be least for the AHBR, somewhat more for the MSBR, and several times as much for the other systems. The AHBR was judged to rank first in regard to nuclear capability, fuel cycle potential, and status of development. (auth)

Alexander, L.G.; Carter, W.L.; Chapman, R.H.; Kinyon, R.W.; Miller, J.W.; Van Winkle, R.

1961-05-24T23:59:59.000Z

20

Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration  

SciTech Connect

The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations.

Benedict, R.W.; Goff, K.M.

1993-01-01T23:59:59.000Z

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


21

A dynamic fuel cycle analysis for a heterogeneous thorium-DUPIC recycle in CANDU reactors  

SciTech Connect

A heterogeneous thorium fuel recycle scenario in a Canada deuterium uranium (CANDU) reactor has been analyzed by the dynamic analysis method. The thorium recycling is performed through a dry process which has a strong proliferation resistance. In the fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 0%. In this analysis, the spent fuel inventory as well as the amount of plutonium, minor actinides, and fission products of a multiple thorium recycling fuel cycle were estimated and compared to those of the once-through fuel cycle. The analysis results have shown that the heterogeneous thorium fuel cycle can be constructed through the dry process technology. It is also shown that the heterogeneous thorium fuel cycle can reduce the spent fuel inventory and save on the natural uranium resources when compared with the once-through cycle. (authors)

Jeong, C. J.; Park, C. J.; Choi, H. [Korea Atomic Energy Research Inst., P.O. Box 150, Yuseong, Daejeon, 305-600 (Korea, Republic of)

2006-07-01T23:59:59.000Z

22

Progress and status of the Integral Fast Reactor (IFR) fuel cycle development  

Science Conference Proceedings (OSTI)

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. 10 refs.

Till, C.E.; Chang, Y.I.

1991-01-01T23:59:59.000Z

23

General analysis of breed-and-burn reactors and limited-separations fuel cycles  

E-Print Network (OSTI)

A new theoretical framework is introduced, the "neutron excess" concept, which is useful for analyzing breed-and-burn (B&B) reactors and their fuel cycles. Based on this concept, a set of methods has been developed which ...

Petroski, Robert C

2011-01-01T23:59:59.000Z

24

Behavior of actinides in the Integral Fast Reactor fuel cycle  

SciTech Connect

The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors` confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

Courtney, J.C. [Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center; Lineberry, M.J. [Argonne National Lab., Idaho Falls, ID (United States). Technology Development Div.

1994-06-01T23:59:59.000Z

25

Status of the Integral Fast Reactor fuel cycle demonstration and waste management practices  

SciTech Connect

Over the past few years, Argonne National Laboratory has been preparing for the demonstration of the fuel cycle for the Integral Fast Reactor (IFR), an advanced reactor concept that takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety and operations, fuel-cycle economics, environmental protection, and safeguards. The IFR fuel cycle, which will be demonstrated at Argonne-West in Idaho, employs a pyrometallurgical process using molten salts and liquid metals to recover actinides from spent fuel. The required facility modifications and process equipment for the demonstration are nearing completion. Their status and the results from initial fuel fabrication work, including the waste management aspects, are presented. Additionally, estimated compositions of the various process waste streams have been made, and characterization and treatment methods are being developed. The status of advanced waste processing equipment being designed and fabricated is described.

Benedict, R.W.; Goff, K.M.; McFarlane, H.F.

1994-07-01T23:59:59.000Z

26

Nuclear fuel cycle costs  

Science Conference Proceedings (OSTI)

The costs for the back-end of the nuclear fuel cycle, which were developed as part of the Nonproliferation Alternative Systems Assessment Program (NASAP), are presented. Total fuel cycle costs are given for the pressurized water reactor once-through and fuel recycle systems, and for the liquid-metal fast breeder reactor system. These calculations show that fuel cycle costs are a small part of the total power costs. For breeder reactors, fuel cycle costs are about half that of the present once-through system. The total power cost of the breeder reactor system is greater than that of light-water reactor at today's prices for uranium and enrichment.

Burch, W.D.; Haire, M.J.; Rainey, R.H.

1982-02-01T23:59:59.000Z

27

New Tool for Proliferation Resistance Evaluation Applied to Uranium and Thorium Fueled Fast Reactor Fuel Cycles  

E-Print Network (OSTI)

The comparison of nuclear facilities based on their barriers to nuclear material proliferation has remained a difficult endeavor, often requiring expert elicitation for each system under consideration. However, objectively comparing systems using a set of computable metrics to derive a single number representing a system is not, in essence, a nuclear nonproliferation specific problem and significant research has been performed for business models. For instance, Multi-Attribute Utility Analysis (MAUA) methods have been used previously to provide an objective insight of the barriers to proliferation. In this paper, the Proliferation Resistance Analysis and Evaluation Tool for Observed Risk (PRAETOR), a multi-tiered analysis tool based on the multiplicative MAUA method, is presented. It folds sixty three mostly independent metrics over three levels of detail to give an ultimate metric for nonproliferation performance comparison. In order to reduce analysts' bias, the weighting between the various metrics was obtained by surveying a total of thirty three nonproliferation specialists and nonspecialists from fields such as particle physics, international policy, and industrial engineering. The PRAETOR was used to evaluate the Fast Breeder Reactor Fuel Cycle (FBRFC). The results obtained using these weights are compared against a uniform weight approach. Results are presented for five nuclear material diversion scenarios: four examples include a diversion attempt on various components of a PUREX fast reactor cycle and one scenario involves theft from a PUREX facility in a LWR cycle. The FBRFC was evaluated with uranium-plutonium fuel and a second time using thorium-uranium fuel. These diversion scenarios were tested with both uniform and expert weights, with and without safeguards in place. The numerical results corroborate nonproliferation truths and provide insight regarding fast reactor facilities' proliferation resistance in relation to known standards.

Metcalf, Richard R.

2009-05-01T23:59:59.000Z

28

Simulated first operating campaign for the Integral Fast Reactor fuel cycle demonstration  

Science Conference Proceedings (OSTI)

This report discusses the Integral Fast Reactor (IFR) which is an innovative liquid-metal-cooled reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid-metal cooling to offer significant improvements in reactor safety, operation, fuel cycle-economics, environmental protection, and safeguards. Over the next few years, the IFR fuel cycle will be demonstrated at Argonne-West in Idaho. Spent fuel from the Experimental Breeder Reactor II (EBR-II) win be processed in its associated Fuel Cycle Facility (FCF) using a pyrochemical method that employs molten salts and liquid metals in an electrorefining operation. As part of the preparation for the fuel cycle demonstration, a computer code, PYRO, was developed at Argonne to model the electrorefining operation using thermodynamic and empirical data. This code has been used extensively to evaluate various operating strategies for the fuel cycle demonstration. The modeled results from the first operating campaign are presented. This campaign is capable of processing more than enough material to refuel completely the EBR-II core.

Goff, K.M.; Mariani, R.D.; Benedict, R.W.; Park, K.H. [Argonne National Lab., Idaho Falls, ID (United States); Ackerman, J.P. [Argonne National Lab., IL (United States)

1993-09-01T23:59:59.000Z

29

Nuclear Fuel Cycle Cost Comparison Between Once-Through and Plutonium Multi-Recycling in Fast Reactors  

Science Conference Proceedings (OSTI)

This report presents results from a parametric study of equilibrium fuel cycle costs for a closed fuel cycle with multi-recycling of plutonium in fast reactors (FRs) compared to an open, once-through fuel cycle using PWRs. The study examines the impact on fuel cycle costs from changes in the unit costs of uranium, advanced PUREX reprocessing of discharged uranium dioxide (UO2) fuel and fast-reactor mixed-oxide (FR-MOX) fuel, and FR-MOX fuel fabrication. In addition, the impact associated with changes in ...

2010-03-15T23:59:59.000Z

30

Effect of changes in DOE pricing policies for enrichment and reprocessing on research reactor fuel cycle costs  

SciTech Connect

Fuel cycle costs with HEU and LEU fuels for the IAEA generic 10 MW reactor are updated to reflect the change in DOE pricing policy for enrichment services as of October 1985 and the published charges for LEU reprocessing services as of February 1986. The net effects are essentially no change in HEU fuel cycle costs and a reduction of about 8 to 10% in the fuel cycle costs for LEU silicide fuel.

Matos, J.E.; Freese, K.E.

1986-11-03T23:59:59.000Z

31

Proliferation resistance for fast reactors and related fuel cycles: issues and impacts  

Science Conference Proceedings (OSTI)

The prospects for a dramatic growth in nuclear power may depend to a significant degree on the effectiveness of, and the resources devoted to, plans to develop and implement technologies and approaches that strengthen proliferation resistance and nuclear materials accountability. The challenges for fast reactors and related fuel cycles are especially critical. They are being explored in the Generation IV Tnternational Forum (GIF) and the Tnternational Atomic Energy Agency's (IAEA's) International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) initiative, as well as by many states that are looking to these systems for the efficient lise of uranium resources and long-term energy security. How do any proliferation risks they may pose compare to other reactors, both existing and under development, and their fuel cycles? Can they be designed with intrinsic (technological) features to make these systems more proliferation resistant? What roles can extrinsic (institutional) features play in proliferation resistance? What are the anticipated safeguards requirements, and will new technologies and approaches need to be developed? How can safeguards be facilitated by the design process? These and other questions require a rethinking of proliferation resistance and the prospects for new technologies and other intrinsic and extrinsic features being developed that are responsive to specific issues for fast reactors and related fuel cycles and to the broader threat environment in which these systems will have to operate. There are no technologies that can wholly eliminate the risk of proliferation by a determined state, but technology and design can playa role in reducing state threats and perhaps in eliminating non-state threats. There will be a significant role for extrinsic factors, especially the various measures - from safeguards and physical protection to export controls - embodied in the international nuclear nonproliferation regime. This paper will offer an assessment of the issues surrounding, and the prospects for, efforts to develop proliferation resistance for fast reactors and related fuel cycles in the context of a nuclear renaissance. The focus of the analysis is on fast reactors.

Pilat, Joseph F [Los Alamos National Laboratory

2010-01-01T23:59:59.000Z

32

Investigation of Browns Ferry 2 Reactor Cycle 12 Fuel Corrosion Failures, Volume 3: Assessment of Results  

Science Conference Proceedings (OSTI)

Boiling water reactor (BWR) fuel rods from 63 bundles of the Reload 10 GE13 (9x9) design developed leaks during Cycle 12 of the Browns Ferry 2 reactor. Root cause evaluations, including poolside and hot cell examinations were performed. The details of the investigation are documented in a series of reports and presentations. This report compiles significant findings of the overall investigation and assesses these results with respect to the cause of failure. In addition, new laboratory data to support th...

2007-12-05T23:59:59.000Z

33

Enhanced CANDU6: Reactor and fuel cycle options - Natural uranium and beyond  

SciTech Connect

The Enhanced CANDU 6{sup R} (ECo{sup R}) is the updated version of the well established CANDU 6 family of units incorporating improved safety characteristics designed to meet or exceed Generation III nuclear power plant expectations. The EC6 retains the excellent neutron economy and fuel cycle flexibility that are inherent in the CANDU reactor design. The reference design is based on natural uranium fuel, but the EC6 is also able to utilize additional fuel options, including the use of Recovered Uranium (RU) and Thorium based fuels, without requiring major hardware upgrades to the existing control and safety systems. This paper outlines the major changes in the EC6 core design from the existing C6 design that significantly enhance the safety characteristics and operating efficiency of the reactor. The use of RU fuel as a transparent replacement fuel for the standard 37-el NU fuel, and several RU based advanced fuel designs that give significant improvements in fuel burnup and inherent safety characteristics are also discussed in the paper. In addition, the suitability of the EC6 to use MOX and related Pu-based fuels will also be discussed. (authors)

Ovanes, M.; Chan, P. S. W.; Mao, J.; Alderson, N.; Hopwood, J. M. [Candu Energy Inc., 2285 Speakman Drive, Mississauga, ON L5K 1B1 (Canada)

2012-07-01T23:59:59.000Z

34

Development of a Safeguards Approach for a Small Graphite Moderated Reactor and Associated Fuel Cycle Facilities  

E-Print Network (OSTI)

Small graphite-moderated and gas-cooled reactors have been around since the beginning of the atomic age. Though their existence in the past has been associated with nuclear weapons programs, they are capable of being used in civilian power programs. The simpler design constraints associated with this type of reactor would make them ideal for developing nations to bolster their electricity generation and help promote a greater standard of living in those nations. However, the same benefits that make this type of reactor desirable also make it suspicious to the international community as a possible means to shorten that state?s nuclear latency. If a safeguards approach could be developed for a fuel cycle featuring one of these reactors, it would ease the tension surrounding their existence and possibly lead to an increased latency through engineered barriers. The development of this safeguards approach follows a six step procedure. First, the fuel cycle was analyzed for the types of facilities found in it and how nuclear material flows between facilities. The goals of the safeguards system were established next, using the normal IAEA standards for the non-detection and false alarm probabilities. The 5 MWe Reactor was modeled for both plutonium production and maximum power capacity. Each facility was analyzed for material throughput and the processes that occur in each facility were researched. Through those processes, diversion pathways were developed to test the proposed safeguards system. Finally, each facility was divided into material balance areas and a traditional nuclear material accountancy system was set up to meet the established safeguards goals for the facility. The DPRK weapons program is a great example of the type of fuel cycle that is the problem. The three major facilities in the fuel cycle, the Fuel Fabrication Facility, the 5 MWe Reactor, and the Radiochemical Laboratory, can achieve the two goals of safeguards using traditional methods. Each facility can be adequately safeguarded using methods and practices that are relatively inexpensive and can obtain material balance periods close to the timeliness limits set forth by the IAEA. The Fuel Fabrication Facility can be safeguarded at both its current needed capacity and its full design capacity using inexpensive measurements. The material balance period needed for both capacities are reasonable. For the 5 MWe reactor, plutonium production is simulated to be 6.7 kg per year and is on the high side of estimates. The Radiochemical Laboratory can also be safeguarded at its current capacity. In fact, the timeliness goal for the facility dictates what the material balance period must be for the chosen set of detectors which make it very reasonable.

Rauch, Eric B.

2009-05-01T23:59:59.000Z

35

Validation and application of a physics database for fast reactor fuel cycle analysis  

SciTech Connect

An effort has been made to automate the execution of fast reactor fuel cycle analysis, using EBR-II as a demonstration vehicle, and to validate the analysis results for application to the IFR closed fuel cycle demonstration at EBR-II and its fuel cycle facility. This effort has included: (1) the application of the standard ANL depletion codes to perform core-follow analyses for an extensive series of EBR-II runs, (2) incorporation of the EBR-II data into a physics database, (3) development and verification of software to update, maintain and verify the database files, (4) development and validation of fuel cycle models and methodology, (5) development and verification of software which utilizes this physics database to automate the application of the ANL depletion codes, methods and models to perform the core-follow analysis, and (6) validation studies of the ANL depletion codes and of their application in support of anticipated near-term operations in EBR-II and the Fuel Cycle Facility. Results of the validation tests indicate the physics database and associated analysis codes and procedures are adequate to predict required quantities in support of early phases of FCF operations.

McKnight, R.D.; Stillman, J.A.; Toppel, B.J.; Khalil, H.S.

1994-03-01T23:59:59.000Z

36

Investigation of Browns Ferry 2 Reactor Cycle 12 Fuel Corrosion Failures  

Science Conference Proceedings (OSTI)

BWR fuel rods from 63 bundles of the Reload 10 GE13 (9x9) design developed leaks during Cycle 12 of the Browns Ferry 2 reactor. Poolside examination subsequently revealed accelerated cladding corrosion as the apparent failure mechanism (EPRI report 1009729). The failure root cause could not be determined from site examinations alone and thus follow-on hot cell examinations were needed to provide additional information that might help to further understand the failure mechanism, the root cause, and linger...

2006-06-19T23:59:59.000Z

37

Proliferation resistance of the fuel cycle for the Integral Fast Reactor  

SciTech Connect

Argonne National Laboratory has developed an electrorefining pyrochemical process for recovery and recycle of metal fuel discharged from the Integral Fast Reactor (FR). This inherently low decontamination process has an overall decontamination factor of only about 100 for the plutonium metal product. As a result, all of the fuel cycle operations must be conducted in heavily shielded cells containing a high-purity argon atmosphere. The FR fuel cycle possesses high resistance to clandestine diversion or overt, state- supported removal of plutonium for nuclear weapons production because of two main factors: the highly radioactive product, which is also contaminated with heat- and neutron-producing isotopes of plutonium and other actinide elements, and the difficulty of removing material from the FR facility through the limited number of cell transfer locks without detection.

Burris, L.

1993-09-01T23:59:59.000Z

38

Fuel-cycle costs for alternative fuels  

Science Conference Proceedings (OSTI)

This paper compares the fuel cycle cost and fresh fuel requirements for a range of nuclear reactor systems including the present day LWR without fuel recycle, an LWR modified to obtain a higher fuel burnup, an LWR using recycle uranium and plutonium fuel, an LWR using a proliferation resistant /sup 233/U-Th cycle, a heavy water reactor, a couple of HTGRs, a GCFR, and several LMFBRs. These reactor systems were selected from a set of 26 developed for the NASAP study and represent a wide range of fuel cycle requirements.

Rainey, R.H.; Burch, W.D.; Haire, M.J.; Unger, W.E.

1980-01-01T23:59:59.000Z

39

Nuclear Fuel Cycle Cost Comparison Between Once-Through and Plutonium Single-Recycling in Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

Within the context of long-term waste management and sustainable nuclear fuel supply, there continue to be discussions regarding whether the United States should consider recycling of light-water reactor (LWR) spent nuclear fuel (SNF) for the current fleet of U.S. LWRs. This report presents a parametric study of equilibrium fuel cycle costs for an open fuel cycle without plutonium recycling (once-through) and with plutonium recycling (single-recycling using mixed-oxide, or MOX, fuel), assuming an all-pre...

2009-02-25T23:59:59.000Z

40

A strategy for transition from a uranium fueled, open cycle SFR to a transuranic fueled, closed cycle sodium cooled fast reactor  

E-Print Network (OSTI)

Reactors utilizing a highly energetic neutron spectrum, often termed fast reactors, offer large fuel utilization improvements over the thermal reactors currently used for nuclear energy generation. Conventional fast reactor ...

Richard, Joshua (Joshua Glenn)

2012-01-01T23:59:59.000Z

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


41

A 48-month extended fuel cycle for the B and W mPower{sup TM} small modular nuclear reactor  

Science Conference Proceedings (OSTI)

The B and W mPower{sup TM} reactor is a small, rail-shippable pressurized water reactor (PWR) with an integral once-through steam generator and an electric power output of 150 MW, which is intended to replace aging fossil power plants of similar output. The core is composed of 69 reduced-height, but otherwise standard, PWR assemblies with the familiar 17 x 17 fuel rod array on a 21.5 cm inter-assembly pitch. The B and W mPower core design and cycle management plan, which were performed using the Studsvik core design code suite, follow the pattern of a typical nuclear reactor fuel cycle design and analysis performed by most nuclear fuel management organizations, such as fuel vendors and utilities. However, B and W is offering a core loading and cycle management plan for four years of continuous power operations without refueling and without the hurdles of chemical shim. (authors)

Erighin, M. A. [Babcock and Wilcox Company, 109 Ramsey Place, Lynchburg, VA 24502 (United States)

2012-07-01T23:59:59.000Z

42

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

Science Conference Proceedings (OSTI)

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

Schneider, K.J.

1982-09-01T23:59:59.000Z

43

Fuel cycles and envisioned roles of fast neutron reactors and hybrids  

Science Conference Proceedings (OSTI)

Future innovative nuclear fuel cycles will require insuring sustainability in terms of safe operation, optimal use of resources, radioactive waste minimization and reduced risk of proliferation. The present paper introduces some basic notions and fundamental fuel cycle strategies. The simulation approach needed to evaluate the impact of the different fuel cycle alternatives will also be shortly discussed.

Salvatores, Massimo [CEA-Cadarache, DEN-Dir, Bat. 101, St-Paul-Lez-Durance 13108 (France)

2012-06-19T23:59:59.000Z

44

Nuclear fuel cycle information workshop  

SciTech Connect

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

1983-01-01T23:59:59.000Z

45

Application of the thorium fuel cycle  

SciTech Connect

An economic analysis of the application of the thorium fuel cycle to thermal reactors is presented. (JWR)

Kasten, P.R.; Tobias, M.L.

1975-01-01T23:59:59.000Z

46

An extended conventional fuel cycle for the B and W mPower{sup TM} small modular nuclear reactor  

SciTech Connect

The B and W mPower{sup TM} reactor is a small pressurized water reactor (PWR) with an integral once-through steam generator and a thermal output of about 500 MW; it is intended to replace aging fossil power plants of similar output. The core is composed of 69 reduced-height PWR assemblies with the familiar 17 x 17 fuel rod array. The Babcock and Wilcox Company (B and W) is offering a core loading and cycle management plan for a four-year cycle based on its presumed attractiveness to potential customers. This option is a once-through fuel cycle in which the entire core is discharged and replaced after four years. In addition, a conventional fuel utilization strategy, employing a periodic partial reload and shuffle, was developed as an alternative to the four-year once-through fuel cycle. This study, which was performed using the Studsvik core design code suite, is a typical multi-cycle projection analysis of the type performed by most fuel management organizations such as fuel vendors and utilities. In the industry, the results of such projections are used by the financial arms of these organizations to assist in making long-term decisions. In the case of the B and W mPower reactor, this analysis demonstrates flexibility for customers who consider the once-through fuel cycle unacceptable from a fuel utilization standpoint. As expected, when compared to the once-through concept, reloads of the B and W mPower reactor will achieve higher batch average discharge exposure, will have adequate shut-down margin, and will have a relatively flat hot excess reactivity trend at the expense of slightly increased peaking. (authors)

Scarangella, M. J. [Babcock and Wilcox Company, 109 Ramsey Place, Lynchburg, VA 24502 (United States)

2012-07-01T23:59:59.000Z

47

Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input  

SciTech Connect

In this study a feasibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850 deg. C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticality was obtained for this reactor.

Meriyanti; Su'ud, Zaki; Rijal, K. [Nuclear Physics and Biophysics Research Division, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Zuhair; Ferhat, A. [National Nuclear Energ Agency of Indonesia (BATAN) (Indonesia); Sekimoto, H. [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology (Japan)

2010-06-22T23:59:59.000Z

48

The fuel cycle economics of improved uranium utilization in light water reactors  

E-Print Network (OSTI)

A simple fuel cycle cost model has been formulated, tested satisfactorily (within better than 3% for a wide range of cases)

Abbaspour, Ali Tehrani

49

Fuel Cycle Optimization of a Helium-Cooled, Sub-Critical, Fast Transmutation of Waste Reactor with a Fusion Neutron Source.  

E-Print Network (OSTI)

??Possible fuel cycle scenarios for a helium-cooled, sub-critical, fast reactor with a fusion neutron source for the transmutation of spent nuclear fuel have been analyzed.… (more)

Maddox, James Warren

2006-01-01T23:59:59.000Z

50

LEU fuel cycle analyses for the Belgian BR2 Research Reactor  

SciTech Connect

Equilibrium fuel cycle characteristics were calculated for reference HEU and two proposed LEU fuel cycles using an 11-group diffusion-theory neutron flux solution in hexagonal-Z geometry. The diffusion theory model was benchmarked with a detailed Monte Carlo core model. The two proposed LEU fuel designs increased the {sup 235}U loading 20% and the fuel meat volume 51%. The first LEU design used {sup 10}B as a burnable absorber. Either proposed LEU fuel element would provide equilibrium fuel cycle characteristics similar to those of the HEU fuel cycle. Irradiation rates of Co control followers and Ir disks in the center of the core were reduced 6 {plus minus} 1% in the LEU equilibrium core compared to reference HEU core. 11 refs., 4 figs., 5 tabs.

Deen, J.R.; Snelgrove, J.L.

1988-01-01T23:59:59.000Z

51

PREPARING THE HIGH FLUX ISOTOPE REACTOR FOR CONVERSION TO LOW ENRICHED URANIUM FUEL ? EXTENDING CYCLE BURNUP  

Science Conference Proceedings (OSTI)

Reactor performance studies have been completed for conceptual plate designs and show that maintaining reactor performance while converting HFIR from high enriched to low enriched uranium (20 wt % 235U) fuel requires extending the end-of-life burnup value for HFIR fuel from the current nominal value of 2200 MWD to 2600 MWD. The current fuel fabrication procedure is discussed and changes that would be required to this procedure are identified. Design and safety related analyses that are required for the certification of a new fuel are identified. Qualification tests and comments regarding the regulatory approval process are provided along with a conceptual schedule.

Primm, Trent [ORNL; Chandler, David [ORNL

2009-01-01T23:59:59.000Z

52

Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts  

Science Conference Proceedings (OSTI)

The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather than graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in Pressurized Water Reactor (PWR) assemblies. In addition to consideration of this 'naive' use of TRISO fuel in LWRs, several refined options are briefly examined and others are identified for further consideration including the use of advanced, high density fuel forms and larger kernel diameters and TRISO packing fractions. The combination of 800 {micro}m diameter kernels of 20% enriched UN and 50% TRISO packing fraction yielded reactivity sufficient to achieve comparable burnup to present-day PWR fuel.

R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal O. Pasamehmetoglu

2012-04-01T23:59:59.000Z

53

FUEL CYCLE PROGRAM, A BOILING WATER REACTOR RESEARCH DEVELOPMENT PROGRAM. First Summary Report for March 1959-July 1960  

SciTech Connect

The Fuel Cycle Development Program is a basic development program for boiling and other water technology. It covers the areas of oxide fuel fabrication. irradiation. and examination; the physics of water-moderated reactore; and boiling-water heat transfer and stability. Schedules for the fuel- cycle program were examined. and it was concluded that portions of the Task A program should be conducted during the period May to Dec. 1959 in order to keep costs of the work as low as possible and to allow initiation of the fuel-cycle program at the earliest possible date after the Vallecitos BWR was returned to service. The basis for the scheduling of the work is discussed. and a chronological summary describing the content of the work is given. Technical progress is outlined and details are summarized. Subsequent reports issued monthly and quarterly will summarize the progress of the prognam. (W.D.M.)

Cook, W.H.

1961-10-31T23:59:59.000Z

54

Preliminary LEU fuel cycle analyses for the Belgian BR2 reactor  

SciTech Connect

Fuel cycle calculations have been performed with reference HEU fuel and LEU fuel using Cd wires or boron as burnable absorbers. The /sup 235/U content in the LEU element has increased 20% to 480g compared to the reference HEU element. The number of fuel plates has remained unchanged while the fuel meat thickness has increased to 0.76 mm from 0.51 mm. The LEU meat density is 5.1 Mg U/m/sup 3/. The reference fuel cycle was a 31 element core operating at 56 MW with a 19.8 day cycle length and eight fresh elements loaded per cycle. Comparable fuel cycle characteristics can be achieved using the proposed LEU fuel element with either Cd wires or boron burnable absorbers. The neutron flux for E/sub n/ > 1 eV changes very little (<5%) in LEU relative to HEU cores. Thermal flux reductions are 5 to 10% in non-fueled positions, and 20 to 30% in fuel elements.

Deen, J.R.; Snelgrove, J.L.

1986-01-01T23:59:59.000Z

55

Management of transuranics using the Integral Fast Reactor (IFR) fuel cycle  

Science Conference Proceedings (OSTI)

The 50 years of activities following the discovery of self-sustaining fission chains have given rise to a buildup of roughly 900 tons of manmade transuranics. Of the total, about 260 tons of Pu{sup 239} were generated for use in weapons while the remainder were generated as a byproduct of electrical power produced worldwide by the commercial thermal nuclear power industry. What is to be done with these actinides? The options for disposition include interminable storage, burial, or recycle for use. The pros and cons of each option are being vigorously debated regarding the impact upon the issues of human and ecological risk -- both current and future; weapons proliferation potential -- both current and future; and total life cycle benefits and costs. As to the options for utilization, commercial uses for actinides (uranium and transuranics) are of limited diversity. The actinides have in the past and will in the future find application in large scale mostly by virtue of their ability to release energy through fission, and here their utility is unmatched -- whether the application be in commercial electricity generation or in armaments. The integral Fast Reactor (IFR) fuel cycle offers a number of features for management of the current and future burden of manmade transuranic materials and for capturing the energy content of the U{sup 238}. These features are discussed here.

Wade, D.C.

1994-01-01T23:59:59.000Z

56

Compatibility Analysis on Existing Reactivity Devices in CANDU 6 Reactors for DUPIC Fuel Cycle  

Science Conference Proceedings (OSTI)

The performance of reactivity devices for a Canada deuterium uranium (CANDU) 6 reactor loaded with Direct Use of Spent Pressurized Water Reactor Fuel In CANDU reactors (DUPIC) fuel is assessed. The reactivity devices studied are the zone controller units, the adjuster rods, and the mechanical control absorbers. For the zone controller system, the bulk reactivity control, spatial power control, and damping capability for spatial oscillation are investigated. For the adjusters, the xenon override, restart after a poison-out, shim operation, and power step-back capabilities are confirmed. The mechanical control absorber is assessed for the function of compensating temperature reactivity feedback following a power reduction. This study shows that the current reactivity device system of a CANDU 6 reactor is compatible with DUPIC fuel for normal and transient operations.

Jeong, Chang-Joon; Choi, Hangbok [Korea Atomic Energy Research Institute (Korea, Republic of)

2000-03-15T23:59:59.000Z

57

Renovation of CPF (Chemical Processing Facility) for Development of Advanced Fast Reactor Fuel Cycle System  

Science Conference Proceedings (OSTI)

CPF (Chemical Processing Facility) was constructed at Nuclear Fuel Cycle Engineering Laboratories of JAEA (Japan Atomic Energy Agency) in 1980 as a basic research field where spent fuel pins from fast reactor (FR) and high level liquid waste can be dealt with. The renovation consists of remodeling of the CA-3 cell and the laboratory A, installation of globe boxes, hoods and analytical equipments to the laboratory C and the analytical laboratory. Also maintenance equipments in the CA-5 cell which had been out of order were repaired. The CA-3 cell is the main cell in which important equipments such as a dissolver, a clarifier and extractors are installed for carrying out the hot test using the irradiated FR fuel. Since the CPF had specialized originally in the research function for the Purex process, it was desired to execute the research and development of such new, various reprocessing processes. Formerly, equipments were arranged in wide space and connected with not only each other but also with utility supply system mainly by fixed stainless steel pipes. It caused shortage of operation space in flexibility for basic experimental study. Old equipments in the CA-3 cell including vessels and pipes were removed after successful decontamination, and new equipments were installed conformably to the new design. For the purpose of easy installation and rearranging the experimental equipments, equipments are basically connected by flexible pipes. Since dissolver is able to be easily replaced, various dissolution experiments is conducted. Insoluble residue generated by dissolution of spent fuel is clarified by centrifugal. This small apparatus is effective to space-saving. Mini mixer settlers or centrifugal contactors are put on to the prescribed limited space in front of the backside wall. Fresh reagents such as solvent, scrubbing and stripping solution are continuously fed from the laboratory A to the extractor by the reagent supply system with semi-automatic observation system. The in-cell crane in CA-5 was renovated to increase driving efficiency. At the renovation for the in-cell crane, full scale mockup test and 3D simulation test had been executed in advance. After the renovation, hot tests in the CPF had been resumed from JFY 2002. New equipments such as dissolver, extractor, electrolytic device, etc. were installed in CA-3 conformably to the new design laid out in order to ensure the function and space. Glove boxes in the analysis laboratory were renewed in order to let it have flexibility from the viewpoint of conducting basic experiments (ex. U crystallization). Glove boxes and hoods were newly installed in the laboratory A for basic research and analysis, especially on MA chemistries. One laboratory (the laboratory C) was established to research about dry reprocessing. The renovation of the CPF has been executed in order to contribute to the development on the advanced fast reactor fuel cycle system, which will give us many sort of technical subject and experimental theme to be solved in the 2. Generation of the CPF.

Shinichi Aose; Takafumi Kitajima; Kouji Ogasawara; Kazunori Nomura; Shigehiko Miyachi; Yoshiaki Ichige; Tadahiro Shinozaki; Shinichi Ohuchi [Japan Atomic Energy Agency:4-33, Tokai-mura, Naka-gun, Ibaraki pref, 319-1194 (Japan)

2008-01-15T23:59:59.000Z

58

Practical introduction of thorium fuel cycles  

SciTech Connect

The pracitcal introduction of throrium fuel cycles implies that thorium fuel cycles compete economically with uranium fuel cycles in economic nuclear power plants. In this study the reactor types under consideration are light water reactors (LWRs), heavy water reactors (HWRs), high-temperature gas-cooled reactors (HTGRs), and fast breeder reactors (FBRs). On the basis that once-through fuel cycles will be used almost exclusively for the next 20 or 25 years, introduction of economic thorium fuel cycles appears best accomplished by commercial introduction of HTGRs. As the price of natural uranium increases, along with commercialization of fuel recycle, there will be increasing incentive to utilize thorium fuel cycles in heavy water reactors and light water reactors as well as in HTGRs. After FBRs and fuel recycle are commercialized, use of thorium fuel cycles in the blanket of FBRs appears advantageous when fast breeder reactors and thermal reactors operate in a symbiosis mode (i.e., where /sup 233/U bred in the blanket of a fast breeder reactor is utilized as fissile fuel in thermal converter reactors).

Kasten, P.R.

1982-01-01T23:59:59.000Z

59

Summary of the radiological assessment of the fuel cycle for a thorium-uranium carbide-fueled fast breeder reactor  

Science Conference Proceedings (OSTI)

A large fraction of the potential fuel for nuclear power reactors employing fissionable materials exists as ores of thorium. In addition, certain characteristics of a fuel system based on breeding of the fissionable isotope {sup 233}U from thorium offer the possibility of a greater resistance to the diversion of fissionable material for the fabrication of nuclear weapons. This report consolidates into a single source the principal content of two previous reports which assess the radiological environmental impact of mining and milling of thorium ore and of the reprocessing and refabrication of spent FBR thorium-uranium carbide fuel.

Tennery, V.J.; Bomar, E.S.; Bond, W.D.; Meyer, H.R.; Morse, L.E.; Till, J.E.; Yalcintas, M.G.

1980-01-01T23:59:59.000Z

60

Fuel Cycle Subcommittee  

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

Report to NEAC Report to NEAC Fuel Cycle Subcommittee Meeting of April 23, 2013 Washington D.C. June 13, 2013 Burton Richter (Chair), Margaret Chu, Darleane Hoffman, Raymond Juzaitis, Sekazi K Mtingwa, Ronald P Omberg, Joy L Rempe, Dominique Warin 2 I Introduction and Summary The Fuel Cycle Subcommittee of NEAC met in Washington on April 23, 2013. The meeting focused on issues relating to the NE advanced reactor program (sections II, III, and IV), and on storage and transportation issues (section V) related to a possible interim storage program that is the first step in moving toward a new permanent repository as recommended by the Blue Ribbon Commission (BRC) and discussed in the recent response by DOE to Congress on the BRC report 1 . The agenda is given in

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


61

Nuclear 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

62

Fuel cycle cost study with HEU and LEU fuels  

SciTech Connect

Fuel cycle costs are compared for a range of /sup 235/U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors.

Matos, J.E.; Freese, K.E.

1984-01-01T23:59:59.000Z

63

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

Science Conference Proceedings (OSTI)

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

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

2011-12-15T23:59:59.000Z

64

The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input  

Science Conference Proceedings (OSTI)

A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of i{sup th} region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.

Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal,; Monado, Fiber; Sekimoto, Hiroshi [Department of Physics Bandung Institute of Technology Jl. Ganesha 10, Bandung 40134, Physics Department, Sriwijaya University, Kampus Indralaya, Ogan Ilir, Sumatera Selatan (Indonesia); Department of Physics Bandung Institute of Technology Jl. Ganesha 10, Bandung 40134 (Indonesia); Department of Physics Bandung Institute of Technology Jl. Ganesha 10, Bandung 40134, Physics Department, Sriwijaya University, Kampus Indralaya, Ogan Ilir, Sumatera Selatan (Indonesia); Reserach of Laboratory for Nuclear Reactors, Tokyo Institute of Technology O-okayama, Meguro-ku, Tokyo 152 (Japan)

2012-06-06T23:59:59.000Z

65

Nuclear Fuel Cycle Integrated System Analysis  

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

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

66

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

SciTech Connect

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

Not Available

1980-06-01T23:59:59.000Z

67

The closed fuel cycle  

Science Conference Proceedings (OSTI)

Available in abstract form only. Full text of publication follows: The fast growth of the world's economy coupled with the need for optimizing use of natural resources, for energy security and for climate change mitigation make energy supply one of the 21. century most daring challenges. The high reliability and efficiency of nuclear energy, its competitiveness in an energy market undergoing a new oil shock are as many factors in favor of the 'renaissance' of this greenhouse gas free energy. Over 160,000 tHM of LWR1 and AGR2 Used Nuclear Fuel (UNF) have already been unloaded from the reactor cores corresponding to 7,000 tons discharged per year worldwide. By 2030, this amount could exceed 400,000 tHM and annual unloading 14,000 tHM/year. AREVA believes that closing the nuclear fuel cycle through the treatment and recycling of Used Nuclear Fuel sustains the worldwide nuclear power expansion. It is an economically sound and environmentally responsible choice, based on the preservation of natural resources through the recycling of used fuel. It furthermore provides a safe and secure management of wastes while significantly minimizing the burden left to future generations. (authors)

Froment, Antoine; Gillet, Philippe [AREVA NC (France)

2007-07-01T23:59:59.000Z

68

Effect of Highly Enriched/Highly Burnt UO2 Fuels on Fuel Cycle Costs, Radiotoxicity, and Nuclear Design Parameters  

Science Conference Proceedings (OSTI)

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

Robert Gregg; Andrew Worrall

69

Nuclear-fuel-cycle costs. Consolidated Fuel-Reprocessing Program  

Science Conference Proceedings (OSTI)

The costs for the back-end of the nuclear fuel cycle, which were developed as part of the Nonproliferation Alternative Systems Assessment Program (NASAP), are presented. Total fuel-cycle costs are given for the pressurized-water reactor once-through and fuel-recycle systems, and for the liquid-metal fast-breeder-reactor system. These calculations show that fuel-cycle costs are a small part of the total power costs. For breeder reactors, fuel-cycle costs are about half that of the present once-through system. The total power cost of the breeder-reactor system is greater than that of light-water reactor at today's prices for uranium and enrichment.

Burch, W.D.; Haire, M.J.; Rainey, R.H.

1981-01-01T23:59:59.000Z

70

Fuel cycle cost, reactor physics and fuel manufacturing considerations for Erbia-bearing PWR fuel with > 5 wt% U-235 content  

Science Conference Proceedings (OSTI)

The efforts to reduce fuel cycle cost have driven LWR fuel close to the licensed limit in fuel fissile content, 5.0 wt% U-235 enrichment, and the acceptable duty on current Zr-based cladding. An increase in the fuel enrichment beyond the 5 wt% limit, while certainly possible, entails costly investment in infrastructure and licensing. As a possible way to offset some of these costs, the addition of small amounts of Erbia to the UO{sub 2} powder with >5 wt% U-235 has been proposed, so that its initial reactivity is reduced to that of licensed fuel and most modifications to the existing facilities and equipment could be avoided. This paper discusses the potentialities of such a fuel on the US market from a vendor's perspective. An analysis of the in-core behavior and fuel cycle performance of a typical 4-loop PWR with 18 and 24-month operating cycles has been conducted, with the aim of quantifying the potential economic advantage and other operational benefits of this concept. Subsequently, the implications on fuel manufacturing and storage are discussed. While this concept has certainly good potential, a compelling case for its short-term introduction as PWR fuel for the US market could not be determined. (authors)

Franceschini, F.; Lahoda, E. J.; Kucukboyaci, V. N. [Westinghouse Electric Co. LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01T23:59:59.000Z

71

Fuel-cycle cost comparisons with oxide and silicide fuels  

SciTech Connect

This paper addresses fuel cycle cost comparisons for a generic 10 MW reactor with HEU aluminide fuel and with LEU oxide and silicide fuels in several fuel element geometries. The intention of this study is to provide a consistent assessment of various design options from a cost point of view. Fuel cycle cost benefits could result if a number of reactors were to utilize fuel elements with the same number or different numbers of the same standard fuel plate. Data are presented to quantify these potential cost benefits. This analysis shows that there are a number of fuel element designs using LEU oxide or silicide fuels that have either the same or lower total fuel cycle costs than the HEU design. Use of these fuels with the uranium densities considered requires that they are successfully demonstrated and licensed.

Matos, J.E.; Freese, K.E.

1982-01-01T23:59:59.000Z

72

VISION: Verifiable Fuel Cycle Simulation Model  

Science Conference Proceedings (OSTI)

The nuclear fuel cycle is a very complex system that includes considerable dynamic complexity as well as detail complexity. In the nuclear power realm, there are experts and considerable research and development in nuclear fuel development, separations technology, reactor physics and waste management. What is lacking is an overall understanding of the entire nuclear fuel cycle and how the deployment of new fuel cycle technologies affects the overall performance of the fuel cycle. The Advanced Fuel Cycle Initiative’s systems analysis group is developing a dynamic simulation model, VISION, to capture the relationships, timing and delays in and among the fuel cycle components to help develop an understanding of how the overall fuel cycle works and can transition as technologies are changed. This paper is an overview of the philosophy and development strategy behind VISION. The paper includes some descriptions of the model and some examples of how to use VISION.

Jacob J. Jacobson; Abdellatif M. Yacout; Gretchen E. Matthern; Steven J. Piet; David E. Shropshire

2009-04-01T23:59:59.000Z

73

Nuclear Fuel Cycle | Department of Energy  

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

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

74

NUCLEAR REACTOR FUEL SYSTEMS  

DOE Patents (OSTI)

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

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

1959-09-15T23:59:59.000Z

75

Status of IFR fuel cycle demonstration  

SciTech Connect

The next major step in Argonne`s Integral Fast Reactor (IFR) Program is demonstration of the pyroprocess fuel cycle, in conjunction with continued operation of EBR-II. The Fuel Cycle Facility (FCF) is being readied for this mission. This paper will address the status of facility systems and process equipment, the initial startup experience, and plans for the demonstration program.

Lineberry, M.J.; Phipps, R.D.; McFarlane, H.F.

1993-09-01T23:59:59.000Z

76

ASSESSMENT OF POSSIBLE CYCLE LENGTHS FOR FULLY-CERAMIC MICRO-ENCAPSULATED FUEL-BASED LIGHT WATER REACTOR CONCEPTS  

Science Conference Proceedings (OSTI)

The use of TRISO-particle-based dispersion fuel within SiC matrix and cladding materials has the potential to allow the design of extremely safe LWRs with failure-proof fuel. This paper examines the feasibility of LWR-like cycle length for such a low enriched uranium fuel with the imposed constraint of strictly retaining the original geometry of the fuel pins and assemblies. The motivation for retaining the original geometry is to provide the ability to incorporate the fuel 'as-is' into existing LWRs while retaining their thermal-hydraulic characteristics. The feasibility of using this fuel is assessed by looking at cycle lengths and fuel failure rates. Other considerations (e.g., safety parameters, etc.) were not considered at this stage of the study. The study includes the examination of different TRISO kernel diameters without changing the coating layer thicknesses. The study shows that a naive use of UO{sub 2} results in cycle lengths too short to be practical for existing LWR designs and operational demands. Increasing fissile inventory within the fuel compacts shows that acceptable cycle lengths can be achieved. In this study, starting with the recognized highest packing fraction practically achievable (44%), higher enrichment, larger fuel kernel sizes, and the use of higher density fuels have been evaluated. The models demonstrate cycle lengths comparable to those of ordinary LWRs. As expected, TRISO particles with extremely large kernels are shown to fail under all considered scenarios. In contrast, the designs that do not depart too drastically from those of the nominal NGNP HTR fuel TRISO particles are shown to perform satisfactorily and display a high rates of survival under all considered scenarios. Finally, it is recognized that relaxing the geometry constraint will result in satisfactory cycle lengths even using UO{sub 2}-loaded TRISO particles-based fuel with enrichment at or below 20 w/o.

R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal Pasamehmetoglu; Francesco Venneri

2012-04-01T23:59:59.000Z

77

Assessment of possible cycle lengths for fully-ceramic micro-encapsulated fuel-based light water reactor concepts  

Science Conference Proceedings (OSTI)

The use of TRISO-particle-based dispersion fuel within SiC matrix and cladding materials has the potential to allow the design of extremely safe LWRs with accident-tolerant fuel. This paper examines the feasibility of LWR-like cycle length for such a low enriched uranium fuel with the imposed constraint of strictly retaining the original geometry of the fuel pins and assemblies. The motivation for retaining the original geometry is to provide the ability to incorporate the fuel 'as-is' into existing LWRs while retaining their thermal-hydraulic characteristics. The feasibility of using this fuel is assessed by looking at cycle lengths and fuel failure rates. Other considerations (e.g., safety parameters, etc.) were not considered at this stage of the study. The study includes the examination of different TRISO kernel diameters without changing the coating layer thicknesses. The study shows that a naive use of UO{sub 2} results in cycle lengths too short to be practical for existing LWR designs and operational demands. Increasing fissile inventory within the fuel compacts shows that acceptable cycle lengths can be achieved. In this study, starting with the recognized highest packing fraction practically achievable (44%), higher enrichment, larger fuel kernel sizes, and the use of higher density fuels have been evaluated. The models demonstrate cycle lengths comparable to those of ordinary LWRs. As expected, TRISO particles with extremely large kernels are shown to fail under all considered scenarios. In contrast, the designs that do not depart too drastically from those of the nominal NGNP HTR fuel TRISO particles are shown to perform satisfactorily and display a high rate of survival under all considered scenarios. Finally, it is recognized that relaxing the geometry constraint will result in satisfactory cycle lengths even using UO{sub 2}-loaded TRISO particles-based fuel with enrichment at or below 20 w/o. (authors)

Sen, R. S.; Pope, M. A.; Ougouag, A. M.; Pasamehmetoglu, K. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Venneri, F. [UltraSafe Nuclear (United States)

2012-07-01T23:59:59.000Z

78

Fuel Cycle and Isotopes Division  

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

Divisions Fuel Cycle and Isotopes Division Jeffrey Binder, Division Director Jeffrey Binder, Division Director The Fuel Cycle and Isotopes Division (FCID) of the Nuclear Science...

79

Strategy for the practical utilization of thorium fuel cycles  

SciTech Connect

There has been increasing interest in the utilization of thorium fuel cycles in nuclear power reactors for the past few years. This is due to a number of factors, the chief being the recent emphasis given to increasing the proliferation resistance of reactor fuel cycles and the thorium cycle characteristic that bred /sup 233/U can be denatured with /sup 238/U (further, a high radioactivity is associated with recycle /sup 233/U, which increases fuel diversion resistance). Another important factor influencing interest in thorium fuel cycles is the increasing cost of U/sub 3/O/sub 8/ ores leading to more emphasis being placed on obtaining higher fuel conversion ratios in thermal reactor systems, and the fact that thorium fuel cycles have higher fuel conversion ratios in thermal reactors than do uranium fuel cycles. Finally, there is increasing information which indicates that fast breeder reactors have significantly higher capital costs than do thermal reactors, such that there is an economic advantage in the long term to have combinations of fast breeder reactors and high-conversion thermal reactors operating together. Overall, it appears that the practical, early utilization of thorium fuel cycles in power reactors requires commercialization of HTGRs operating first on stowaway fuel cycles, followed by thorium fuel recycle. In the longer term, thorium utilization involves use of thorium blankets in fast breeder reactors, in combination with recycling the bred /sup 233/U to HTGRs (preferably), or to other thermal reactors.

Kasten, P.R.

1978-01-01T23:59:59.000Z

80

FUEL CYCLE POTENTIAL WASTE FOR DISPOSITION  

SciTech Connect

The United States (U.S.) currently utilizes a once-through fuel cycle where used nuclear fuel (UNF) is stored on-site in either wet pools or in dry storage systems with ultimate disposal in a deep mined geologic repository envisioned. Within the Department of Energy's (DOE) Office of Nuclear Energy (DOE-NE), the Fuel Cycle Research and Development Program (FCR&D) develops options to the current commercial fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while minimizing proliferation risks by conducting research and development of advanced fuel cycles, including modified open and closed cycles. The safe management and disposition of used nuclear fuel and/or nuclear waste is a fundamental aspect of any nuclear fuel cycle. Yet, the routine disposal of used nuclear fuel and radioactive waste remains problematic. Advanced fuel cycles will generate different quantities and forms of waste than the current LWR fleet. This study analyzes the quantities and characteristics of potential waste forms including differing waste matrices, as a function of a variety of potential fuel cycle alternatives including: (1) Commercial UNF generated by uranium fuel light water reactors (LWR). Four once through fuel cycles analyzed in this study differ by varying the assumed expansion/contraction of nuclear power in the U.S; (2) Four alternative LWR used fuel recycling processes analyzed differ in the reprocessing method (aqueous vs. electro-chemical), complexity (Pu only or full transuranic (TRU) recovery) and waste forms generated; (3) Used Mixed Oxide (MOX) fuel derived from the recovered Pu utilizing a single reactor pass; and (4) Potential waste forms generated by the reprocessing of fuels derived from recovered TRU utilizing multiple reactor passes.

Jones, R.; Carter, J.

2010-10-13T23:59:59.000Z

Note: This page contains sample records for the topic "fuel cycle reactors" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
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81

FUEL CYCLE POTENTIAL WASTE FOR DISPOSITION  

SciTech Connect

The United States (U.S.) currently utilizes a once-through fuel cycle where used nuclear fuel (UNF) is stored on-site in either wet pools or in dry storage systems with ultimate disposal in a deep mined geologic repository envisioned. Within the Department of Energy's (DOE) Office of Nuclear Energy (DOE-NE), the Fuel Cycle Research and Development Program (FCR&D) develops options to the current commercial fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while minimizing proliferation risks by conducting research and development of advanced fuel cycles, including modified open and closed cycles. The safe management and disposition of used nuclear fuel and/or nuclear waste is a fundamental aspect of any nuclear fuel cycle. Yet, the routine disposal of used nuclear fuel and radioactive waste remains problematic. Advanced fuel cycles will generate different quantities and forms of waste than the current LWR fleet. This study analyzes the quantities and characteristics of potential waste forms including differing waste matrices, as a function of a variety of potential fuel cycle alternatives including: (1) Commercial UNF generated by uranium fuel light water reactors (LWR). Four once through fuel cycles analyzed in this study differ by varying the assumed expansion/contraction of nuclear power in the U.S. (2) Four alternative LWR used fuel recycling processes analyzed differ in the reprocessing method (aqueous vs. electro-chemical), complexity (Pu only or full transuranic (TRU) recovery) and waste forms generated. (3) Used Mixed Oxide (MOX) fuel derived from the recovered Pu utilizing a single reactor pass. (4) Potential waste forms generated by the reprocessing of fuels derived from recovered TRU utilizing multiple reactor passes.

Carter, J.

2011-01-03T23:59:59.000Z

82

Advanced Fuel Cycle Cost Basis  

SciTech Connect

This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

2008-03-01T23:59:59.000Z

83

Advanced Fuel Cycle Cost Basis  

SciTech Connect

This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 26 cost modules—24 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, and high-level waste.

D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert

2007-04-01T23:59:59.000Z

84

Advanced Fuel Cycle Cost Basis  

SciTech Connect

This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

2009-12-01T23:59:59.000Z

85

NEUTRONIC REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A fuel slug for a reactor which acts as a safety device is described. The fuel slug is an aluminum tube with a foil lining the inside surface of the tube, the foil being fabricated of uranium in a lead matrix.

Horning, W.A.; Lanning, D.D.; Donahue, D.J.

1959-10-01T23:59:59.000Z

86

NUCLEAR REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

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

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

1963-06-11T23:59:59.000Z

87

World nuclear fuel cycle requirements 1991  

Science Conference Proceedings (OSTI)

The nuclear fuel cycle consists of mining and milling uranium ore, processing the uranium into a form suitable for generating electricity, burning'' the fuel in nuclear reactors, and managing the resulting spent nuclear fuel. This report presents projections of domestic and foreign requirements for natural uranium and enrichment services as well as projections of discharges of spent nuclear fuel. These fuel cycle requirements are based on the forecasts of future commercial nuclear power capacity and generation published in a recent Energy Information Administration (EIA) report. Also included in this report are projections of the amount of spent fuel discharged at the end of each fuel cycle for each nuclear generating unit in the United States. The International Nuclear Model is used for calculating the projected nuclear fuel cycle requirements. 14 figs., 38 tabs.

Not Available

1991-10-10T23:59:59.000Z

88

Evaluation of conceptual flowsheets for incorporating Light Water Reactor (LWR) fuel materials in an advanced nuclear fuel cycle  

SciTech Connect

A preliminary study by a group of experts at ORNL has generated and evaluated a number of aqueous and non-aqueous flowsheets for recovering transuranium actinides from LWR fuel for use as fuel in an LMR and, at the same time, for transmutation of the wastes to less hazardous materials. The need for proliferation resistance was a consideration in the flowsheets. The current state of development of the flowsheets was evaluated and recommendations for additional study were made. 3 refs., 6 figs.

Bell, J.T.; Burch, W.D.; Collins, E.D.; Forsberg, C.W.; Prince, B.E.; Bond, W.D.; Campbell, D.O.; Delene, J.G.; Mailen, J.C.

1990-08-01T23:59:59.000Z

89

DESTRUCTIVE EXAMINATION OF 3-CYCLE LWR (LIGHT WATER REACTOR) FUEL RODS FROM TURKEY POINT UNIT 3 FOR THE CLIMAX - SPENT FUEL TEST  

DOE Green Energy (OSTI)

The destructive examination results of five light water reactor rods from the Turkey Point Unit 3 reactor are presented. The examinations included fission gas collection and analyses, burnup and hydrogen analyses, and a metallographic evaluation of the fuel, cladding, oxide, and hydrides. The rods exhibited a low fission gas release with all other results appearing representative for pressurized water reator fuel rods with similar burnups (28 GWd/MTU) and operating histories.

ATKIN SD

1981-06-01T23:59:59.000Z

90

Fuel Cycle System Analysis Handbook  

Science Conference Proceedings (OSTI)

This Handbook aims to improve understanding and communication regarding nuclear fuel cycle options. It is intended to assist DOE, Campaign Managers, and other presenters prepare presentations and reports. When looking for information, check here. The Handbook generally includes few details of how calculations were performed, which can be found by consulting references provided to the reader. The Handbook emphasizes results in the form of graphics and diagrams, with only enough text to explain the graphic, to ensure that the messages associated with the graphic is clear, and to explain key assumptions and methods that cause the graphed results. Some of the material is new and is not found in previous reports, for example: (1) Section 3 has system-level mass flow diagrams for 0-tier (once-through), 1-tier (UOX to CR=0.50 fast reactor), and 2-tier (UOX to MOX-Pu to CR=0.50 fast reactor) scenarios - at both static and dynamic equilibrium. (2) To help inform fast reactor transuranic (TRU) conversion ratio and uranium supply behavior, section 5 provides the sustainable fast reactor growth rate as a function of TRU conversion ratio. (3) To help clarify the difference in recycling Pu, NpPu, NpPuAm, and all-TRU, section 5 provides mass fraction, gamma, and neutron emission for those four cases for MOX, heterogeneous LWR IMF (assemblies mixing IMF and UOX pins), and a CR=0.50 fast reactor. There are data for the first 10 LWR recycle passes and equilibrium. (4) Section 6 provides information on the cycle length, planned and unplanned outages, and TRU enrichment as a function of fast reactor TRU conversion ratio, as well as the dilution of TRU feedstock by uranium in making fast reactor fuel. (The recovered uranium is considered to be more pure than recovered TRU.) The latter parameter impacts the required TRU impurity limits specified by the Fuels Campaign. (5) Section 7 provides flows for an 800-tonne UOX separation plant. (6) To complement 'tornado' economic uncertainty diagrams, which show at a glance combined uncertainty information, section 9.2 has a new set of simpler graphs that show the impact on fuel cycle costs for once through, 1-tier, and 2-tier scenarios as a function of key input parameters.

Steven J. Piet; Brent W. Dixon; Dirk Gombert; Edward A. Hoffman; Gretchen E. Matthern; Kent A. Williams

2009-06-01T23:59:59.000Z

91

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

SciTech Connect

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

Schneider, K.J.

1982-09-01T23:59:59.000Z

92

Cermet fuel reactors  

Science Conference Proceedings (OSTI)

Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

1987-09-01T23:59:59.000Z

93

FUEL ASSAY REACTOR  

DOE Patents (OSTI)

A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

1962-12-25T23:59:59.000Z

94

Optimization of the Mode of the Uranium-233 Accumulation for Application in Thorium Self-Sufficient Fuel Cycle of Candu Power Reactor  

SciTech Connect

Results of calculation studies of the first stage of self-sufficient thorium cycle for CANDU reactor are presented in the paper. The first stage is preliminary accumulation of {sup 233}U in the CANDU reactor itself. Parameters of active core and scheme of fuel reloading were accepted the same as those for CANDU reactor. It was assumed for calculations, that enriched {sup 235}U or plutonium was used as additional fissile material to provide neutrons for {sup 233}U production. Parameters of 10 different variants of the elementary cell of active core were calculated for the lattice pitch, geometry of fuel channels, and fuel assembly of the CANDU reactor. The results presented in the paper allow to determine the time of accumulation of the required amount of {sup 233}U and corresponding number of targets going into processing for {sup 233}U extraction. Optimum ratio of the accumulation time to number of processed targets can be determined using the cost of electric power produced by the reactor and cost of targets along with their processing. (authors)

Bergelson, Boris; Gerasimov, Alexander [Institute of Theoretical and Experimental Physics, B. Cheremushkinskaya 25, 117259 Moscow (Russian Federation); Tikhomirov, Georgy [Moscow Engineering Physics Institute, Kashirskoe Shosse 31, Moscow (Russian Federation)

2006-07-01T23:59:59.000Z

95

USCEA fuel cycle '93  

SciTech Connect

The US Council for Energy Awareness sponsored the Fuel Cycle '93 conference in Dallas, Texas, on March 21-24, 1993. Over 250 participants attended, numerous papers were presented, and several panel discussions were held. The focus of most industry participants remains the formation of USEC and the pending US-Russian HEU agreement. Following are brief summaries of two key papers and the Fuel Market Issues panel discussion.

Not Available

1993-04-01T23:59:59.000Z

96

Reliability Engineering Approach to Probabilistic Proliferation Resistance Analysis of the Example Sodium Fast Reactor Fuel Cycle Facility  

E-Print Network (OSTI)

International Atomic Energy Agency (IAEA) safeguards are one method of proliferation resistance which is applied at most nuclear facilities worldwide. IAEA safeguards act to prevent the diversion of nuclear materials from a facility through the deterrence of detection. However, even with IAEA safeguards present at a facility, the country where the facility is located may still attempt to proliferate nuclear material by exploiting weaknesses in the safeguards system. The IAEA's mission is to detect the diversion of nuclear materials as soon as possible and ideally before it can be weaponized. Modern IAEA safeguards utilize unattended monitoring systems (UMS) to perform nuclear material accountancy and maintain the continuity of knowledge with regards to the position of nuclear material at a facility. This research focuses on evaluating the reliability of unattended monitoring systems and integrating the probabilistic failure of these systems into the comprehensive probabilistic proliferation resistance model of a facility. To accomplish this, this research applies reliability engineering analysis methods to probabilistic proliferation resistance modeling. This approach is demonstrated through the analysis of a safeguards design for the Example Sodium Fast Reactor Fuel Cycle Facility (ESFR FCF). The ESFR FCF UMS were analyzed to demonstrate the analysis and design processes that an analyst or designer would go through when evaluating/designing the proliferation resistance component of a safeguards system. When comparing the mean time to failure (MTTF) for the system without redundancies versus one with redundancies, it is apparent that redundancies are necessary to achieve a design without routine failures. A reliability engineering approach to probabilistic safeguards system analysis and design can be used to reach meaningful conclusions regarding the proliferation resistance of a UMS. The methods developed in this research provide analysts and designers alike a process to follow to evaluate the reliability of a UMS.

Cronholm, Lillian Marie

2011-08-01T23:59:59.000Z

97

Current Comparison of Advanced Nuclear Fuel Cycles  

SciTech Connect

This paper compares potential nuclear fuel cycle strategies – once-through, recycling in thermal reactors, sustained recycle with a mix of thermal and fast reactors, and sustained recycle with fast reactors. Initiation of recycle starts the draw-down of weapons-usable material and starts accruing improvements for geologic repositories and energy sustainability. It reduces the motivation to search for potential second geologic repository sites. Recycle in thermal-spectru

Steven Piet; Trond Bjornard; Brent Dixon; Robert Hill; Gretchen Matthern; David Shropshire

2007-04-01T23:59:59.000Z

98

NEUTRONIC REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A reactor fuel element of the capillary tube type is described. The element consists of a thin walled tube, sealed at both ends, and having an interior coatlng of a fissionable material, such as uranium enriched in U-235. The tube wall is gas tight and is constructed of titanium, zirconium, or molybdenum.

Kesselring, K.A.; Seybolt, A.U.

1958-12-01T23:59:59.000Z

99

NEUTRONIC REACTOR FUEL PUMP  

DOE Patents (OSTI)

A reactor fuel pump is described which offers long life, low susceptibility to radiation damage, and gaseous fission product removal. An inert-gas lubricated bearing supports a journal on one end of the drive shsft. The other end has an impeller and expansion chamber which effect pumping and gas- liquid separation. (T.R.H.)

Cobb, W.G.

1959-06-01T23:59:59.000Z

100

Nuclear Fuel Cycle Cost Comparison Between Once-Through and Fully Closed Cycles  

Science Conference Proceedings (OSTI)

This report presents results from a parametric study of equilibrium fuel cycle costs for a closed fuel cycle with multi-recycling of plutonium (Pu) and minor actinides in fast reactors (FRs) compared to an open, once-through fuel cycle using pressurized water reactors (PWRs). The study examines the impact on fuel cycle costs from changes in the unit costs of uranium, advanced plutonium and uranium recovery by extraction (PUREX) reprocessing of discharged fast-reactor mixed-oxide (FR-MOX) fuel, and fabric...

2010-11-04T23:59:59.000Z

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


101

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

Science Conference Proceedings (OSTI)

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

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

102

PROCEEDINGS OF THE THORIUM FUEL CYCLE SYMPOSIUM, GATLINBURG, TENNESSEE, DECEMBER 5-7, 1962  

SciTech Connect

Thirty-three papers presented at the Thorium Fuel Cycle symposium are given. Topics covered include fuel-cycle technology, raw materials, reactor physics, and reactor concepts. Separate abstracts were prepared for each of the papers. (M.C.G.)

1963-10-31T23:59:59.000Z

103

An Economic Analysis of Select Fuel Cycles Using the Steady-State Analysis Model for Advanced Fuel Cycles Schemes (SMAFS)  

Science Conference Proceedings (OSTI)

The U.S. Department of Energy's (DOE) Global Nuclear Energy Partnership (GNEP) is currently considering alternatives to the current U.S. once-through fuel cycle. This report evaluates the relative economics of three alternative fuel cycles to determine those cost components important to overall fuel cycle costs and total generation costs. The analysis determined that the unit cost of nuclear reactors is the most important nuclear generation cost parameter in future fuel cycles. The report also evaluates ...

2007-12-20T23:59:59.000Z

104

Physics of fusion-fuel cycles  

SciTech Connect

The evaluation of nuclear fusion fuels for a magnetic fusion economy must take into account the various technological impacts of the various fusion fuel cycles as well as the relative reactivity and the required ..beta..'s and temperatures necessary for economic steady-state burns. This paper will review some of the physics of the various fusion fuel cycles (D-T, catalyzed D-D, D-/sup 3/He, D-/sup 6/Li, and the exotic fuels: /sup 3/He/sup 3/He and the proton-based fuels such as P-/sup 6/Li, P-/sup 9/Be, and P-/sup 11/B) including such items as: (1) tritium inventory, burnup, and recycle, (2) neutrons, (3) condensable fuels and ashes, (4) direct electrical recovery prospects, (5) fissile breeding, etc. The advantages as well as the disadvantages of the different fusion fuel cycles will be discussed. The optimum fuel cycle from an overall standpoint of viability and potential technological considerations appears to be catalyzed D-D, which could also support smaller relatively clean, lean-D, rich-/sup 3/He satellite reactors as well as fission reactors.

McNally, J.R. Jr.

1981-01-01T23:59:59.000Z

105

Proliferation Resistant Nuclear Reactor Fuel  

Science Conference Proceedings (OSTI)

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

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

2011-02-18T23:59:59.000Z

106

Overview of the nuclear fuel cycle  

SciTech Connect

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

Leuze, R.E.

1982-01-01T23:59:59.000Z

107

Fossil fuel furnace reactor  

DOE Patents (OSTI)

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

Parkinson, William J. (Los Alamos, NM)

1987-01-01T23:59:59.000Z

108

Transportation implications of a closed fuel cycle.  

Science Conference Proceedings (OSTI)

Transportation for each step of a closed fuel cycle is analyzed in consideration of the availability of appropriate transportation infrastructure. The United States has both experience and certified casks for transportation that may be required by this cycle, except for the transport of fresh and used MOX fuel and fresh and used Advanced Burner Reactor (ABR) fuel. Packaging that had been used for other fuel with somewhat similar characteristics may be appropriate for these fuels, but would be inefficient. Therefore, the required neutron and gamma shielding, heat dissipation, and criticality were calculated for MOX and ABR fresh and spent fuel. Criticality would not be an issue, but the packaging design would need to balance neutron shielding and regulatory heat dissipation requirements.

Bullard, Tim (University of Nevada - Reno); Bays, Samuel (Idaho National Laboratory); Dennis, Matthew L.; Weiner, Ruth F.; Sorenson, Ken Bryce; Dixon, Brent (Idaho National Laboratory); Greiner, Miles (University of Nevada - Reno)

2010-10-01T23:59:59.000Z

109

Water reactor fuel cladding  

Science Conference Proceedings (OSTI)

This patent describes a nuclear reactor fuel element cladding tube. It comprises: an outer cylindrical layer of a first zirconium alloy selected from the group consisting of Zircaloy-2 and Zircaloy-4; an inner cylindrical layer of a second zirconium alloy consisting essentially of about 0.19 to 0.6 wt.% tin, about 0.19 to less than 0.5 wt.% iron, about 100 to 700 ppm oxygen, less than 2000 ppm total impurities, and the remainder essentially zirconium; the inner layer characterized by aqueous corrosion resistance substantially the same as the first zirconium alloy; the inner layer characterized by improved resistance to PCI crack propagation under reactor operating conditions compared to the first zirconium alloy and substantially the same PCI crack propagation resistance compared to unalloyed zirconium; and the inner cylindrical layer is metallurgically bonded to the outer layer.

Foster, J.P.; McDonald, S.G.

1990-06-12T23:59:59.000Z

110

Reprocessing in breeder fuel cycles  

SciTech Connect

Over the past decade, the United States has developed plans and carried out programs directed toward the demonstration of breeder fuel reprocessing in connection with early breeder demonstration reactors. Although subject to continuing debate, progress continued on the construction of the Clinch River Breeder Reactor (CRBR) with startup anticipated near the end of this decade, while plans for the CRBR and its associated fuel cycle are still being firmed up, the basic R and D programs required to carry out the demonstrations have continued. Policies call for breeder recycle to begin in the early to mid-1990s. An important objective of the reprocessing program is to develop advanced technology for the recovery of fissile materials in systems that minimize environmental emissions and doses to plant workers, and that also provide effective fissile material safeguards. Major improvements include technology for remote operation and maintenance, low-flow ventilation systems coupled with more effective off-gas treatment, and advanced process monitoring for control and safeguards.

Burch, W.D.; Groenier, W.S.

1983-06-01T23:59:59.000Z

111

Reprocessing in breeder fuel cycles  

Science Conference Proceedings (OSTI)

Over the past decade, the United States has developed plans and carried out programs directed toward the demonstration of breeder fuel reprocessing in connection with the first breeder demonstration reactor. A renewed commitment to moving forward with the construction of the Clinch River Breeder Reactor (CRBR) has been made, with startup anticipated near the end of this decade. While plans for the CRBR and its associated fuel cycle are still being firmed up, the basic research and development programs required to carry out the demonstrations have continued. This paper updates the status of the reprocessing plans and programs. Policies call for breeder recycle to begin in the early to mid-1990's. Contents of this paper are: (1) evolving plans for breeder reprocessing (demonstration reprocessing plant, reprocessing head-end colocated at an existing facility); (2) relationship to LWR reprocessing; (3) integrated equipment test (IET) facility and related hardware development activities (mechanical considerations in shearing and dissolving, remote operations and maintenance demonstration phase of IET, integrated process demonstration phase of IET, separate component development activities); and (4) supporting process R and D.

Burch, W.D.; Groenier, W.S.

1982-01-01T23:59:59.000Z

112

NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT  

DOE Patents (OSTI)

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

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

1962-08-14T23:59:59.000Z

113

Fuel Cycle Research & Development | Department of Energy  

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

Fuel Cycle Research & Fuel Cycle Research & Development Fuel Cycle Research & Development Fuel Cycle Research & Development The mission of the Fuel Cycle Research and Development (FCRD) program is to conduct research and development to help develop sustainable fuel cycles, as described in the Nuclear Energy Research and Development Roadmap. Sustainable fuel cycle options are those that improve uranium resource utilization, maximize energy generation, minimize waste generation, improve safety, and limit proliferation risk. The FCRD program will develop a suite of options to enable future policymakers to make informed decisions about how best to manage used fuel from nuclear reactors. The overall goal is to demonstrate the technologies necessary to allow commercial deployment of solutions for the sustainable management of used

114

REACTOR FUEL ELEMENTS TESTING CONTAINER  

DOE Patents (OSTI)

This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

Whitham, G.K.; Smith, R.R.

1963-01-15T23:59:59.000Z

115

FUEL CYCLE PROGRAM. A BOILING WATER REACTOR RESEARCH AND DEVELOPMENT PROGRAM. Eleventh Quarterly Progress Report, January-March 1963  

SciTech Connect

Even though VBWR shutdowns were required for location and removal of five failed fuel assemblies (HPD Program), the increase in fuel exposure was good. Fuel exposures wili pass the values at which cold worked stainless steel cladding was failing under the HPD Program. Failure of the 0.005-inch cold worked stainless steel clad fuel rods in assembly 8L was traced to strain cycling fatigue. A study of tapered fuel rods indicates a potential advantage for us of a variable water/fuel ratio along the flow channel. Natural circulation tests in the hydraulic stability loop were conducted over a range of conditions from stable, to oscillatory with exponential decay, to self-sustaining oscillation of constant amplitude, to unstable oscillations with divergent amplitude. The response to impulses in power input shows the effect of the time delay for transporting steam voids up through the riser. The data permit calculation of oscillation frequency, damping coefficient, time lags, and show the magnitude and character of pressure and velocity changes. The data, which have an experimental scatter of plus or minus 10% maximum, show that burnout heat fiux: decreases with increasing flow up to 2 x 10/sup 8/ lb/hr-ft/sup 2/; has a maximum for hydraulic diameter between 0.25 and 0.5 inch; and decreases for pressure increases between 600 to 1400 psi. A correlating equation for the data is given. The data are compared to results of others. Tests of special geometries show that the burnout heat flux: decreases 22 to 50% when the heated rod is within 0.033 inch of the channel wall; is unchanged upstream of a plate-type spacer; decreases 35 to 50% when the rod surface is roughened by sandblasting; is increased 20 to 40% by use of a rough liner. The four-rod test section is operating satisfactorily and 17 critical heat fiux data points are obtained at 1000 psia and flows of 0.5, 1.0, and 1.5 x 10/sup 6/ lb/hr-ft/sup 2/. In each case the critical heat flux occurred at the exit end and on the side of the rod facing the corner of the channel. The evaluation of film trippers (rough liner) on the unheated channel walls indicates considerable promise for increasing the burnout heat flux limit. The theory of operation is that the liquid film on the unheated wail is sheared off and dispersed, thus adding to the liquid film on the heated rod. Measurements with a heater rod bowed so that it is in contact with the channel wall show that the critical heat flux is decreased by a factor of two or more from values with normal clearance. Temperature measurements on the rod, when operating past the critical heat flux, were in the order of magnitude of 1000 deg F for heat fluxes of about 500,000--600,000 Btu/hr-ft/sup 2/. Chemical analyses for radial variations in isotopic composition within a fuel pellet are nearly completed and are compiled for interpretation. (N.W.R.)

Howard, C.L. comp.

1963-04-01T23:59:59.000Z

116

TURRET: A HIGH TEMPERATURE GAS-CYCLE REACTOR PROPOSAL  

SciTech Connect

A nitrogen-cooled graphite-moderated nuclear reactor experiment is proposed to drive a closed-cycle gas turbine power plant at 1300 deg F. The annular core of the reactor can be rotated inside the reflector to permit fuel loading and discharge while operating at full power. Small cylindrical fuel elements of graphite are solutionimpregnated with partially enriched uranium. The fuel is recycled by incineration of the elements, chemical fresh graphite tn a small batch process. The unclad, uncoated fuel should permit high burn-up and simple fuel processing, but allows fission product diffusion into the gas stream. While methods are proposed for the removal of these from the gas, the Song-term consequences on turbine operation are unknown. The compatibility of nitrogen gas with the fuel has been studied experimentally. The radial movement of fuel gives a reactor with a constant power profile and no excess reactivity. The temperature is regulated by the fuel charging rate. (auth)

Hammond, R.P.; Busey, H.M.; Chapman, K.R.; Durham, F.P.; Rogers, J.D.; Wykoff, W.R.

1958-01-23T23:59:59.000Z

117

Fuel Cycle Science & Technology | Nuclear Science | ORNL  

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

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

118

Parametric Study of Front-End Nuclear Fuel Cycle Costs  

Science Conference Proceedings (OSTI)

This study provides an overview of front-end fuel cost components for nuclear plants, specifically uranium concentrates, uranium conversion services, uranium enrichment services, and nuclear fuel fabrication services. A parametric analysis of light-water reactor (LWR) fuel cycle costs is also included in order to quantify the impacts that result from changes in the cost of one or more front-end components on overall fuel cycle costs.

2009-02-20T23:59:59.000Z

119

Fuel Reformation: Microchannel Reactor Design  

DOE Green Energy (OSTI)

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

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

2005-09-01T23:59:59.000Z

120

Immobilization of Fast Reactor First Cycle Raffinate  

Science Conference Proceedings (OSTI)

This paper describes the results of work to bring forward the timing for the immobilization of first cycle raffinate from reprocessing fuel from the Dounreay Prototype Fast Reactor (PFR). First cycle raffinate is the liquor which contains > 99% of the fission products separated from spent fuel during reprocessing. Approximately 203 m3 of raffinate from the reprocessing of PFR fuel is held in four tanks at the UKAEA's site at Dounreay, Scotland. Two methods of immobilization of this high level waste (HLW) have been considered: vitrification and cementation. Vitrification is the standard industry practice for the immobilization of first cycle raffinate, and many papers have been presented on this technique elsewhere. However, cementation is potentially feasible for immobilizing first cycle raffinate because the heat output is an order of magnitude lower than typical HLW from commercial reprocessing operations such as that at the Sellafield site in Cumbria, England. In fact, it falls within the upper end of the UK definition of intermediate level waste (ILW). Although the decision on which immobilization technique will be employed has yet to be made, initial development work has been undertaken to identify a suitable cementation formulation using inactive simulant of the raffinate. An approach has been made to the waste disposal company Nirex to consider the disposability of the cemented product material. The paper concentrates on the process development work that is being undertaken on cementation to inform the decision making process for selection of the immobilization method.

Langley, K. F.; Partridge, B. A.; Wise, M.

2003-02-26T23:59:59.000Z

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


121

Software Requirements Specification Verifiable Fuel Cycle Simulation (VISION) Model  

SciTech Connect

The purpose of this Software Requirements Specification (SRS) is to define the top-level requirements for a Verifiable Fuel Cycle Simulation Model (VISION) of the Advanced Fuel Cycle (AFC). This simulation model is intended to serve a broad systems analysis and study tool applicable to work conducted as part of the AFCI (including costs estimates) and Generation IV reactor development studies.

D. E. Shropshire; W. H. West

2005-11-01T23:59:59.000Z

122

Impact of actinide recycle on nuclear fuel cycle health risks  

SciTech Connect

The purpose of this background paper is to summarize what is presently known about potential impacts on the impacts on the health risk of the nuclear fuel cycle form deployment of the Advanced Liquid Metal Reactor (ALMR){sup 1} and Integral Fast Reactor (IF){sup 2} technology as an actinide burning system. In a companion paper the impact on waste repository risk is addressed in some detail. Therefore, this paper focuses on the remainder of the fuel cycle.

Michaels, G.E.

1992-06-01T23:59:59.000Z

123

MOLTEN FLUORIDE NUCLEAR REACTOR FUEL  

DOE Patents (OSTI)

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

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

1960-01-01T23:59:59.000Z

124

DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR  

DOE Patents (OSTI)

A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

1962-08-14T23:59:59.000Z

125

Fusion fuel cycle solid radioactive wastes  

SciTech Connect

Eight conceptual deuterium-tritium fueled fusion power plant designs have been analyzed to identify waste sources, materials and quantities. All plant designs include the entire D-T fuel cycle within each plant. Wastes identified include radiation-damaged structural, moderating, and fertile materials; getter materials for removing corrosion products and other impurities from coolants; absorbents for removing tritium from ventilation air; getter materials for tritium recovery from fertile materials; vacuum pump oil and mercury sludge; failed equipment; decontamination wastes; and laundry waste. Radioactivity in these materials results primarily from neutron activation and from tritium contamination. For the designs analyzed annual radwaste volume was estimated to be 150 to 600 m/sup 3//GWe. This may be compared to 500 to 1300 m/sup 3//GWe estimated for the LMFBR fuel cycle. Major waste sources are replaced reactor structures and decontamination waste.

Gore, B.F.; Kaser, J.D.; Kabele, T.J.

1978-06-01T23:59:59.000Z

126

Kinetics of pyroprocesses in ATW fuel cycles  

SciTech Connect

Accelerator-driven transmutation of waste (ATW) combines the technologies of accelerators and reactors to treat the nuclear waste problem. An ATW system uses a high-current accelerator to generate spallation neutrons to initiate the transmutation of actinides and select fission products in a subcritical nuclear assembly surrounding the target volume. For high burnup and efficient operation, an ATW system requires simple, reliable, and efficient fuel preparation and cleanup procedures to periodically remove {open_quotes}neutron poisons.{close_quotes} We have identified several fuel cycles based on pyroprocessing.

Li, Ning; Hu, Y.C.; Park, B.G. [Los Alamos National Lab., NM (United States)

1997-12-01T23:59:59.000Z

127

FUSED REACTOR FUELS  

DOE Patents (OSTI)

This invention relates to a nuciear reactor fuel composition comprising (1) from about 0.01 to about 50 wt.% based on the total weight of said composition of at least one element selected from the class consisting of uranium, thorium, and plutonium, wherein said eiement is present in the form of at least one component selected from the class consisting of oxides, halides, and salts of oxygenated anions, with components comprising (2) at least one member selected from the class consisting of (a) sulfur, wherein the sulfur is in the form of at least one entity selected irom the class consisting of oxides of sulfur, metal sulfates, metal sulfites, metal halosulfonates, and acids of sulfur, (b) halogen, wherein said halogen is in the form of at least one compound selected from the class of metal halides, metal halosulfonates, and metal halophosphates, (c) phosphorus, wherein said phosphorus is in the form of at least one constituent selected from the class consisting of oxides of phosphorus, metal phosphates, metal phosphites, and metal halophosphates, (d) at least one oxide of a member selected from the class consisting of a metal and a metalloid wherein said oxide is free from an oxide of said element in (1); wherein the amount of at least one member selected from the class consisting of halogen and sulfur is at least about one at.% based on the amount of the sum of said sulfur, halogen, and phosphorus atom in said composition; and wherein the amount of said 2(a), 2(b) and 2(c) components in said composition which are free from said elements of uranium, thorium, arid plutonium, is at least about 60 wt.% based on the combined weight of the components of said composition which are free from said elements of uranium, thorium, and plutonium. (AEC)

Mayer, S.W.

1962-11-13T23:59:59.000Z

128

FUEL ELEMENT FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

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

Bassett, C.H.

1961-05-16T23:59:59.000Z

129

Analysis of Nuclear Proliferation Resistance of DUPIC Fuel Cycle  

E-Print Network (OSTI)

with other fuel cycle cases. The other fuel cycles considered in this study are PWR of once-through mode (PWR-OT), PWR of reprocessing mode (PWR-MOX), in which spent PWR fuel is reprocessed and recovered plutonium is used for making MOX (Mixed Oxide), CANDU with once-through mode (CANDU-OT), PWR fuel and CANDU fuel in a oncethrough mode with reactor grid equivalent to DUPIC fuel cycle (PWR-CANDU-OT). This study is focused on intrinsic barriers, especially, radiation field of the diverted material, which could be a significant accessibility barrier, amount of special nuclear material based on 1 GWe-yr that has to be diverted and the quality of the separated fissile material. It is indicated from plutonium analysis of each fuel cycle that the MOX spent fuel is containing the largest plutonium per MTHM but PWR-MOX option based on 1 GWe-yr has the best benefit in total plutonium consumption aspects. The DUPIC option is containing a little higher total plutonium based on 1 GWe-yr than the PWR-MOX case, but the DUPIC option has the lowest fissile plutonium content which could be another measure for proliferation resistance. On the whole, the CANDU-OT option has the largest fissile plutonium as well as total plutonium per GWe-yr, which means negative points in nuclear proliferation resistance aspects. It is indicated from the radiation field analysis that fresh DUPIC fuel could play an important radiation barrier role, more than even CANDU spent fuels. In conclusion, due to those inherent features, the DUPIC fuel cycle could include technical characteristics that comply naturally with the Spent Fuel Standard, at all steps along the DUPIC linkage between PWR and CANDU. KEYWORDS: DUPIC (direct use of spent PWR fuel in CANDU), (DUPIC) fuel cycle, nuclear fuel cycle analysis, nuclear proliferaion resistance, proliferation resistance barrier, safeguards, plutonium analysis, candu type reactors, spent fuels, fuel cycles I.

Won Il Ko; Ho Dong Kim

2001-01-01T23:59:59.000Z

130

Answering Key Fuel Cycle Questions  

Science Conference Proceedings (OSTI)

Given the range of fuel cycle goals and criteria, and the wide range of fuel cycle options, how can the set of options eventually be narrowed in a transparent and justifiable fashion? It is impractical to develop all options. We suggest an approach that starts by considering a range of goals for the Advanced Fuel Cycle Initiative (AFCI) and then posits seven questions, such as whether Cs and Sr isotopes should be separated from spent fuel and, if so, what should be done with them. For each question, we consider which of the goals may be relevant to eventually providing answers. The AFCI program has both ''outcome'' and ''process'' goals because it must address both waste already accumulating as well as completing the fuel cycle in connection with advanced nuclear power plant concepts. The outcome objectives are waste geologic repository capacity and cost, energy security and sustainability, proliferation resistance, fuel cycle economics, and safety. The process objectives are rea diness to proceed and adaptability and robustness in the face of uncertainties.

Piet, S.J.; Dixon, B.W.; Bennett, R.G.; Smith, J.D.; Hill, R.N.

2004-10-03T23:59:59.000Z

131

FUEL ELEMENT FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

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

Bassett, C.H.

1961-11-21T23:59:59.000Z

132

Filling Knowledge Gaps with Five Fuel Cycle Studies  

SciTech Connect

During FY 2010, five studies were conducted of technology families’ applicability to various fuel cycle strategies to fill in knowledge gaps in option space and to better understand trends and patterns. Here, a “technology family” is considered to be defined by a type of reactor and by selection of which actinides provide fuel. This report summarizes the higher-level findings; the detailed analyses and results are documented in five individual reports, as follows: • Advanced once through with uranium fuel in fast reactors (SFR), • Advanced once through (uranium fuel) or single recycle (TRU fuel) in high temperature gas cooled reactors (HTGR), • Sustained recycle with Th/U-233 in light water reactors (LWRs), • Sustained recycle with Th/U-233 in molten salt reactors (MSR), and • Several fuel cycle missions with Fusion-Fission Hybrid (FFH). Each study examined how the designated technology family could serve one or more designated fuel cycle missions, filling in gaps in overall option space. Each study contains one or more illustrative cases that show how the technology family could be used to meet a fuel cycle mission, as well as broader information on the technology family such as other potential fuel cycle missions for which insufficient information was available to include with an illustrative case. None of the illustrative cases can be considered as a reference, baseline, or nominal set of parameters for judging performance; the assessments were designed to assess areas of option space and were not meant to be optimized. There is no implication that any of the cases or technology families are necessarily the best way to meet a given fuel cycle mission. The studies provide five examples of 1-year fuel cycle assessments of technology families. There is reasonable coverage in the five studies of the performance areas of waste management and uranium utilization. The coverage of economics, safety, and proliferation resistance and physical protection in the five studies was spotty. Some studies did not have existing or past work to draw on in one or more of these areas. Resource constraints limited the amount of new analyses that could be performed. Little or no assessment was done of how soon any of the technologies could be deployed and therefore how quickly they could impact domestic or international fuel cycle performance. There were six common R&D needs, such as the value of advanced fuels, cladding, coating, and structure that would survive high neutron fluence. When a technology family is considered for use in a new fuel cycle mission, fuel cycle performance characteristics are dependent on both the design choices and the fuel cycle approach. For example, the use of the sodium-cooled fast reactor to provide recycle in either breeder or burner mode has been studied for decades, but the SFR could be considered for once-through fuel cycle with the physical reactor design and fuel management parameters changed. In addition, the sustained recycle with Th/U-233 in LWR could be achieved with a heterogeneous assembly and derated power density. Therefore, it may or may not be adjustable for other fuel cycle missions although a reactor intended for one fuel cycle mission is built. Simple parameter adjustment in applying a technology family to a new fuel cycle mission should be avoided and, if observed, the results viewed with caution.

Steven J. Piet; Jess Gehin; William Halsey; Temitope Taiwo

2010-11-01T23:59:59.000Z

133

Preliminary analysis of alternative fuel cycles for proliferation evaluation  

SciTech Connect

The ERDA Division of Nuclear Research and Applications proposed 67 nuclear fuel cycles for assessment as to their nonproliferation potential. The object of the assessment was to determine which fuel cycles pose inherently low risk for nuclear weapon proliferation while retaining the major benefits of nuclear energy. This report is a preliminary analysis of these fuel cycles to develop the fuel-recycle data that will complement reactor data, environmental data, and political considerations, which must be included in the overall evaluation. This report presents the preliminary evaluations from ANL, HEDL, ORNL, and SRL and is the basis for a continuing in-depth study. (DLC)

Steindler, M. J.; Ripfel, H. C.F.; Rainey, R. H.

1977-01-01T23:59:59.000Z

134

FUEL ELEMENT FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

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

Bassett, C.H.

1961-05-01T23:59:59.000Z

135

NEUTRONIC REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A fuel element possessing good stability and heat conducting properties is described. The fuel element comprises an outer tube formed of material selected from the group consisting of stainhess steel, V, Ti. Mo. or Zr, a fuel tube concentrically fitting within the outer tube and containing an oxide of an isotope selected from the group consisting of U/sup 235/, U/sup 233/, and Pu/sup 239/, and a hollow, porous core concentrically fitting within the fuel tube and formed of an oxide of an element selected from the group consisting of Mg, Be, and Zr.

Shackleford, M.H.

1958-12-16T23:59:59.000Z

136

Technology Insights and Perspectives for Nuclear Fuel Cycle Concepts  

SciTech Connect

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

137

NEUTRONIC REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A nuclear fuel element comprising a plurality of nuclear fuel bearing strips is presented. The strips are folded along their longitudinal axes to an angle of about 60 deg and are secured at each end by ferrule to form an elongated assembly suitable for occupying a cylindrical coolant channel.

Gurinsky, D.H.; Powell, R.W.; Fox, M.

1959-11-24T23:59:59.000Z

138

FUEL ELEMENT FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

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

Dickson, J.J.

1963-09-24T23:59:59.000Z

139

Fuel cycles for the 80's  

SciTech Connect

Papers presented at the American Nuclear Society's topical meeting on the fuel cycle are summarized. Present progress and goals in the areas of fuel fabrication, fuel reprocessing, spent fuel storage, accountability, and safeguards are reported. Present governmental policies which affect the fuel cycle are also discussed. Individual presentations are processed for inclusion in the Energy Data Base.(DMC)

Not Available

1980-01-01T23:59:59.000Z

140

EFFECT OF REDUCED U-235 PRICE ON FUEL CYCLE COSTS  

SciTech Connect

A study was made to determine the effect of changes in natural uranium cost and in separative work charges on fuel cycle costs in nuclear power plants. Reactors considered were a Dresden-type boiling water reactor (BWR) and a Yankee- type pressurized water reactor (PWR), with net power ratings of 100, 300, and 500 Mwe. Fuel cycle costs were calculated for these reactors, using either enriched uranium or U/sup 235/-thorium as the fuel material. The price schedule for uranium was based on a feed material cost of /kg uranium as UF/sub 6/ and separative work costs of /kg uranium (Schedule B) and /kg uranium (Schedule C). The present AEC price schedule for enriched uranium was also used for purposes of a reference case. The results indicate that a reduction in present enriched uranium price to that given by Schedule B would reduce fuel cycle costs for the BWR plants by 0.4 to 0.5 mill/kwh for the enriched-uranium cycle, and 0.4 to 0.7 mill/kwh for the thorium cycle. Reductions in fuel cycle costs for the PWR plants were 0.5 to 0.7 and 0.4 to 0.75 mill/kwh, respectively, for the same situations. (auth)

Bennett, L.L.

1962-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "fuel cycle reactors" from the National Library of EnergyBeta (NLEBeta).
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141

Back-end costs of alternative nuclear fuel cycles  

Science Conference Proceedings (OSTI)

As part of its charter, the Alternate Fuel Cycle Evaluation Program (AFCEP) was directed to evaluate the back-end of the nuclear fuel cycle in support of the Nonproliferation Alternative Systems Assessment Program (NASAP). The principal conclusion from this study is that the costs for recycling a broad range of reactor fuels will not have a large impact on total fuel cycle costs. For the once-through fuel cycle, the costs of fresh fuel fabrication, irradiated fuel storage, and associated transportation is about 1.2 to 1.3 mills/kWh. For the recycle of uranium and plutonium into thermal reactors, the back-cycle costs (i.e., the costs of irradiated fuel storage, transportation, reprocessing, refabrication, and waste disposal) will be from 3 to 3.5 mills/kWh. The costs for the recycle of uranium and plutonium into fast breeder reactors will be from 4.5 to 5 mills/kWh. Using a radioactive spikant or a denatured /sup 233/U-Th cycle will increase power costs for both recycle cases by about 1 mill/kWh. None of these costs substantially influence the total cost of nuclear power, which is in the range of 4 cents/kWh. The fuel cycle costs used in this study do not include costs incurred prior to fuel fabrication; that is, the cost of the uranium or thorium, the costs for enrichment, or credit for fissile materials in the discharged fuel, which is not recycled with the system.

Rainey, R.H.; Burch, W.D.; Haire, M.J.; Unger, W.E.

1980-01-01T23:59:59.000Z

142

Nuclear reactor composite fuel assembly  

DOE Patents (OSTI)

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

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

1980-01-01T23:59:59.000Z

143

Interim assessment of the denatured /sup 233/U fuel cycle: feasibility and nonproliferation characteristics  

SciTech Connect

A fuel cycle that employs /sup 233/U denatured with /sup 238/U and mixed with thorium fertile material is examined with respect to its proliferation-resistance characteristics and its technical and economic feasibility. The rationale for considering the denatured /sup 233/U fuel cycle is presented, and the impact of the denatured fuel on the performance of Light-Water Reactors, Spectral-Shift-Controlled Reactors, Gas-Cooled Reactors, Heavy-Water Reactors, and Fast Breeder Reactors is discussed. The scope of the R, D and D programs to commercialize these reactors and their associated fuel cycles is also summarized and the resource requirements and economics of denatured /sup 233/U cycles are compared to those of the conventional Pu/U cycle. In addition, several nuclear power systems that employ denatured /sup 233/U fuel and are based on the energy center concept are evaluated.

Abbott, L.S.; Bartine, D.E.; Burns, T.J. (eds.)

1979-12-01T23:59:59.000Z

144

Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2  

Science Conference Proceedings (OSTI)

The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

2002-09-01T23:59:59.000Z

145

Safety aspects of the IFR pyroprocess fuel cycle  

SciTech Connect

This paper addresses the important safety considerations related to the unique Integral Fast Reactor (IFR) fuel cycle technology, the pyroprocess. Argonne has been developing the IFR since 1984. It is a liquid metal cooled reactor, with a unique metal alloy fuel, and it utilizes a radically new fuel cycle. An existing facility, the Hot Fuel Examination Facility-South (HFEF/S) is being modified and equipped to provide a complete demonstration of the fuel cycle. This paper will concentrate on safety aspects of the future HFEF/S operation, slated to begin late next year. HFEF/S is part of Argonne's complex of reactor test facilities located on the Idaho National Engineering Laboratory. HFEF/S was originally put into operation in 1964 as the EBR-II Fuel Cycle Facility (FCF) (Stevenson, 1987). From 1964--69 FCF operated to demonstrate an earlier and incomplete form of today's pyroprocess, recycling some 400 fuel assemblies back to EBR-II. The FCF mission was then changed to one of an irradiated fuels and materials examination facility, hence the name change to HFEF/S. The modifications consist of activities to bring the facility into conformance with today's much more stringent safety standards, and, of course, providing the new process equipment. The pyroprocess and the modifications themselves are described more fully elsewhere (Lineberry, 1987; Chang, 1987). 18 refs., 5 figs., 2 tabs.

Forrester, R.J.; Lineberry, M.J.; Charak, I.; Tessier, J.H.; Solbrig, C.W.; Gabor, J.D.

1989-01-01T23:59:59.000Z

146

Assessment of Browns Ferry 2 Cycle 12 Fuel Corrosion Failures  

Science Conference Proceedings (OSTI)

Boiling water reactor (BWR) fuel rods from 63 bundles of the Reload 10 GE13 (9x9) design developed leaks during Cycle 12 at Browns Ferry 2 (BF-2). Corrosion failures also occurred in Browns Ferry 3 (BF-3) and Vermont Yankee (VY) in a similar time frame. These fuel failures were investigated in the spent fuel pool and in two separate hot cell examination campaigns. This report compiles and assesses the significant findings of the root cause investigation.

2011-11-30T23:59:59.000Z

147

Development Plan for the Fuel Cycle Simulator  

Science Conference Proceedings (OSTI)

The Fuel Cycle Simulator (FCS) project was initiated late in FY-10 as the activity to develop a next generation fuel cycle dynamic analysis tool for achieving the Systems Analysis Campaign 'Grand Challenge.' This challenge, as documented in the Campaign Implementation Plan, is to: 'Develop a fuel cycle simulator as part of a suite of tools to support decision-making, communication, and education, that synthesizes and visually explains the multiple attributes of potential fuel cycles.'

Brent Dixon

2011-09-01T23:59:59.000Z

148

NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY  

DOE Patents (OSTI)

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

Stengel, F.G.

1963-12-24T23:59:59.000Z

149

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

150

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

151

Fuel handling apparatus for a nuclear reactor  

DOE Patents (OSTI)

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

Hawke, Basil C. (Solana Beach, CA)

1987-01-01T23:59:59.000Z

152

Advanced Fuel Cycle Economic Sensitivity Analysis  

Science Conference Proceedings (OSTI)

A fuel cycle economic analysis was performed on four fuel cycles to provide a baseline for initial cost comparison using the Gen IV Economic Modeling Work Group G4 ECON spreadsheet model, Decision Programming Language software, the 2006 Advanced Fuel Cycle Cost Basis report, industry cost data, international papers, the nuclear power related cost study from MIT, Harvard, and the University of Chicago. The analysis developed and compared the fuel cycle cost component of the total cost of energy for a wide range of fuel cycles including: once through, thermal with fast recycle, continuous fast recycle, and thermal recycle.

David Shropshire; Kent Williams; J.D. Smith; Brent Boore

2006-12-01T23:59:59.000Z

153

The Basis for Developing Samarium AMS for Fuel Cycle Analysis  

SciTech Connect

Modeling of nuclear reactor fuel burnup indicates that the production of samarium isotopes can vary significantly with reactor type and fuel cycle. The isotopic concentrations of {sup 146}Sm, {sup 149}Sm, and {sup 151}Sm are potential signatures of fuel reprocessing, if analytical techniques can overcome the inherent challenges of lanthanide chemistry, isobaric interferences, and mass/charge interferences. We review the current limitations in measurement of the target samarium isotopes and describe potential approaches for developing Sm-AMS. AMS sample form and preparation chemistry will be discussed as well as possible spectrometer operating conditions.

Buchholz, B A; Biegalski, S R; Whitney, S M; Tumey, S J; Weaver, C J

2008-10-13T23:59:59.000Z

154

Fuel Cycle Research and Development Program  

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

Development Program Presentation to Office of Environmental Management Tank Waste Corporate Board James C. Bresee, ScD, JD Advisory Board Member Office of Nuclear Energy July 29, 2009 July 29, 2009 Fuel Cycle Research and Development DM 195665 2 Outline Fuel Cycle R&D Mission Changes from the Former Advanced Fuel Cycle Initiative The Science-Based Approach Key Collaborators Budget History Program Elements Summary July 29, 2009 Fuel Cycle Research and Development DM 195665 3 Fuel Cycle R&D Mission The mission of Fuel Cycle Research and Development is to develop options to current fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while reducing proliferation risks by conducting

155

Fuel Cycle CrossCut Group  

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

CrossCut Group CrossCut Group 1 NERAC Briefing: Assessment of Dose of Closed vs Open Gen-IV Fuel Cycles David Wade NERAC Meeting September 30, 2002 Fuel Cycle CrossCut Group 2 Public Dose and Worker Dose Comparison of Open vs Closed Fuel Cycles * Gen-IV fuel cycle options are meant to address all stated Gen-IV Goals - Dose to workers and to the public is one of the numerous elements to be evaluated by Gen-IV R&D - The Fuel Cycle Crosscut Group was assigned to take an early look at dose implication tradeoffs of open and closed fuel cycles * FCCG Interpretation of Assignment: - Collect already-existing evaluations and prepare a briefing on what is currently known Fuel Cycle CrossCut Group 3 Approach * Look at Actual Historical Doses Based on Operational Experience - Data compiled by the United Nations Scientific Committee on the Effects of Atomic

156

Fast Reactor Fuel Type and Reactor Safety Performance  

Science Conference Proceedings (OSTI)

Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and thermophysical properties of the fuel and their compatibility with the reactor coolant, with corresponding differences in the challenges presented to the reactor developers. Accident phenomena are discussed for the sodium-cooled fast reactor based on the mechanistic progression of conditions from accident initiation to accident termination, whether a benign state is achieved or more severe consequences are expected. General principles connecting accident phenomena and fuel properties are developed from the oxide and metal fuel safety analyses, providing guidelines that can be used as part of the evaluation for selection of fuel type for the sodium-cooled fast reactor.

R. Wigeland; J. Cahalan

2009-09-01T23:59:59.000Z

157

Answering Key Fuel Cycle Questions  

Science Conference Proceedings (OSTI)

The Advanced Fuel Cycle Initiative (AFCI) program has both “outcome” and “process” goals because it must address both waste already accumulating as well as completing the fuel cycle in connection with advanced nuclear power plant concepts. The outcome objectives are waste geological repository capacity and cost, energy security and sustainability, proliferation resistance, fuel cycle economics, and safety. The process objectives are readiness to proceed and adaptability and robustness in the face of uncertainties. A classic decision-making approach to such a multi-attribute problem would be to weight individual quantified criteria and calculate an overall figure of merit. This is inappropriate for several reasons. First, the goals are not independent. Second, the importance of different goals varies among stakeholders. Third, the importance of different goals is likely to vary with time, especially the “energy future.” Fourth, some key considerations are not easily or meaningfully quantifiable at present. Instead, at this point, we have developed 16 questions the AFCI program should answer and suggest an approach of determining for each whether relevant options improve meeting each of the program goals. We find that it is not always clear which option is best for a specific question and specific goal; this helps identify key issues for future work. In general, we suggest attempting to create as many win-win decisions (options that are attractive or neutral to most goals) as possible. Thus, to help clarify why the program is exploring the options it is, and to set the stage for future narrowing of options, we have developed 16 questions, as follows: · What are the AFCI program goals? · Which potential waste disposition approaches do we plan for? · What are the major separations, transmutation, and fuel options? · How do we address proliferation resistance? · Which potential energy futures do we plan for? · What potential external triggers do we plan for? · Should we separate uranium? · If we separate uranium, should we recycle it, store it or dispose of it? · Is it practical to plan to fabricate and handle “hot” fuel? · Which transuranic elements (TRU) should be separated and transmuted? · Of those TRU separated, which should be transmuted together? · Should we separate and/or transmute Cs and Sr isotopes that dominate near-term repository heating? · Should we separate and/or transmute very long-lived Tc and I isotopes? · Which separation technology? · What mix of transmutation technologies? · What fuel technology best supports the above decisions?

Steven J. Piet; Brent W. Dixon; J. Stephen Herring; David E. Shropshire; Mary Lou Dunzik-Gougar

2003-10-01T23:59:59.000Z

158

Poolside Fuel Inspection and Fuel Desposit Evaluation at LaSalle-1 (Cycle 10)  

Science Conference Proceedings (OSTI)

LaSalle Unit 1 is a boiling water reactor (BWR) plant running a noble metals chemical addition (NMCA) chemistry program starting with Cycle 10. Fuel inspection and deposit samples were obtained from two fuel assemblies after Cycle 10. These two assemblies contained rods failed by mechanisms unrelated to corrosion or crud. Results of a poolside examination, bulk sample analysis of 15 crud samples, and detailed flake analysis are reported in this document.

2005-11-28T23:59:59.000Z

159

Fuel cycle and waste management demonstration in the IFR Program  

Science Conference Proceedings (OSTI)

Argonne's National Laboratory's Integral Fast Reactor (IFR) is the main element in the US advanced reactor development program. A unique fuel cycle and waste process technology is being developed for the IFR. Demonstration of this technology at engineering scale will begin within the next year at the EBR-II test facility complex in Idaho. This paper describes the facility being readied for this demonstration, the process to be employed, the equipment being built, and the waste management approach.

Lineberry, M.J.; Phipps, R.D.; Benedict, R.W. (Argonne National Lab., Idaho Falls, ID (United States)); Laidler, J.J.; Battles, J.E.; Miller, W.E. (Argonne National Lab., IL (United States))

1992-01-01T23:59:59.000Z

160

Fuel cycle and waste management demonstration in the IFR Program  

SciTech Connect

Argonne`s National Laboratory`s Integral Fast Reactor (IFR) is the main element in the US advanced reactor development program. A unique fuel cycle and waste process technology is being developed for the IFR. Demonstration of this technology at engineering scale will begin within the next year at the EBR-II test facility complex in Idaho. This paper describes the facility being readied for this demonstration, the process to be employed, the equipment being built, and the waste management approach.

Lineberry, M.J.; Phipps, R.D.; Benedict, R.W. [Argonne National Lab., Idaho Falls, ID (United States); Laidler, J.J.; Battles, J.E.; Miller, W.E. [Argonne National Lab., IL (United States)

1992-09-01T23:59:59.000Z

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


161

Westinghouse Advanced Reactors Division Plutonium Fuel Laboratories  

Office of Legacy Management (LM)

ADVANCED REACTORS DIVISION FUEL LABORATORIES CHESWICK, PENNSYLVANIA Department of Energy Office of Policy, Safety and Environment Office of Operational Safety Environmental...

162

Accident Tolerant Fuels for Light Water Reactors  

Science Conference Proceedings (OSTI)

Presentation Title, Accident Tolerant Fuels for Light Water Reactors. Author(s), Steven J. Zinkle, Kurt A. Terrani, Lance L. Snead. On-Site Speaker (Planned) ...

163

Advanced Nuclear Fuel Cycles -- Main Challenges and Strategic Choices  

Science Conference Proceedings (OSTI)

This report presents the results of a critical review of the technological challenges to the growth of nuclear energy, emerging advanced technologies that would have to be deployed, and fuel cycle strategies that could conceivably involve interim storage, plutonium recycling in thermal and fast reactors, reprocessed uranium recycling, and transmutation of minor actinide elements and fission products before eventual disposal of residual wastes.

2010-09-02T23:59:59.000Z

164

Closed DTU fuel cycle with Np recycle and waste transmutation  

Science Conference Proceedings (OSTI)

A nuclear energy scenario for the 21st century that included a denatured thorium-uranium-oxide (DTU) fuel cycle and new light water reactors (LWRs) supported by accelerator-driven transmutation of waste (ATW) systems was previously described. This coupled system with the closed DTU fuel cycle provides several improvements beyond conventional LWR (CLWR) (once-through, UO{sub 2} fuel) nuclear technology: increased proliferation resistance, reduced waste, and efficient use of natural resources. However, like CLWR fuel cycles, the spent fuel in the first one-third core discharged after startup contains higher-quality Pu than the equilibrium fuel cycle. To eliminate this high-grade Pu, Np is separated and recycled with Th and U--rather than with higher actinides [(HA) including Pu]. The presence of Np in the LWR feed greatly increases the production of {sup 238}Pu so that a few kilograms of Pu generated enough alpha-decay heat that the separated Pu is highly resistant to proliferation. This alternate process also simplifies the pyrochemical separation of fuel elements (Th and U) from HAs. To examine the advantages of this concept, the authors modeled a US deployment scenario for nuclear energy that includes DTU-LWRs plus ATW`s to burn the actinides produced by these LWRs and to close the back-end of the DTU fuel cycle.

Beller, D.E.; Sailor, W.C.; Venneri, F. [Los Alamos National Lab., NM (United States); Herring, J.S. [Idaho National Engineering and Environmental Lab., ID (United States)

1999-09-01T23:59:59.000Z

165

WEB RESOURCES: The Nuclear Fuel Cycle - TMS  

Science Conference Proceedings (OSTI)

Feb 12, 2007 ... A compilation of links to websites describing the nuclear fuel cycle. A link to a short overview of the entire cycle is included as well as a ...

166

Fuel Cycle Research and Development Presentation Title  

Science Conference Proceedings (OSTI)

Separations and Waste Form. Campaign Objectives. ?Develop the next generation of fuel cycle separation and waste management technologies that enable a.

167

2012 Fuel Cycle MPACT Working Group  

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

meeting is to review findings and help advance research and development in the Fuel Cycle Materials Protection, Accounting and Control Technologies area. It will include a campaign...

168

DECONTAMINATION OF NEUTRON-IRRADIATED REACTOR FUEL  

DOE Patents (OSTI)

A pyrometallurgical method of decontaminating neutronirradiated reactor fuel is presented. In accordance with the invention, neutron-irradiated reactor fuel may be decontaminated by countercurrently contacting the fuel with a bed of alkali and alkaine fluorides under an inert gas atmosphere and inductively melting the fuel and tracking the resulting descending molten fuel with induction heating as it passes through the bed. By this method, a large, continually fresh surface of salt is exposed to the descending molten fuel which enhances the efficiency of the scrubbing operation.

Buyers, A.G.; Rosen, F.D.; Motta, E.E.

1959-12-22T23:59:59.000Z

169

Proliferation resistant fuel for pebble bed modular reactors  

SciTech Connect

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

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

2012-07-01T23:59:59.000Z

170

International Nuclear Fuel Cycle Fact Book. Revision 5  

SciTech Connect

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

171

International nuclear fuel cycle fact book. Revision 4  

SciTech Connect

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

172

United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support  

SciTech Connect

The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

Douglas Morrell

2011-03-01T23:59:59.000Z

173

Uncertainty Analyses of Advanced Fuel Cycles  

SciTech Connect

The Department of Energy is developing technology, experimental protocols, computational methods, systems analysis software, and many other capabilities in order to advance the nuclear power infrastructure through the Advanced Fuel Cycle Initiative (AFDI). Our project, is intended to facilitate will-informed decision making for the selection of fuel cycle options and facilities for development.

Laurence F. Miller; J. Preston; G. Sweder; T. Anderson; S. Janson; M. Humberstone; J. MConn; J. Clark

2008-12-12T23:59:59.000Z

174

VISION: Verifiable Fuel Cycle Simulation Model  

SciTech Connect

The nuclear fuel cycle consists of a set of complex components that work together in unison. In order 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. The Advanced Fuel Cycle Initiative’s systems analysis group 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.

Jacob Jacobson; A. M. Yacout; Gretchen Matthern; Steven Piet; David Shropshire; Tyler Schweitzer

2010-11-01T23:59:59.000Z

175

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

176

Assessment of Hatch 1, Cycle 21 Fuel Failures  

Science Conference Proceedings (OSTI)

On 19 October 2003, Hatch Unit One located near Baxley, Georgia, experienced six duty-related fuel failures following a control blade notching adjustment. The Hatch 1 reactor was shut down for the End of Cycle 21 refueling outage on 14 February 2004. The duty associated with the Hatch 1 notching event was analyzed in detail and found to be relatively high. Global Nuclear Fuel (GNF), in collaboration with EPRI Fuel Reliability Program and Southern Nuclear Company, sponsored a hot cell examination to estab...

2008-12-08T23:59:59.000Z

177

Evaluation of DD and DT fusion fuel cycles for different fusion-fission energy systems  

SciTech Connect

A study has been carried out in order to investigate the characteristics of an energy system to produce a new source of fissile fuel for existing fission reactors. The denatured fuel cycles were used because it gives additional proliferation resistance compared to other fuel cycles. DT and DD fusion drivers were examined in this study with a thorium or uranium blanket for each fusion driver. Various fuel cycles were studied for light-water and heavy-water reactors. The cost of electricity for each energy system was calculated.

Gohar, Y.

1980-01-01T23:59:59.000Z

178

Nuclear fuel recycling in 4 minutes | Argonne National Laboratory  

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

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

179

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

180

Energy storage for tokamak reactor cycles  

DOE Green Energy (OSTI)

The inherent characteristic of a tokamak reactor requiring periodic plasma quench and reignition introduces the problem of energy storage to permit continuous electrical output to the power grid. The cycle under consideration in this paper is a 1000 second burn followed by a 100 second reignition phase. The physical size of a typical toroidal plasma reaction chamber for a tokamak reactor has been described earlier. The thermal energy storage requirements described in this reference will serve as a basis for much of the ensuing discussion.

Buchanan, C.H.

1979-01-01T23:59:59.000Z

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


181

Technology gap analysis on sodium-cooled reactor fuel handling system supporting advanced burner reactor development.  

Science Conference Proceedings (OSTI)

The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The advanced burner reactor (ABR) is a fast reactor concept which supports the GNEP fuel cycle system. Since the integral fast reactor (IFR) and advanced liquid-metal reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP fast reactor program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology. The fuel-handling system is a key element of any fast reactor design. The major functions of this system are to receive, test, store, and then load fresh fuel into the core; unload from the core; then clean, test, store, and ship spent fuel. Major requirements are that the system must be reliable and relatively easy to maintain. In addition, the system should be designed so that it does not adversely impact plant economics from the viewpoints of capital investment or plant operations. In this gap analysis, information on fuel-handling operating experiences in the following reactor plants was carefully reviewed: EBR-I, SRE, HNPF, Fermi, SEFOR, FFTF, CRBR, EBR-II, DFR, PFR, Rapsodie, Phenix, Superphenix, KNK, SNR-300, Joyo, and Monju. The results of this evaluation indicate that a standardized fuel-handling system for a commercial fast reactor is yet to be established. However, in the past sodium-cooled reactor plants, most major fuel-handling components-such as the rotatable plug, in-vessel fuel-handling machine, ex-vessel fuel transportation cask, ex-vessel sodium-cooled storage, and cleaning stations-have accumulated satisfactory construction and operation experiences. In addition, two special issues for future development are described in this report: large capacity interim storage and transuranic-bearing fuel handling.

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

2009-03-01T23:59:59.000Z

182

Nuclear Fuel Cycle | Department of Energy  

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

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

183

Core Designs and Economic Analyses of Homogeneous Thoria-Urania Fuel in Light Water Reactors  

SciTech Connect

The objective is to develop equilibrium fuel cycle designs for a typical pressurized water reactor (PWR) loaded with homogeneously mixed uranium-thorium dioxide (ThO{sub 2}-UO{sub 2}) fuel and compare those designs with more conventional UO{sub 2} designs.The fuel cycle analyses indicate that ThO{sub 2}-UO{sub 2} fuel cycles are technically feasible in modern PWRs. Both power peaking and soluble boron concentrations tend to be lower than in conventional UO{sub 2} fuel cycles, and the burnable poison requirements are less.However, the additional costs associated with the use of homogeneous ThO{sub 2}-UO{sub 2} fuel in a PWR are significant, and extrapolation of the results gives no indication that further increases in burnup will make thoria-urania fuel economically competitive with the current UO{sub 2} fuel used in light water reactors.

Saglam, Mehmet; Sapyta, Joe J.; Spetz, Stewart W.; Hassler, Lawrence A. [Framatome ANP, Inc. (France)

2004-07-15T23:59:59.000Z

184

Back end of an enduring fuel cycle  

SciTech Connect

An enduring nuclear fuel cycle is an essential part of sustainable consumption, the process whereby world`s riches are consumed in a responsible manner so that future generations can continue to enjoy at least some of them. In many countries, the goal of sustainable development has focused attention on the benefits of nuclear technologies. However, sustenance of the nuclear fuel cycle is dependent on sensible management of all the resources of the fuel cycle, including energy, spent fuels, and all of its side streams. The nuclear fuel cycle for energy production has suffered many traumas since the mid seventies. The common basis of technologies producing nuclear explosives and consumable nuclear energy has been a preoccupation for some, predicament for others, and a perception problem for many. It is essential to reestablish a reliable back end of the nuclear fuel cycle that can sustain the resource requirements of an enduring full cycle. This paper identifies some pragmatic steps necessary to reverse the trend and to maintain a necessary fuel cycle option for the future.

Pillay, K.K.S.

1998-03-01T23:59:59.000Z

185

International Nuclear Fuel Cycle Fact Book  

Science Conference Proceedings (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

186

Production and Handling Slide 1: The Uranium Fuel Cycle  

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

and Handling The Uranium Fuel Cycle Skip Presentation Navigation Next Slide Last Presentation Table of Contents The Uranium Fuel Cycle Refer to caption below for image...

187

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

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

15 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,...

188

Report of the Fuel Cycle Research and Development Subcommittee...  

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

Report of the Fuel Cycle Research and Development Subcommittee of the Nuclear Energy Advisory Committee Report of the Fuel Cycle Research and Development Subcommittee of the...

189

NETL - Petroleum-Based Fuels Life Cycle Greenhouse Gas Analysis...  

Open Energy Info (EERE)

Petroleum-Based Fuels Life Cycle Greenhouse Gas Analysis 2005 Baseline Model Jump to: navigation, search Name NETL - Petroleum-Based Fuels Life Cycle Greenhouse Gas Analysis 2005...

190

Projections of Full-Fuel-Cycle Energy and Emissions Metrics  

E-Print Network (OSTI)

2012a. “Analysis & Projections - Models & Documentation. ”Projections of Full-Fuel-Cycle Energy and Emissions MetricsGovernment purposes. Projections of Full-Fuel-Cycle Energy

Coughlin, Katie

2013-01-01T23:59:59.000Z

191

FUEL PROGRAMMING FOR SODIUM GRAPHITE REACTORS  

SciTech Connect

The effect of fuel programming, i.e., the scheme used for changing fuel in a core, on the reactivity and specific power of a sodium graphite reactor is discussed Fuel programs considered Include replacing fuel a core-load at a time or a radial zone at a time, replacing fuel to manutain the same average exposure of fuel elements throughout the core, and replacing and transferring fuel elements to maintain more highly exposed fuel in the center or at the periphery of the core. Flux and criticality calculations show the degree of power flattening and the concurrent decrease in effective multiplication which results from maintaining more exposed fuel toward the core center. Corverse effects are shown for the case of maintaining more exposed fuel near the core periphery. The excess reactivity which must be controlled in the various programs is considered. Illustrative schedules for implementing each of these programs in an SGR are presented. (auth)

Connolly, T.J.

1959-10-15T23:59:59.000Z

192

NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM  

DOE Patents (OSTI)

This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

Moore, W.T.

1958-09-01T23:59:59.000Z

193

HTGR Technology Family Assessment for a Range of Fuel Cycle Missions  

SciTech Connect

This report examines how the HTGR technology family can provide options for the once through, modified open cycle (MOC), or full recycle fuel cycle strategies. The HTGR can serve all the fuel cycle missions that an LWR can; both are thermal reactors. Additional analyses are warranted to determine if HTGR “full recycle” service could provide improved consumption of transuranic (TRU) material than LWRs (as expected), to analyze the unique proliferation resistance issues associated with the “pebble bed” approach, and to further test and analyze methods to separate TRISO-coated fuel particles from graphite and/or to separate used HTGR fuel meat from its TRISO coating. The feasibility of these two separation issues is not in doubt, but further R&D could clarify and reduce the cost and enable options not adequately explored at present. The analyses here and the now-demonstrated higher fuel burnup tests (after the illustrative designs studied here) should enable future MOC and full recycle HTGR concepts to more rapidly consume TRU, thereby offering waste management advantages. Interest in “limited separation” or “minimum fuel treatment” separation approaches motivates study of impurity-tolerant fuel fabrication. Several issues are outside the scope of this report, including the following: thorium fuel cycles, gas-cooled fast reactors, the reliability of TRISO-coated particles (billions in a reactor), and how soon any new reactor or fuel type could be licensed and then deployed and therefore impact fuel cycle performance measures.

Steven J. Piet; Samuel E. Bays; Nick Soelberg

2010-08-01T23:59:59.000Z

194

Incorporation of excess weapons material into the IFR fuel cycle  

SciTech Connect

The Integral Fast Reactor (IFR) provides both a diversion resistant closed fuel cycle for commercial power generation and a means of addressing safeguards concerns related to excess nuclear weapons material. Little head-end processing and handling of dismantled warhead materials is required to convert excess weapons plutonium (Pu) to IFR fuel and a modest degree of proliferation protection is available immediately by alloying weapons Pu to an IFR fuel composition. Denaturing similar to that of spent fuel is obtained by short cycle (e.g. 45 day) use in an IFR reactor, by mixing which IFR recycle fuel, or by alloying with other spent fuel constituents. Any of these permanent denaturings could be implemented as soon as an operating IFR and/or an IFR recycle capability of reasonable scale is available. The initial Pu charge generated from weapons excess Pu can then be used as a permanent denatured catalyst, enabling the IFR to efficiently and economically generate power with only a natural or depleted uranium feed. The Pu is thereafter permanently safeguarded until consumed, with essentially none going to a waste repository.

Hannum, W.H.; Wade, D.C.

1993-09-01T23:59:59.000Z

195

Reactor Physics Assessment of the Inclusion of Unseparated Neptunium in MOX Reactor Fuel  

Science Conference Proceedings (OSTI)

Reducing the number of actinide separation streams in a spent fuel recovery process would reduce the cost and complexity of the process, and lower the quantity and numbers of solvents needed. It is more difficult and costly to separate Np and recombine it with Am-Cm prior to co-conversion than to simply co-strip it with the U-Pu-Np. Inclusion of the Np in mixed oxide (MOX) fuel for light water reactor (LWR) applications should not seriously affect the operating behavior of the reactor, nor should it pose insurmountable fuel design issues. In this work, the U, Pu, and Np from typical discharged and cooled PWR spent nuclear fuel are assumed to be used together in the preparation of MOX fuel for use in a pressurized water reactor (PWR). The reactor grade Pu isotopic vector is used in the model and the relative mass ratio of the Pu and Np content (Np/Pu mass is 0.061) from the cooled spent fuel is maintained but the overall Pu-Np MOX wt% is adjusted with respect to the U content (assumed to be at 0.25 wt% 235U enrichment) to offset reactivity and cycle length effects. The SCALE 5.1 scientific package (especially modules TRITON, NEWT, ORIGEN-S, ORIGEN-ARP) was used for the calculations presented in this paper. A typical Westinghouse 17x17 fuel assembly design was modeled at nominal PWR operating conditions. It was seen that U-Pu-Np MOX fuel with NpO2 and PuO2 representing 11.5wt% of the total MOX fuel would be similar to standard MOX fuel in which PuO2 is 9wt% of the fuel. The reactivity, isotopic composition, and neutron and ? sources, and the decay heat details for the discharged MOX fuel are presented and discussed in this paper.

Ellis, Ronald James [ORNL

2009-01-01T23:59:59.000Z

196

Fuel Cycle Research and Development Program  

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

Waste Corporate Board James C. Bresee, ScD, JD Advisory Board Member Office of Nuclear Energy July 29, 2009 July 29, 2009 Fuel Cycle Research and Development DM 195665 2 Outline...

197

Projections of Full-Fuel-Cycle Energy and Emissions Metrics  

E-Print Network (OSTI)

Nuclear Fuel ..to characterize the nuclear fuel cycle (Wu et al. Renewableby the heat content of nuclear fuel. In this analysis we use

Coughlin, Katie

2013-01-01T23:59:59.000Z

198

Hybrid Cycles with Hydrogen as Fuel  

Science Conference Proceedings (OSTI)

The gas turbine and steam turbine combined cycle fueled with hydrogen have an overall high efficiency. The virtues of the supercritical steam turbine, the high temperature gas turbine and the low pressure steam turbine are fully expressed in this system. ... Keywords: gas turbine, new energy, combined cycle, hydrogen energy, thermal efficiency, energy conversion

Jing Rulin; Xu Hong; Hu Sangao; Gao Dan; Guo Xiaodan; Ni Weidou

2009-10-01T23:59:59.000Z

199

Parametric Study of Front-End Nuclear Fuel Cycle Costs Using Reprocessed Uranium  

Science Conference Proceedings (OSTI)

This study evaluates front-end nuclear fuel cycle costs assuming that uranium recovered during the reprocessing of commercial light-water reactor (LWR) spent nuclear fuel is available to be recycled and used in the place of natural uranium. This report explores the relationship between the costs associated with using a natural uranium fuel cycle, in which reprocessed uranium (RepU) is not recycled, with those associated with using RepU.

2010-01-26T23:59:59.000Z

200

Safety analysis of IFR fuel processing in the Argonne National Laboratory Fuel Cycle Facility  

SciTech Connect

The Integral Fast Reactor (IFR) concept developed by Argonne National Laboratory (ANL) includes on-site processing and recycling of discharged core and blanket fuel materials. The process is being demonstrated in the Fuel Cycle Facility (FCF) at ANL`s Idaho site. This paper describes the safety analyses that were performed in support of the FCF program; the resulting safety analysis report was the vehicle used to secure authorization to operate the facility and carry out the program, which is now under way. This work also provided some insights into safety-related issues of a commercial IFR fuel processing facility. These are also discussed.

Charak, I; Pedersen, D.R. [Argonne National Lab., IL (United States); Forrester, R.J.; Phipps, R.D. [Argonne National Lab., Idaho Falls, ID (United States)

1993-09-01T23:59:59.000Z

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


201

Solar Thermochemical Fuels Production: Solar Fuels via Partial Redox Cycles with Heat Recovery  

SciTech Connect

HEATS Project: The University of Minnesota is developing a solar thermochemical reactor that will efficiently produce fuel from sunlight, using solar energy to produce heat to break chemical bonds. The University of Minnesota is envisioning producing the fuel by using partial redox cycles and ceria-based reactive materials. The team will achieve unprecedented solar-to-fuel conversion efficiencies of more than 10% (where current state-of-the-art efficiency is 1%) by combined efforts and innovations in material development, and reactor design with effective heat recovery mechanisms and demonstration. This new technology will allow for the effective use of vast domestic solar resources to produce precursors to synthetic fuels that could replace gasoline.

None

2011-12-19T23:59:59.000Z

202

Cost-effective fuel cycle closure  

SciTech Connect

The U.S. government is moving toward meeting its obligation to accept spent fuel from commercial light water reactors (LWRs) in 1998 by providing an interim storage facility. Site work and analysis of the deep, geologic repository at Yucca Mountain will continue at a reduced level of effort. This provides the time required to reevaluate the use of spent-fuel recycling instead of direct disposal. A preliminary assessment of this option is presented in this paper.

Ehrman, C.S. [Burns & Roe, Inc., Oradell, NJ (United States); Boardman, C.E. [General Electric Company, San Jose, CA (United States)

1995-12-31T23:59:59.000Z

203

Fuel Cycle Technology Documents | Department of Energy  

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

Technology Technology Documents Fuel Cycle Technology Documents January 11, 2013 Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste Issued on January 11, 2013, the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste is a framework for moving toward a sustainable program to deploy an integrated system capable of transporting, storing, and disposing of used nuclear fuel and high-level radioactive waste from civilian nuclear power generation, defense, national security and other activities. October 30, 2012 2012 Fuel Cycle Technologies Annual Review Meeting Transaction Report The United States must continue to ensure improvements and access to this technology so we can meet our economic, environmental and energy security

204

Fuel assembly for nuclear reactors  

DOE Patents (OSTI)

A new and improved fuel assembly is formed to minimize the amount of parasitic structural material wherein a plurality of hollow tubular members are juxtaposed to the fuel elements of the assembly. The tubular members may serve as guide tubes for control elements and are secured to a number of longitudinally spaced grid members along the fuel assembly. The grid members include means thereon engaging each of the fuel elements to laterally position the fuel elements in a predetermined array. Openings in the bottom of each hollow member serve as a shock absorber to cushion shock transmitted to the structure when the control elements are rapidly inserted in their corresponding tubular members.

Creagan, Robert J. (Pitcairn, PA); Frisch, Erling (Pittsburgh, PA)

1977-01-01T23:59:59.000Z

205

PLUTONIUM FUEL PROCESSING AND FABRICATION FOR FAST CERAMIC REACTORS  

SciTech Connect

>A study was made of the processes available for fabrication of plutonium-containing fuel from a fast ceramic reacter, and for chemical reprocessing of irradiated fuel. Radiations from recycled plutonium are evaluated. Adaptation of conventional glove-box handling procedures to the fabrication of recycle plutonium appears practical. It is concluded that acceptable costs are obtainable using moderate extensions of conventional glove- box fabrication methods and wet processing techniques, provided a significant volume of production is available. The minimum economic scale for the preferred chemical reprocessing method, anion exchange, is about 500 Mw(e) of reactor capacity. The minimum scale of economic operation for the fuel refabrication facility corresponds to three 500 Mw(e) reactors, if only steady-state refueling provides the fabrication load. The minimum volume required falls to one 500 Mw(e) reactor, if the continued growth of capacity provides fabrication volume equal to that for refueling. The chemical reprocessing costs obtained range from 0.27 mills/kwh for 1500 Mw(e) of reactor capacity, to 0.10 mills/kwh for 3000 Mw(e) of capacity. The estimated fuel fabrication cost is l/kg of uranium and plutonium in the core region (excluding axial and radial blankets) or .06/ g of plutonium content, When axial blankets, fabricated in the same rods, are included; the combined average is 34/kg of uranium and plutonium. Radial blanket fabrication cost is /kg of uranium. The overall average of all fuel and blankets is /kg of uranium and plutonium. The fabrication cost is 0.29 mills/kwh for a production rate corresponding to 3000 Mw(e) of capacity (or 1500 Mw(e) of capacity plus growth equivalent to one additional reactor core per year). For one 525 Mw(e) reactor, (plus equivalent growth volume) the fabrication cost becomes 0.42 mills/ kwh. (All fuel throughputs are based on fuel life of 100,000 MWD/T.) Using the estimates developed, the total fuel cycle cost for a typical fast reactor design using PuO/sub 2/UO/sub 2/ fuel is estimated to be about 0.9 mills/kwh. (auth)

Zebroski, E.L.; Alter, H.W.; Collins, G.D.

1962-02-01T23:59:59.000Z

206

Fuel Cycle Research and Development Presentation Title  

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

SiC Research for SiC Research for Accident Tolerant Fuels Shannon Bragg-Sitton Idaho National Laboratory Advanced LWR Fuels Technical Lead Advanced Fuels Campaign Advanced LWR Fuels Pathway Lead Light Water Reactor Sustainability Program August 2013 Outline  Overview of DOE SiC research  Severe accident modeling: MELCOR analysis w/SiC  Recent characterization test results - Oxidation kinetics - Irradiation studies - Fuel-clad interactions - Elastic property measurement - Thermal properties - Failure model analysis - Quench testing  Technology development - ASTM standards development - SiC/SiC joining technology 2 SiC Gap Analysis and Feasibility Study  SiC Gap Analysis / Feasibility - Milestone report issued July 30, 2013 - Incorporates results of work funded

207

International nuclear fuel cycle fact book  

Science Conference Proceedings (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

208

Concept for a small, colocated fuel cycle facility for oxide breeder fuels  

SciTech Connect

As part of a United States Department of Energy (USDOE) program to examine innovative liquid-metal reactor (LMR) system designs over the past three years, the Oak Ridge National Laboratory (ORNL) and the Westinghouse Hanford Company (WHC) collaborated on studies of mixed oxide fuel cycle options. A principal effort was an advanced concept for a small integrated fuel cycle colocated with a 1300-MW(e) reactor station. The study provided a scoping design and a basis on which to proceed with implementation of such a facility if future plans so dictate. The facility integrated reprocessing, waste management, and refabrication functions in a single facility of nominal 35-t/year capacity utilizing the latest technology developed in fabrication programs at WHC and in reprocessing at ORNL. The concept was based on many years of work at both sites and extensive design studies of prior years.

Burch, W.D.; Stradley, J.G.; Lerch, R.E.

1987-01-01T23:59:59.000Z

209

Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 3. Alternatives for interim storage and transportation  

SciTech Connect

Volume III of the five-volume report contains information on alternatives for interim storage and transportation. Section titles are: interim storage of spent fuel elements; interim storage of chop-leach fuel bundle residues; tank storage of high-level liquid waste; interim storage of solid non-high-level wastes; interim storage of solidified high-level waste; and, transportation alternatives. (JGB)

1976-05-01T23:59:59.000Z

210

Fuel cycle analysis of once-through nuclear systems.  

SciTech Connect

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

211

Solid oxide fuel cell combined cycles  

DOE Green Energy (OSTI)

The integration of the solid oxide fuel cell and combustion turbine technologies can result in combined-cycle power plants, fueled with natural gas, that have high efficiencies and clean gaseous emissions. Results of a study are presented in which conceptual designs were developed for 3 power plants based upon such an integration, and ranging in rating from 3 to 10 MW net ac. The plant cycles are described and characteristics of key components summarized. Also, plant design-point efficiency estimates are presented as well as values of other plant performance parameters.

Bevc, F.P. [Westinghouse Electric Corp., Orlando, FL (United States). Power Generation Business Unit; Lundberg, W.L.; Bachovchin, D.M. [Westinghouse Electric Corp., Pittsburgh, PA (United States). Science and Technology Center

1996-12-31T23:59:59.000Z

212

AN ANALYSIS OF POWER REACTOR FUEL REPROCESSING  

SciTech Connect

This report presents an analysis of the projected economies and processing capacity requirements for a power reactor fuel reprocessing industry based on the recovery of fertile and fissionable materials from presently proposed power reactors within tbe confines of the continental United 8tates for the next five to ten years. An analysis of the present general state of development of a technology required for such an Industry is given. A summary of results of power reactor reprocessing chemical and engineering development at Oak Ridge National Laboratory from July 1955 through December 1956 is given. (auth)

Culler, F.L. Jr.; Blanco, R.E.; Goeller, H.E.; Watson, C.D.

1957-03-27T23:59:59.000Z

213

Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 2. Alternatives for waste treatment  

DOE Green Energy (OSTI)

Volume II of the five-volume report is devoted to the description of alternatives for waste treatment. The discussion is presented under the following section titles: fuel reprocessing modifications; high-level liquid waste solidification; treatment and immobilization of chop-leach fuel bundle residues; treatment of noncombustible solid wastes; treatment of combustible wastes; treatment of non-high-level liquid wastes; recovery of transuranics from non-high-level wastes; immobilization of miscellaneous non-high-level wastes; volatile radioisotope recovery and off-gas treatment; immobilization of volatile radioisotopes; retired facilities (decontamination and decommissioning); and, modification and use of selected fuel reprocessing wastes. (JGB)

Not Available

1976-05-01T23:59:59.000Z

214

Fuel cell and advanced turbine power cycle  

SciTech Connect

Solar has a vested interest in integration of gas turbines and high temperature fuels (particularly solid oxide fuel cells[SOFC]); this would be a backup for achieving efficiencies on the order of 60% with low exhaust emissions. Preferred cycle is with the fuel cell as a topping system to the gas turbine; bottoming arrangements (fuel cells using the gas turbine exhaust as air supply) would likely be both larger and less efficient unless complex steam bottoming systems are added. The combined SOFC and gas turbine will have an advantage because it will have lower NOx emissions than any heat engine system. Market niche for initial product entry will be the dispersed or distributed power market in nonattainment areas. First entry will be of 1-2 MW units between the years 2000 and 2004. Development requirements are outlined for both the fuel cell and the gas turbine.

White, D.J.

1996-12-31T23:59:59.000Z

215

Optimally moderated nuclear fission reactor and fuel source therefor  

DOE Patents (OSTI)

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

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

2008-07-22T23:59:59.000Z

216

Selection of Isotopes and Elements for Fuel Cycle Analysis  

Science Conference Proceedings (OSTI)

Fuel cycle system analysis simulations examine how the selection among fuel cycle options for reactors, fuel, separation, and waste management impact uranium ore utilization, waste masses and volumes, radiotoxicity, heat to geologic repositories, isotope-dependent proliferation resistance measures, and so forth. Previously, such simulations have tended to track only a few actinide and fission product isotopes, those that have been identified as important to a few criteria from the standpoint of recycled material or waste, taken as a whole. After accounting for such isotopes, the residual mass is often characterized as “fission product other” or “actinide other”. However, detailed assessment of separation and waste management options now require identification of key isotopes and residual mass for Group 1A/2A elements (Rb, Cs, Sr, Ba), inert gases (Kr, Xe), halogens (Br, I), lanthanides, transition metals, transuranic (TRU), uranium, actinide decay products. The paper explains the rationale for a list of 81 isotopes and chemical elements to better support separation and waste management assessment in dynamic system analysis models such as Verifiable Fuel Cycle Simulation (VISION)

Steven J. Piet

2009-04-01T23:59:59.000Z

217

Electrorefining {open_quotes}N{close_quotes} reactor fuel  

SciTech Connect

Principles of purifying of uranium metal by electrorefining are reviewed. Metal reactor fuel after irradiation is a form of impure uranium. Dissolution and deposition electrorefining processes were developed for spent metal fuel under the Integral Fast Reactor Program. Application of these processes to the conditioning of spent N-reactor fuel slugs is examined.

Gay, E.C.; Miller, W.E.

1995-02-01T23:59:59.000Z

218

PLUTONIUM METALLIC FUELS FOR FAST REACTORS  

Science Conference Proceedings (OSTI)

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

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

2007-02-07T23:59:59.000Z

219

FUEL CELL/MICRO-TURBINE COMBINED CYCLE  

SciTech Connect

A wide variety of conceptual design studies have been conducted that describe ultra-high efficiency fossil power plant cycles. The most promising of these ultra-high efficiency cycles incorporate high temperature fuel cells with a gas turbine. Combining fuel cells with a gas turbine increases overall cycle efficiency while reducing per kilowatt emissions. This study has demonstrated that the unique approach taken to combining a fuel cell and gas turbine has both technical and economic merit. The approach used in this study eliminates most of the gas turbine integration problems associated with hybrid fuel cell turbine systems. By using a micro-turbine, and a non-pressurized fuel cell the total system size (kW) and complexity has been reduced substantially from those presented in other studies, while maintaining over 70% efficiency. The reduced system size can be particularly attractive in the deregulated electrical generation/distribution environment where the market may not demand multi-megawatt central stations systems. The small size also opens up the niche markets to this high efficiency, low emission electrical generation option.

Larry J. Chaney; Mike R. Tharp; Tom W. Wolf; Tim A. Fuller; Joe J. Hartvigson

1999-12-01T23:59:59.000Z

220

Indirect-fired gas turbine dual fuel cell power cycle  

DOE Patents (OSTI)

The present invention relates generally to an integrated fuel cell power plant, and more specifically to a combination of cycles wherein a first fuel cell cycle tops an indirect-fired gas turbine cycle and a second fuel cell cycle bottoms the gas turbine cycle so that the cycles are thermally integrated in a tandem operating arrangement. The United States Government has rights in this invention pursuant to the employer-employee relationship between the United States Department of Energy and the inventors.

Micheli, P.L.; Williams, M.C.; Sudhoff, F.A.

1998-04-01T23:59:59.000Z

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


221

Fuel Cycle Scenario Definition, Evaluation, and Trade-offs  

SciTech Connect

This report aims to clarify many of the issues being discussed within the AFCI program, including Inert Matrix Fuel (IMF) versus Mixed Oxide (MOX) fuel, single-pass versus multi-pass recycling, thermal versus fast reactors, potential need for transmutation of technetium and iodine, and the value of separating cesium and strontium. It documents most of the work produced by INL, ANL, and SNL personnel under their Simulation, Evaluation, and Trade Study (SETS) work packages during FY2005 and the first half of FY2006. This report represents the first attempt to calculate a full range of metrics, covering all four AFCI program objectives - waste management, proliferation resistance, energy recovery, and systematic management/economics/safety - using a combination of "static" calculations and a system dynamic model, DYMOND. In many cases, we examine the same issue both dynamically and statically to determine the robustness of the observations. All analyses are for the U.S. reactor fleet. This is a technical report, not aimed at a policy-level audience. A wide range of options are studied to provide the technical basis for identifying the most attractive options and potential improvements. Option improvement could be vital to accomplish before the AFCI program publishes definitive cost estimates. Information from this report will be extracted and summarized in future policy-level reports. Many dynamic simulations of deploying those options are included. There are few "control knobs" for flying or piloting the fuel cycle system into the future, even though it is dark (uncertain) and controls are sluggish with slow time response: what types of reactors are built, what types of fuels are used, and the capacity of separation and fabrication plants. Piloting responsibilities are distributed among utilities, government, and regulators, compounding the challenge of making the entire system work and respond to changing circumstances. We identify four approaches that would increase our ability to pilot the fuel cycle system: (1) have a recycle strategy that could be implemented before the 2030-2050 approximate period when current reactors retire so that replacement reactors fit into the strategy, (2) establish an option such as multi-pass blended-core IMF as a downward plutonium control knob and accumulate waste management benefits early, (3) establish fast reactors with flexible conversion ratio as a future control knob that slowly becomes available if/when fast reactors are added to the fleet, and (4) expand exploration of blended assemblies and cores, which appear to have advantages and agility. Initial results suggest multi-pass full-core MOX appears to be a less effective way than multi-pass blended core IMF to manage the fuel cycle system because it requires higher TRU throughput while more slowly accruing waste management benefits. Single-pass recycle approaches for LWRs (we did not study the VHTR) do not meet AFCI program objectives and could be considered a "dead end". Fast reactors appear to be effective options but a significant number of fast reactors must be deployed before the benefit of such strategies can be observed.

Steven J. Piet; Gretchen E. Matthern; Jacob J. Jacobson; Christopher T. Laws; Lee C. Cadwallader; Abdellatif M. Yacout; Robert N. Hill; J. D. Smith; Andrew S. Goldmann; George Bailey

2006-08-01T23:59:59.000Z

222

Natural Fueling of a Tokamak Fusion Reactor  

E-Print Network (OSTI)

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

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

2009-01-01T23:59:59.000Z

223

NEUTRONIC REACTOR AND FUEL ELEMENT THEREFOR  

DOE Patents (OSTI)

This patent relates to a reactor design of the type which employs solid fuel elements disposed in channels within the moderator through which channels and around the fuel elements is conveyed a coolant fiuid. The coolant channels are comprised of aluminum tubes extending through a solid moderator such as graphite and the fuel elements are comprised of an elongated solid body of natural uranium jacketed in an aluminum jacket with the ends thereof closed by aluminum caps of substantially greater thickness than the jacket was and in good thermal contact with the fuel material to facilitate the conduction of heat from the central portion of said ends to the coolant surrounding the fuel element to prevent overheating of said central portion.

Szilard, L.; Young, G.J.

1958-03-01T23:59:59.000Z

224

Nuclear reactor core and fuel element therefor  

SciTech Connect

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

Fortescue, P.

1986-02-11T23:59:59.000Z

225

Westinghouse fuel cell combined cycle systems  

DOE Green Energy (OSTI)

Efficiency (voltage) of the solid oxide fuel cell (SOFC) should increase with operating pressure, and a pressurized SOFC could function as the heat addition process in a Brayton cycle gas turbine (GT) engine. An overall cycle efficiency of 70% should be possible. In cogeneration, half of the waste heat from a PSOFC/GT should be able to be captured in process steam and hot water, leading to a fuel effectiveness of about 85%. In order to make the PSOFC/GT a commercial reality, satisfactory operation of the SOFC at elevated pressure must be verified, a pressurized SOFC generator module must be designed, built, and tested, and the combined cycle and parameters must be optimized. A prototype must also be demonstrated. This paper describes progress toward making the PSOFC/GT a reality.

Veyo, S.

1996-12-31T23:59:59.000Z

226

PYROCHEMICAL DECONTAMINATION METHOD FOR REACTOR FUEL  

DOE Patents (OSTI)

A pyro-chemical method is presented for decontaminating neutron irradiated uranium and separating plutonium therefrom by contact in the molten state with a metal chloride salt. Uranium trichloride and uranium tetrachloride either alone or in admixture with alkaline metal and alkaline eanth metal fluorides under specified temperature and specified phase ratio conditions extract substantially all of the uranium from the irradiated uranium fuel together with certain fission products. The phases are then separated leaving purified uranium metal. The uranium and plutonium in the salt phase can be reduced to forin a highly decontaminated uraniumplutonium alloy. The present method possesses advantages for economically decontaminating irradiated nuclear fuel elements since irradiated fuel may be proccessed immediately after withdrawal from the reactor and the uranium need not be dissolved and later reduced to the metallic form. Accordingly, the uranium may be economically refabricated and reinserted into the reactor.

Buyers, A.G.

1959-06-30T23:59:59.000Z

227

FUEL ELEMENT FOR NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Carney, K.G. Jr.

1959-07-14T23:59:59.000Z

228

The FIT 2.0 Model - Fuel-cycle Integration and Tradeoffs  

Science Conference Proceedings (OSTI)

All mass streams from fuel separation and fabrication are products that must meet some set of product criteria – fuel feedstock impurity limits, waste acceptance criteria (WAC), material storage (if any), or recycle material purity requirements such as zirconium for cladding or lanthanides for industrial use. These must be considered in a systematic and comprehensive way. The FIT model and the “system losses study” team that developed it [Shropshire2009, Piet2010b] are steps by the Fuel Cycle Technology program toward an analysis that accounts for the requirements and capabilities of each fuel cycle component, as well as major material flows within an integrated fuel cycle. This will help the program identify near-term R&D needs and set longer-term goals. This report describes FIT 2, an update of the original FIT model.[Piet2010c] FIT is a method to analyze different fuel cycles; in particular, to determine how changes in one part of a fuel cycle (say, fuel burnup, cooling, or separation efficiencies) chemically affect other parts of the fuel cycle. FIT provides the following: Rough estimate of physics and mass balance feasibility of combinations of technologies. If feasibility is an issue, it provides an estimate of how performance would have to change to achieve feasibility. Estimate of impurities in fuel and impurities in waste as function of separation performance, fuel fabrication, reactor, uranium source, etc.

Steven J. Piet; Nick R. Soelberg; Layne F. Pincock; Eric L. Shaber; Gregory M Teske

2011-06-01T23:59:59.000Z

229

Welding austenitic steel clads for fast reactor fuel pins  

SciTech Connect

ABS>From symposium on fuel and elements for fast reactors; Brussels. Belgium (2 Jul 1973). Developmental programs aimed at fabrication of stainless steelclad PuO/sub 2/ fuel pins are described. Information and data are included on welding fast reactor fuel cans, methods of reducing the incidence of weld cracking, effects of weld stresses, and fuel plug design. (JRD)

Papeleux, P.; Flipot, A.J.; Lafontaine, I.

1973-01-01T23:59:59.000Z

230

A framework and methodology for nuclear fuel cycle transparency.  

Science Conference Proceedings (OSTI)

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

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

2006-02-01T23:59:59.000Z

231

Fusion reactors for synthetic fuels  

DOE Green Energy (OSTI)

Some of the types of processes now being considered for synthetic fuels production from fusion energy, together with an example of each type are listed. The process efficiency is defined as the chemical energy in the generated hydrogen (at the higher heating value (HHV)) divided by the total fusion energy release, including alpha particles and secondary neutron reactions in the blanket. Except where specifically noted, both high and low temperature blanket heats are counted as part of total fusion energy release.

Powell, J.R.

1979-01-01T23:59:59.000Z

232

Production and Handling Slide 3: The Uranium Fuel Cycle  

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

First Slide Previous Slide Next Slide Last Presentation Table of Contents The Uranium Fuel Cycle See caption below for image description The second step in the uranium fuel cycle...

233

Production and Handling Slide 23: The Uranium Fuel Cycle  

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

Presentation Table of Contents The Uranium Fuel Cycle Refer to caption below for image description The fourth major step in the uranium fuel cycle is uranium enrichment. Slide 23...

234

Projections of Full-Fuel-Cycle Energy and Emissions Metrics  

E-Print Network (OSTI)

of a Natural Gas Combined-Cycle Power Generation System.combined with separate accounting for the use of energy in fuel production, is referred to as full- fuel- cycle (

Coughlin, Katie

2013-01-01T23:59:59.000Z

235

Fuel Reliability Program: Post-Irradiation Examination and Performance Assessment of ATRIUM-10 BWR Fuel from Browns Ferry-3 Reactor  

Science Conference Proceedings (OSTI)

ATRIUM-10 design (10x10 lattice) fuel was irradiated for one 24-month period during Cycle 12 to 25 MWd/kgU rod-average exposure at Tennessee Valley Authority's Browns Ferry Unit 3 reactor. The project goal was to characterize the behavior of modern boiling water reactor (BWR) fuel at low exposures to assess early-life performance in a well-documented reactor environment. This report includes results from hot cell post-irradiation examinations. In a future Electric Power Research Institute (EPRI) report, ...

2011-06-09T23:59:59.000Z

236

Investigation of Browns Ferry 2 Reactor, Cycle 12 Corrosion Failures: Volume 1: Poolside Examination Results  

Science Conference Proceedings (OSTI)

BWR fuel rods of the GE13 (9x9) design developed leaks during Cycle 12 of the Browns Ferry 2 reactor. This report presents the results of an examination of this fuel and similar, non-leaking assemblies in the storage pool at the Browns Ferry site. The objective was to define the extent of leakage and the conditions associated with the leaking rods.

2004-12-22T23:59:59.000Z

237

World nuclear fuel cycle requirements 1990  

Science Conference Proceedings (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

238

International nuclear fuel cycle fact book. Revision 6  

SciTech Connect

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

239

Study of CANDU thorium-based fuel cycles by deterministic and Monte Carlo methods  

Science Conference Proceedings (OSTI)

In the framework of the Generation IV forum, there is a renewal of interest in self-sustainable thorium fuel cycles applied to various concepts such as Molten Salt Reactors [1, 2] or High Temperature Reactors [3, 4]. Precise evaluations of the U-233 production potential relying on existing reactors such as PWRs [5] or CANDUs [6] are hence necessary. As a consequence of its design (online refueling and D{sub 2}O moderator in a thermal spectrum), the CANDU reactor has moreover an excellent neutron economy and consequently a high fissile conversion ratio [7]. For these reasons, we try here, with a shorter term view, to re-evaluate the economic competitiveness of once-through thorium-based fuel cycles in CANDU [8]. Two simulation tools are used: the deterministic Canadian cell code DRAGON [9] and MURE [10], a C++ tool for reactor evolution calculations based on the Monte Carlo code MCNP [11]. (authors)

Nuttin, A.; Guillemin, P. [LPSC Grenoble ENSPG (France); Courau, T. [EDF R and D Clamart (France); Marleau, G. [Ecole Polytechnique de Montreal (Canada); Meplan, O. [LPSC Grenoble UJF (France); David, S.; Michel-Sendis, F.; Wilson, J. N. [IPN Orsay CNRS (France)

2006-07-01T23:59:59.000Z

240

Maximum Fuel Energy Saving of a Brayton Cogeneration Cycle  

Science Conference Proceedings (OSTI)

An endoreversible Joule-Brayton cogeneration cycle has been optimized with fuel energy saving as an assessment criterion. The effects of power-to-heat ratio, cycle temperature ratio, and user temperature ratio on maximum fuel energy saving and efficiency ... Keywords: cogeneration cycle, fuel energy saving, thermodynamic optimization

Xiaoli Hao; Guoqiang Zhang

2009-10-01T23:59:59.000Z

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


241

Fuel Rod Cooling in Natural Uranium Reactors  

SciTech Connect

An analysis is presented of the transfer of heat from a cylindrical fuel rod surrounded by a fast flowing coolant in an annular duct, with maximum power output limited by fuel rod temperatures, coolant pressure drop and pumping power requirements. A method is also presented for comparing and evaluating various liquid and gaseous coolants within these limitations. The report also shows and discusses some calculated results obtained for the systems considred in the study of natural U reactors for the production of Pu and useful power (NAA-SR-137).

Trilling, C.A.

1952-01-28T23:59:59.000Z

242

FUEL ELEMENT FOR A NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Davidson, J.K.

1963-11-19T23:59:59.000Z

243

International Fuel Services and Commercial Engagement | Department...  

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

International Fuel Services and Commercial Engagement Nuclear Reactor Technologies Fuel Cycle Technologies International Nuclear Energy Policy and Cooperation Bilateral...

244

Fuel handling system for a nuclear reactor  

DOE Patents (OSTI)

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

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

1986-01-01T23:59:59.000Z

245

Financing Strategies for Nuclear Fuel Cycle Facility  

SciTech Connect

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

246

Advanced Reactor Development and Technology - Nuclear Engineering...  

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

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

247

Optimum Discharge Burnup and Cycle Length for PWRs  

Science Conference Proceedings (OSTI)

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

Jeffrey R. Secker; Baard J. Johansen; David L. Stucker; Odelli Ozer; Kostadin Ivanov; Serkan Yilmaz; E. H. Young

248

Safeguarding and Protecting the Nuclear Fuel Cycle  

Science Conference Proceedings (OSTI)

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

Trond Bjornard; Humberto Garcia; William Desmond; Scott Demuth

2010-11-01T23:59:59.000Z

249

NUCLEAR REACTOR AND THERMIONIC FUEL ELEMENT THEREFOR  

DOE Patents (OSTI)

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

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

1963-12-01T23:59:59.000Z

250

LIFE Materials: Fuel Cycle and Repository Volume 11  

Science Conference Proceedings (OSTI)

The fusion-fission LIFE engine concept provides a path to a sustainable energy future based on safe, carbon-free nuclear power with minimal nuclear waste. The LIFE design ultimately offers many advantages over current and proposed nuclear energy technologies, and could well lead to a true worldwide nuclear energy renaissance. When compared with existing and other proposed future nuclear reactor designs, the LIFE engine exceeds alternatives in the most important measures of proliferation resistance and waste minimization. The engine needs no refueling during its lifetime. It requires no removal of fuel or fissile material generated in the LIFE engine. It leaves no weapons-attractive material at the end of life. Although there is certainly a need for additional work, all indications are that the 'back end' of the fuel cycle does not to raise any 'showstopper' issues for LIFE. Indeed, the LIFE concept has numerous benefits: (1) Per unit of electricity generated, LIFE engines would generate 20-30 times less waste (in terms of mass of heavy metal) requiring disposal in a HLW repository than does the current once-through fuel cycle. (2) Although there may be advanced fuel cycles that can compete with LIFE's low mass flow of heavy metal, all such systems require reprocessing, with attendant proliferation concerns; LIFE engines can do this without enrichment or reprocessing. Moreover, none of the advanced fuel cycles can match the low transuranic content of LIFE waste. (3) The specific thermal power of LIFE waste is initially higher than that of spent LWR fuel. Nevertheless, this higher thermal load can be managed using appropriate engineering features during an interim storage period, and could be accommodated in a Yucca-Mountain-like repository by appropriate 'staging' of the emplacement of waste packages during the operational period of the repository. The planned ventilation rates for Yucca Mountain would be sufficient for LIFE waste to meet the thermal constraints of the repository design. (4) A simple, but arguably conservative, estimate for the dose from a repository containing 63,000 MT of spent LIFE fuel would have similar performance to the currently planned Yucca Mountain Repository. This indicates that a properly designed 'LIFE Repository' would almost certainly meet the proposed Nuclear Regulatory Commission standards for dose to individuals, even though the waste in such a repository would have produced 20-30 times more generated electricity than the reference case for Yucca Mountain. The societal risk/benefit ratio for a LIFE repository would therefore be significantly better than for currently planned repositories for LWR fuel.

Shaw, H; Blink, J A

2008-12-12T23:59:59.000Z

251

Tests of prototype salt stripper system for IFR fuel cycle  

Science Conference Proceedings (OSTI)

One of the waste treatment steps for the on-site reprocessing of spent fuel from the Integral Fast Reactor fuel cycles is stripping of the electrolyte salt used in the electrorefining process. This involves the chemical reduction of the actinides and rare earth chlorides forming metals which then dissolve in a cadmium pool. To develop the equipment for this step, a prototype salt stripper system has been installed in an engineering scale argon-filled glovebox. Pumping trails were successful in transferring 90 kg of LiCl-KCl salt containing uranium and rare earth metal chlorides at 500{degree}C from an electrorefiner to the stripper vessel at a pumping rate of about 5 L/min. The freeze seal solder connectors which were used to join sections of the pump and transfer line performed well. Stripping tests have commenced employing an inverted cup charging device to introduce a Cd-15 wt % Li alloy reductant to the stripper vessel.

Carls, E.L.; Blaskovitz, R.J.; Johnson, T.R. [Argonne National Lab., IL (United States); Ogata, T. [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

1993-09-01T23:59:59.000Z

252

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

E-Print Network (OSTI)

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

Fricano, Joseph William

2012-01-01T23:59:59.000Z

253

Preliminary Advanced Test Reactor LEU Fuel Conversion Feasibility Study  

SciTech Connect

The Advanced Test Reactor (ATR) is a high power density, high neutron flux research reactor operating in the United States. The ATR has large irradiation test volumes located in high flux areas. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth with a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2–s. As a result, the ATR is a representative candidate for assessing the necessary modifications and evaluating the subsequent operating effects associated with low-enriched uranium (LEU) fuel conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed for the fuel cycle burnup comparison analysis. Using the current HEU 235U enrichment of 93.0 % as a baseline, an analysis can be performed to determine the LEU uranium density and 235U enrichment required in the fuel meat to yield an equivalent Keff between the HEU core and a LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the 235U loading in the LEU core, such that the differences in Keff between the HEU and LEU core can be minimized for operation at 150 EFPD with a total core power of 115 MW. The Monte-Carlo with ORIGEN-2 (MCWO) method was used to calculate Keff versus EFPDs. The MCWO-calculated results for the LEU case demonstrated adequate excess reactivity such that the LEU core conversion designer should be able to optimize the 235U content of each fuel plate, so that the Keff and relative radial fission heat flux profile are similar to the reference ATR HEU case. However, to demonstrate that the LEU core fuel cycle performance can meet the Upgraded Final Safety Analysis Report (UFSAR) safety requirements, a further study will be required in order to investigate the detailed radial, axial, and azimuthal heat flux profile variations versus EFPDs.

G. S. Chang; R. G. Ambrosek

2005-11-01T23:59:59.000Z

254

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

SciTech Connect

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

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

2012-07-01T23:59:59.000Z

255

A fuel for sub-critical fast reactor  

Science Conference Proceedings (OSTI)

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

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

2012-06-19T23:59:59.000Z

256

Fuel Summary Report: Shippingport Light Water Breeder Reactor  

SciTech Connect

The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

Illum, D.B.; Olson, G.L.; McCardell, R.K.

1999-01-01T23:59:59.000Z

257

Fossil fuel combined cycle power generation method  

SciTech Connect

A method for converting fuel energy to electricity includes the steps of converting a higher molecular weight gas into at least one mixed gas stream of lower average molecular weight including at least a first lower molecular weight gas and a second gas, the first and second gases being different gases, wherein the first lower molecular weight gas comprises H.sub.2 and the second gas comprises CO. The mixed gas is supplied to at least one turbine to produce electricity. The mixed gas stream is divided after the turbine into a first gas stream mainly comprising H.sub.2 and a second gas stream mainly comprising CO. The first and second gas streams are then electrochemically oxidized in separate fuel cells to produce electricity. A nuclear reactor can be used to supply at least a portion of the heat the required for the chemical conversion process.

Labinov, Solomon D. (Knoxville, TN); Armstrong, Timothy R. (Clinton, TN); Judkins, Roddie R. (Knoxville, TN)

2008-10-21T23:59:59.000Z

258

User Guide for VISION 3.4.7 (Verifiable Fuel Cycle Simulation) Model  

SciTech Connect

The purpose of this document is to provide a guide for using the current version of the Verifiable Fuel Cycle Simulation (VISION) model. This is a complex model with many parameters and options; the user is strongly encouraged to read this user guide before attempting to run the model. This model is an R&D work in progress and may contain errors and omissions. It is based upon numerous assumptions. This model is intended to assist in evaluating 'what if' scenarios and in comparing fuel, reactor, and fuel processing alternatives at a systems level. The model is not intended as a tool for process flow and design modeling of specific facilities nor for tracking individual units of fuel or other material through the system. The model is intended to examine the interactions among the components of a fuel system as a function of time varying system parameters; this model represents a dynamic rather than steady-state approximation of the nuclear fuel system. VISION models the nuclear cycle at the system level, not individual facilities, e.g., 'reactor types' not individual reactors and 'separation types' not individual separation plants. Natural uranium can be enriched, which produces enriched uranium, which goes into fuel fabrication, and depleted uranium (DU), which goes into storage. Fuel is transformed (transmuted) in reactors and then goes into a storage buffer. Used fuel can be pulled from storage into either separation or disposal. If sent to separations, fuel is transformed (partitioned) into fuel products, recovered uranium, and various categories of waste. Recycled material is stored until used by its assigned reactor type. VISION is comprised of several Microsoft Excel input files, a Powersim Studio core, and several Microsoft Excel output files. All must be co-located in the same folder on a PC to function. You must use Powersim Studio 8 or better. We have tested VISION with the Studio 8 Expert, Executive, and Education versions. The Expert and Education versions work with the number of reactor types of 3 or less. For more reactor types, the Executive version is currently required. The input files are Excel2003 format (xls). The output files are macro-enabled Excel2007 format (xlsm). VISION 3.4 was designed with more flexibility than previous versions, which were structured for only three reactor types - LWRs that can use only uranium oxide (UOX) fuel, LWRs that can use multiple fuel types (LWR MF), and fast reactors. One could not have, for example, two types of fast reactors concurrently. The new version allows 10 reactor types and any user-defined uranium-plutonium fuel is allowed. (Thorium-based fuels can be input but several features of the model would not work.) The user identifies (by year) the primary fuel to be used for each reactor type. The user can identify for each primary fuel a contingent fuel to use if the primary fuel is not available, e.g., a reactor designated as using mixed oxide fuel (MOX) would have UOX as the contingent fuel. Another example is that a fast reactor using recycled transuranic (TRU) material can be designated as either having or not having appropriately enriched uranium oxide as a contingent fuel. Because of the need to study evolution in recycling and separation strategies, the user can now select the recycling strategy and separation technology, by year.

Jacob J. Jacobson; Robert F. Jeffers; Gretchen E. Matthern; Steven J. Piet; Wendell D. Hintze

2011-07-01T23:59:59.000Z

259

Fuel Cycle Research & Development Documents | Department of Energy  

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

Initiatives » Fuel Cycle Technologies » Fuel Cycle Research & Initiatives » Fuel Cycle Technologies » Fuel Cycle Research & Development » Fuel Cycle Research & Development Documents Fuel Cycle Research & Development Documents November 8, 2011 2011 Fuel Cycle Technologies Annual Review Meeting As the largest domestic source of low-carbon energy, nuclear power is making major contributions toward meeting our nation's current and future energy demands. The United States must continue to ensure improvements and access to this technology so we can meet our economic, environmental and energy security goals. We rely on nuclear energy because it provides a consistent, reliable and stable source of base load electricity with an excellent safety record in the United States. July 11, 2011 Nuclear Separations Technologies Workshop Report

260

Post Irradiation Evaluation of BWR Fuel From Hope Creek Reactor  

Science Conference Proceedings (OSTI)

Occasionally, in some BWRs, fuel pellet washout from a single degraded fuel rod has resulted in high offgas levels that were sufficient to impede the reactor operation. In addition, certain sound fuel rods have exhibited high eddy-current liftoff values during routine poolside measurements. Investigators pursued these two recent BWR fuel issues by performing detailed hotcell examinations on selected fuel rods from the Hope Creek reactor. The results provided insights into the mechanisms involved and poss...

1997-03-12T23:59:59.000Z

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


261

Detection of Anomalous Reactor Activity Using Antineutrino Count Rate Evolution Over the Course of a Reactor Cycle  

E-Print Network (OSTI)

This paper analyzes the sensitivity of antineutrino count rate measurements to changes in the fissile content of civil power reactors. Such measurements may be useful in IAEA reactor safeguards applications. We introduce a hypothesis testing procedure to identify statistically significant differences between the antineutrino count rate evolution of a standard 'baseline' fuel cycle and that of an anomalous cycle, in which plutonium is removed and replaced with an equivalent fissile worth of uranium. The test would allow an inspector to detect anomalous reactor activity, or to positively confirm that the reactor is operating in a manner consistent with its declared fuel inventory and power level. We show that with a reasonable choice of detector parameters, the test can detect replacement of 73 kg of plutonium in 90 days with 95% probability, while controlling the false positive rate at 5%. We show that some improvement on this level of sensitivity may be expected by various means, including use of the method in conjunction with existing reactor safeguards methods. We also identify a necessary and sufficient daily antineutrino count rate to achieve the quoted sensitivity, and list examples of detectors in which such rates have been attained.

Vera Bulaevskaya; Adam Bernstein

2010-09-11T23:59:59.000Z

262

System Losses Study - FIT (Fuel-cycle Integration and Tradeoffs)  

SciTech Connect

This team aimed to understand the broad implications of changes of operating performance and parameters of a fuel cycle component on the entire system. In particular, this report documents the study of the impact of changing the loss of fission products into recycled fuel and the loss of actinides into waste. When the effort started in spring 2009, an over-simplified statement of the objective was “the number of nines” – how would the cost of separation, fuel fabrication, and waste management change as the number of nines of separation efficiency changed. The intent was to determine the optimum “losses” of TRU into waste for the single system that had been the focus of the Global Nuclear Energy Program (GNEP), namely sustained recycle in burner fast reactors, fed by transuranic (TRU) material recovered from used LWR UOX-51 fuel. That objective proved to be neither possible (insufficient details or attention to the former GNEP options, change in national waste management strategy from a Yucca Mountain focus) nor appropriate given the 2009-2010 change to a science-based program considering a wider range of options. Indeed, the definition of “losses” itself changed from the loss of TRU into waste to a generic definition that a “loss” is any material that ends up where it is undesired. All streams from either separation or fuel fabrication are products; fuel feed streams must lead to fuels with tolerable impurities and waste streams must meet waste acceptance criteria (WAC) for one or more disposal sites. And, these losses are linked in the sense that as the loss of TRU into waste is reduced, often the loss or carryover of waste into TRU or uranium is increased. The effort has provided a mechanism for connecting these three Campaigns at a technical level that had not previously occurred – asking smarter and smarter questions, sometimes answering them, discussing assumptions, identifying R&D needs, and gaining new insights. The FIT model has been a forcing function, helping the team in this endeavor. Models don’t like “TBD” as an input, forcing us to make assumptions and see if they matter. A major addition in FY 2010 was exploratory analysis of “modified open fuel” cycles, employing “minimum fuel treatment” as opposed to full aqueous or electrochemical separation treatment. This increased complexity in our analysis and analytical tool development because equilibrium conditions do not appear sustainable in minimum fuel treatment cases, as was assumed in FY 2009 work with conventional aqueous and electrochemical separation. It is no longer reasonable to assume an equilibrium situation exists in all cases.

Steven J. Piet; Nick R. Soelberg; Samuel E. Bays; Robert S. Cherry; Denia Djokic; Candido Pereira; Layne F. Pincock; Eric L. Shaber; Melissa C. Teague; Gregory M. Teske; Kurt G. Vedros

2010-09-01T23:59:59.000Z

263

Analysis of the ATW fuel cycle using the REBUS-3 code system  

Science Conference Proceedings (OSTI)

Partitioning and transmutation strategies are under study in several countries as a means of reducing the long-term hazards of spent fuel and other high-level nuclear waste. Various reactor and accelerator-driven system concepts have been proposed to transmute the long-lived radioactive nuclei of waste into stable or short-lived species. Among these concepts, the accelerator-driven transmutation of waste (ATW) system has been proposed by the Los Alamos National Laboratory for rapid destruction of transuranic actinides and long-lived fission products ({sup 99}Tc and {sup 129}I). The current reference ATW concept employs a subcritical, liquid-metal-cooled, fast-spectrum nuclear subsystem. Because the discharged fuel is recycled, analysis of ATW nuclear performance requires modeling of the external cycle as well as the in-core fuel management. The fuel cycle analysis of ATW can be performed rigorously using Monte Carlo calculations coupled with detailed depletion calculations. However, the inefficiency of this approach makes it impractical, particularly in view of (a) the large number of fuel cycle calculations needed for design optimization and (b) the need to represent complex in-core and out-of-core fuel cycle operations. To meet the need for design-oriented capabilities, tools previously developed for fast reactor calculations are being adapted for application to ATW. Here, the authors describe the extension and application of the REBUS-3 code to ATW fuel cycle analysis.

Yang, W.S.; Khalil, H.S.

1999-07-01T23:59:59.000Z

264

Objectives, Strategies, and Challenges for the Advanced Fuel Cycle Initiative  

Science Conference Proceedings (OSTI)

This paper will summarize the objectives, strategies, and key chemical separation challenges for the Advanced Fuel Cycle Initiative (AFCI). The major objectives are as follows: Waste management - defer the need for a second geologic repository for a century or more, Proliferation resistance - be more resistant than the existing PUREX separation technology or uranium enrichment, Energy sustainability - turn waste management liabilities into energy source assets to ensure that uranium ore resources do not become a constraint on nuclear power, and Systematic, safe, and economic management of the entire fuel cycle. There are four major strategies for the disposal of civilian spent fuel: Once-through - direct disposal of all discharged nuclear fuel, Limited recycle - recycle transuranic elements once and then direct disposal, Continuous recycle - recycle transuranic elements repeatedly, and Sustained recycle - same as continuous except previously discarded depleted uranium is also recycled. The key chemical separation challenges stem from the fact that the components of spent nuclear fuel vary greatly in their influence on achieving program objectives. Most options separate uranium to reduce the weight and volume of waste and the number and cost of waste packages that require geologic disposal. Separated uranium can also be used as reactor fuel. Most options provide means to recycle transuranic (TRU) elements - plutonium (Pu), neptunium (Np), americium (Am), curium (Cm). Plutonium must be recycled to obtain repository, proliferation, and energy recovery benefits. U.S. non-proliferation policy forbids separation of plutonium by itself; therefore, one or more of the other transuranic elements must be kept with the plutonium; neptunium is considered the easiest option. Recycling neptunium also provides repository benefits. Americium recycling is also required to obtain repository benefits. At the present time, curium recycle provides relatively little benefit; indeed, recycling curium in thermal reactors would significantly increase the hazard (hence cost) of the resulting fuel. Most options separate short-lived fission products cesium and strontium to allow them to decay in separate storage facilities tailored to that need, rather than complicate long-term geologic disposal. This can also reduce the number and cost of waste packages requiring geologic disposal. These savings are balanced by costs for separation and recycle systems. Several long-lived fission products, such as technetium-99 and iodine-129 go to geologic disposal in improved waste forms, recognizing that transmutation of these isotopes would be a slow process; however, the program has not precluded their transmutation as a future alternative.

Steven Piet; Brent Dixon; David Shropshire; Robert Hill; Roald Wigeland; Erich Schneider; J. D. Smith

2005-04-01T23:59:59.000Z

265

Indirect-fired gas turbine dual fuel cell power cycle  

DOE Patents (OSTI)

A fuel cell and gas turbine combined cycle system which includes dual fuel cell cycles combined with a gas turbine cycle wherein a solid oxide fuel cell cycle operated at a pressure of between 6 to 15 atms tops the turbine cycle and is used to produce CO.sub.2 for a molten carbonate fuel cell cycle which bottoms the turbine and is operated at essentially atmospheric pressure. A high pressure combustor is used to combust the excess fuel from the topping fuel cell cycle to further heat the pressurized gas driving the turbine. A low pressure combustor is used to combust the excess fuel from the bottoming fuel cell to reheat the gas stream passing out of the turbine which is used to preheat the pressurized air stream entering the topping fuel cell before passing into the bottoming fuel cell cathode. The CO.sub.2 generated in the solid oxide fuel cell cycle cascades through the system to the molten carbonate fuel cell cycle cathode.

Micheli, Paul L. (Sacramento, CA); Williams, Mark C. (Morgantown, WV); Sudhoff, Frederick A. (Morgantown, WV)

1996-01-01T23:59:59.000Z

266

LIFE vs. LWR: End of the Fuel Cycle  

Science Conference Proceedings (OSTI)

The worldwide energy consumption in 2003 was 421 quadrillion Btu (Quads), and included 162 quads for oil, 99 quads for natural gas, 100 quads for coal, 27 quads for nuclear energy, and 33 quads for renewable sources. The projected worldwide energy consumption for 2030 is 722 quads, corresponding to an increase of 71% over the consumption in 2003. The projected consumption for 2030 includes 239 quads for oil, 190 quads for natural gas, 196 quads for coal, 35 quads for nuclear energy, and 62 quads for renewable sources [International Energy Outlook, DOE/EIA-0484, Table D1 (2006) p. 133]. The current fleet of light water reactors (LRWs) provides about 20% of current U.S. electricity, and about 16% of current world electricity. The demand for electricity is expected to grow steeply in this century, as the developing world increases its standard of living. With the increasing price for oil and gasoline within the United States, as well as fear that our CO2 production may be driving intolerable global warming, there is growing pressure to move away from oil, natural gas, and coal towards nuclear energy. Although there is a clear need for nuclear energy, issues facing waste disposal have not been adequately dealt with, either domestically or internationally. Better technological approaches, with better public acceptance, are needed. Nuclear power has been criticized on both safety and waste disposal bases. The safety issues are based on the potential for plant damage and environmental effects due to either nuclear criticality excursions or loss of cooling. Redundant safety systems are used to reduce the probability and consequences of these risks for LWRs. LIFE engines are inherently subcritical, reducing the need for systems to control the fission reactivity. LIFE engines also have a fuel type that tolerates much higher temperatures than LWR fuel, and has two safety systems to remove decay heat in the event of loss of coolant or loss of coolant flow. These features of LIFE are expected to result in a more straightforward licensing process and are also expected to improve the public perception of risk from nuclear power generation, transportation of nuclear materials, and nuclear waste disposal. Waste disposal is an ongoing issue for LWRs. The conventional (once-through) LWR fuel cycle treats unburned fuel as waste, and results in the current fleet of LWRs producing about twice as much waste in their 60 years of operation as is legally permitted to be disposed of in Yucca Mountain. Advanced LWR fuel cycles would recycle the unused fuel, such that each GWe-yr of electricity generation would produce only a small waste volume compared to the conventional fuel cycle. However, the advanced LWR fuel cycle requires chemical reprocessing plants for the fuel, multiple handling of radioactive materials, and an extensive transportation network for the fuel and waste. In contrast, the LIFE engine requires only one fueling for the plant lifetime, has no chemical reprocessing, and has a single shipment of a small amount of waste per GWe-yr of electricity generation. Public perception of the nuclear option will be improved by the reduction, for LIFE engines, of the number of shipments of radioactive material per GWe-yr and the need to build multiple repositories. In addition, LIFE fuel requires neither enrichment nor reprocessing, eliminating the two most significant pathways to proliferation from commercial nuclear fuel to weapons programs.

Farmer, J C; Blink, J A; Shaw, H F

2008-10-02T23:59:59.000Z

267

Characteristics of Spent Fuel from Plutonium Disposition Reactors. Vol. 3: A Westinghouse Pressurized-Water Reactor Design  

Science Conference Proceedings (OSTI)

This report discusses the results of a simulation study involving the burnup of mixed-oxide (MOX) fuel in a Westinghouse pressurized-water reactor (PWR). The MOX was composed of uranium and plutonium oxides, where the plutonium was of weapons-grade composition. The study was part of the Fissile Materials Disposition Program and considered the possibility of fueling commercial reactors with weapons plutonium. The isotopic composition, the activities, and the decay heat, together with the gamma and neutron dose rates are discussed for the spent fuel. For the steady-state situation involving this PWR burning MOX fuel, two burn histories are reported. In one case, an assembly is burned in the reactor for two cycles, and in the second case and assembly is burned for three cycles. Furthermore, assemblies containing wet annular burnable absorbers (WABAs) and assemblies that do not contain WABAs are considered in all cases. The two-cycle cases have a burnup of 35 GWd/t, and the three-cycle cases have a burnup of 52.5 GWd/t.

Murphy, B.D.

1997-07-01T23:59:59.000Z

268

Microsoft Word - Fuel Cycle Subcomm report final v2.docx  

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

of the Fuel Cycle of the Fuel Cycle Subcommittee of NEAC June 15, 2011 Washington, D.C. Members: Burton Richter (Chairman) Darleane Hoffman Raymond Juzaitis Sekazi Mtingwa Ron Omberg Joy Rempe Dominique Warin Fuel Cycle Subcommittee Report 6/15/2011 2 I. Introduction and Summary The Fuel Cycle subcommittee of NEAC met April 25-26 in Albuquerque, New Mexico. The main topics of discussion were the Used Nuclear Fuel (UNF) disposal program, the System Study Program's methodology that is to be used to set priorities for R&D on advanced fuel cycles, and the University Programs. In addition to these, we were briefed on the budget, but have no comments other than a hope for a good outcome and restrict ourselves to general advice until more is known. A current complication in the design of the Fuel Cycle R&D FCRD program is the Blue

269

Spent nuclear fuel discharges from U.S. reactors 1994  

Science Conference Proceedings (OSTI)

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

NONE

1996-02-01T23:59:59.000Z

270

DOE Hydrogen Analysis Repository: Life Cycle Assessment of Hydrogen Fuel  

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

Life Cycle Assessment of Hydrogen Fuel Cell and Gasoline Vehicles Life Cycle Assessment of Hydrogen Fuel Cell and Gasoline Vehicles Project Summary Full Title: Life Cycle Assessment of Hydrogen Fuel Cell and Gasoline Vehicles Project ID: 143 Principal Investigator: Ibrahim Dincer Brief Description: Examines the social, environmental and economic impacts of hydrogen fuel cell and gasoline vehicles. Purpose This project aims to investigate fuel cell vehicles through environmental impact, life cycle assessment, sustainability, and thermodynamic analyses. The project will assist in the development of highly qualified personnel in such areas as system analysis, modeling, methodology development, and applications. Performer Principal Investigator: Ibrahim Dincer Organization: University of Ontario Institute of Technology

271

Future nuclear fuel cycles: prospects and challenges  

Science Conference Proceedings (OSTI)

Solvent extraction has played, from the early steps, a major role in the development of nuclear fuel cycle technologies, both in the front end and back end. Today's stakes in the field of energy enhance further than before the need for a sustainable management of nuclear materials. Recycling actinides appears as a main guideline, as much for saving resources as for minimizing the final waste impact, and many options can be considered. Strengthened by the important and outstanding performance of recent PUREX processing plants, solvent-extraction processes seem a privileged route to meet the new and challenging requirements of sustainable future nuclear systems. (author)

Boullis, Bernard [Commissariat a l'Energie Atomique, Direction de l'Energie Nucleaire, Centre de Saclay, 91191, Gif-sur-Yvette cedex (France)

2008-07-01T23:59:59.000Z

272

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

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

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

273

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

E-Print Network (OSTI)

Framework   for   Nuclear   Fuel   Cycle   Concepts,”  Of   Used   Nuclear   Fuel”,   Nuclear  Engineering  and  Radiotoxicity  of  Spent  Nuclear   Fuel,”   Integrated  

Djokic, Denia

2013-01-01T23:59:59.000Z

274

An improved characterization method for international accountancy measurements of fresh and irradiated mixed oxide (MOX) fuel: helping achieve continual monitoring and safeguards through the fuel cycle  

SciTech Connect

Nuclear fuel accountancy measurements are conducted at several points through the nuclear fuel cycle to ensure continuity of knowledge (CofK) of special nuclear material (SNM). Non-destructive assay (NDA) measurements are performed on fresh fuel (prior to irradiation in a reactor) and spent nuclear fuel (SNF) post-irradiation. We have developed a fuel assembly characterization system, based on the novel concept of 'neutron fingerprinting' with multiplicity signatures to ensure detailed CofK of nuclear fuel through the entire fuel cycle. The neutron fingerprint in this case is determined by the measurement of the various correlated neutron signatures, specific to fuel isotopic composition, and therefore offers greater sensitivity to variations in fissile content among fuel assemblies than other techniques such as gross neutron counting. This neutron fingerprint could be measured at the point of fuel dispatch (e.g. from a fuel fabrication plant prior to irradiation, or from a reactor site post-irradiation), monitored during transportation of the fuel assembly, and measured at a subsequent receiving site (e.g. at the reactor site prior to irradiation, or reprocessing facility post-irradiation); this would confirm that no unexpected changes to the fuel composition or amount have taken place during transportation and/or reactor operations. Changes may indicate an attempt to divert material for example. Here, we present the current state of the practice of fuel measurements for both fresh mixed oxide (MOX) fuel and SNF (both MOX and uranium dioxide). This is presented in the framework of international safeguards perspectives from the US and UK. We also postulate as to how the neutron fingerprinting concept could lead to improved fuel characterization (both fresh MOX and SNF) resulting in: (a) assured CofK of fuel across the nuclear fuel cycle, (b) improved detection of SNM diversion, and (c) greater confidence in safeguards of SNF transportation.

Evans, Louise G [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Tobin, S. J. [Los Alamos National Laboratory; Boyer, B. D. [Los Alamos National Laboratory; Menlove, H. O. [Los Alamos National Laboratory; Schear, M. A. [Los Alamos National Laboratory; Worrall, Andrew [U.K., NNL

2010-11-24T23:59:59.000Z

275

An improved characterization method for international accountancy measurements of fresh and irradiated mixed oxide (MOX) fuel: helping achieve continual monitoring and safeguards through the fuel cycle  

SciTech Connect

Nuclear fuel accountancy measurements are conducted at several points through the nuclear fuel cycle to ensure continuity of knowledge (CofK) of special nuclear material (SNM). Non-destructive assay (NDA) measurements are performed on fresh fuel (prior to irradiation in a reactor) and spent nuclear fuel (SNF) post-irradiation. We have developed a fuel assembly characterization system, based on the novel concept of 'neutron fingerprinting' with multiplicity signatures to ensure detailed CofK of nuclear fuel through the entire fuel cycle. The neutron fingerprint in this case is determined by the measurement of the various correlated neutron signatures, specific to fuel isotopic composition, and therefore offers greater sensitivity to variations in fissile content among fuel assemblies than other techniques such as gross neutron counting. This neutron fingerprint could be measured at the point of fuel dispatch (e.g. from a fuel fabrication plant prior to irradiation, or from a reactor site post-irradiation), monitored during transportation of the fuel assembly, and measured at a subsequent receiving site (e.g. at the reactor site prior to irradiation, or reprocessing facility post-irradiation); this would confirm that no unexpected changes to the fuel composition or amount have taken place during transportation and/ or reactor operations. Changes may indicate an attempt to divert material for example. Here, we present the current state of the practice of fuel measurements for both fresh mixed oxide (MOX) fuel and SNF (both MOX and uranium dioxide). This is presented in the framework of international safeguards perspectives from the US and UK. We also postulate as to how the neutron fingerprinting concept could lead to improved fuel characterization (both fresh MOX and SNF) resulting in: (a) assured CofK of fuel across the nuclear fuel cycle, (b) improved detection of SNM diversion, and (c) greater confidence in safeguards of SNF transportation.

Evans, Louise G [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Tobin, S. J. [Los Alamos National Laboratory; Menlove, H. O. [Los Alamos National Laboratory; Schear, M. A. [Los Alamos National Laboratory; Worrall, Andrew [U.K. NNL

2011-01-13T23:59:59.000Z

276

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

277

Optimization of hydride fueled pressurized water reactor cores  

E-Print Network (OSTI)

This thesis contributes to the Hydride Fuels Project, a collaborative effort between UC Berkeley and MIT aimed at investigating the potential benefits of hydride fuel use in light water reactors (LWRs). This pursuit involves ...

Shuffler, Carter Alexander

2004-01-01T23:59:59.000Z

278

Design and fuel management of PWR cores to optimize the once-through fuel cycle  

SciTech Connect

The once-through fuel cycle has been analyzed to see if there are substantial prospects for improved uranium ore utilization in current light water reactors, with a specific focus on pressurized water reactors. The types of changes which have been examined are: (1) re-optimization of fuel pin diameter and lattice pitch, (2) axial power shaping by enrichment gradation in fresh fuel, (3) use of 6-batch cores with semi-annual refueling, (4) use of 6-batch cores with annual refueling, hence greater extended (approximately doubled) burnup, (5) use of radial reflector assemblies, (6) use of internally heterogeneous cores (simple seed/blanket configurations), (7) use of power/temperature coastdown at the end of life to extend burnup, (8) use of metal or diluted oxide fuel, (9) use of thorium, and (10) use of isotopically separated low sigma/sub a/ cladding material. State-of-the-art LWR computational methods, LEOPARD/PDQ-7/FLARE-G, were used to investigate these modifications.

Fujita, E.K.; Driscoll, M.J.; Lanning, D.D.

1978-08-01T23:59:59.000Z

279

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

E-Print Network (OSTI)

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

Boscher, Thomas

2005-01-01T23:59:59.000Z

280

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

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


281

Program on Technology Innovation: Readiness of Existing and New U.S. Reactors for Mixed-Oxide (MOX) Fuel  

Science Conference Proceedings (OSTI)

Expanding interest in nuclear power and advanced fuel cycles indicate that use of mixed-oxide (MOX) fuel in the current and new U.S. reactor fleet could become an option for utilities in the coming decades. In light of this renewed interest, EPRI has reviewed the substantial knowledge base on MOX fuel irradiation in light water reactors (LWRs). The goal was to evaluate the technical feasibility of MOX fuel use in the U.S. reactor fleet for both existing and advanced LWR designs (Generation III/III+).

2009-05-29T23:59:59.000Z

282

Microsoft Word - Fuel Cycle Potential Waste Inventory for Disposition...  

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

Fuel Cycle Potential Waste Inventory for Disposition Prepared for U.S. Department of Energy Used Nuclear Fuel Joe T. Carter, SRNL Alan J. Luptak, INL Jason Gastelum, PNNL Christine...

283

Summary of Off-Normal Events in US Fuel Cycle Facilities for AFCI Applications  

SciTech Connect

This report is a collection and review of system operation and failure experiences for facilities comprising the fission reactor fuel cycle, with the exception of reactor operations. This report includes mines, mills, conversion plants, enrichment plants, fuel fabrication plants, transportation of fuel materials between these centers, and waste storage facilities. Some of the facilities discussed are no longer operating; others continue to produce fuel for the commercial fission power plant industry. Some of the facilities discussed have been part of the military’s nuclear effort; these are included when the processes used are similar to those used for commercial nuclear power. When reading compilations of incidents and accidents, after repeated entries it is natural to form an opinion that there exists nothing but accidents. For this reason, production or throughput values are described when available. These adverse operating experiences are compiled to support the design and decisions needed for the Advanced Fuel Cycle Initiative (AFCI). The AFCI is to weigh options for a new fission reactor fuel cycle that is efficient, safe, and productive for US energy security.

L. C. Cadwallader; S. J. Piet; S. O. Sheetz; D. H. McGuire; W. B. Boore

2005-09-01T23:59:59.000Z

284

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

SciTech Connect

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

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

1989-01-01T23:59:59.000Z

285

NMSS handbook for decommissioning fuel cycle and materials licensees  

Science Conference Proceedings (OSTI)

The US Nuclear Regulatory Commission amended its regulations to set forth the technical and financial criteria for decommissioning licensed nuclear facilities. These regulations were further amended to establish additional recordkeeping requirements for decommissioning; to establish timeframes and schedules for the decommissioning; and to clarify that financial assurance requirements must be in place during operations and updated when licensed operations cease. Reviews of the Site Decommissioning Management Plan (SDMP) program found that, while the NRC staff was overseeing the decommissioning program at nuclear facilities in a manner that was protective of public health and safety, progress in decommissioning many sites was slow. As a result NRC determined that formal written procedures should be developed to facilitate the timely decommissioning of licensed nuclear facilities. This handbook was developed to aid NRC staff in achieving this goal. It is intended to be used as a reference document to, and in conjunction with, NRC Inspection Manual Chapter (IMC) 2605, ``Decommissioning Inspection Program for Fuel Cycle and Materials Licensees.`` The policies and procedures discussed in this handbook should be used by NRC staff overseeing the decommissioning program at licensed fuel cycle and materials sites; formerly licensed sites for which the licenses were terminated; sites involving source, special nuclear, or byproduct material subject to NRC regulation for which a license was never issued; and sites in the NRC`s SDMP program. NRC staff overseeing the decommissioning program at nuclear reactor facilities subject to regulation under 10 CFR Part 50 are not required to use the procedures discussed in this handbook.

Orlando, D.A.; Hogg, R.C.; Ramsey, K.M. [and others

1997-03-01T23:59:59.000Z

286

Production and Handling Slide 37: The Uranium Fuel Cycle  

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

Table of Contents The Uranium Fuel Cycle Refer to caption below for image description The enrichment process generates two streams of uranium hexafluoride, one enriched in...

287

Fuel cycle assessment: A compendium of models, methodologies, and approaches  

SciTech Connect

The purpose of this document is to profile analytical tools and methods which could be used in a total fuel cycle analysis. The information in this document provides a significant step towards: (1) Characterizing the stages of the fuel cycle. (2) Identifying relevant impacts which can feasibly be evaluated quantitatively or qualitatively. (3) Identifying and reviewing other activities that have been conducted to perform a fuel cycle assessment or some component thereof. (4) Reviewing the successes/deficiencies and opportunities/constraints of previous activities. (5) Identifying methods and modeling techniques/tools that are available, tested and could be used for a fuel cycle assessment.

Not Available

1994-07-01T23:59:59.000Z

288

Production and Handling Slide 5: The Uranium Fuel Cycle  

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

Refer to caption below for image description The third step in the uranium fuel cycle involves the conversion of "yellowcake" to uranium hexafluoride (UF6), the chemical form...

289

Production and Handling Slide 43: The Uranium Fuel Cycle  

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

Presentation Table of Contents The Uranium Fuel Cycle Refer to caption below for image description Enriched uranium hexafluoride, generally containing 3 to 5% uranium-235, is sent...

290

Microsoft Word - Fuel Cycle Subcomm report final v2.docx  

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

of the Fuel Cycle Subcommittee of NEAC June 15, 2011 Washington, D.C. Members: Burton Richter (Chairman) Darleane Hoffman Raymond Juzaitis Sekazi Mtingwa Ron Omberg Joy Rempe...

291

Nuclear Weapons Proliferation and the Civilian Nuclear Fuel Cycle...  

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

Engineering Sciences October 12-14, 2011, Northwestern University Evanston, Illinois Nuclear Weapons Proliferation and the Civilian Nuclear Fuel Cycle: Understanding and Reducing...

292

Description of alternative steady-state fuel cycles  

SciTech Connect

This study provides a first cut analysis for the FRAD program of a range of reference, steady-state, fresh and spent fuel compositions for the development of alternative fuels refabrication technology. Included are the resource requirements and separative work requirements and the material flows for each fuel cycle evaluated. However, since steady-state represents only a portion of the complete fuel cycle, a more in depth evaluation of each alternative fuel cycle will follow this analysis. Each of the fuel types investigated is composed of either plutonium-uranium (Pu-U), denatured uranium-thorium (DU-Th), plutonium-thorium (Pu-Th), highly enriched uranium-thorium (HEU-Th) or low enriched uranium (LEU). Seven ''closed cycles'' were formed by coupling two or more of the above fuel types. The closed cycle concept assumes that all fissile material recovered from spent fuel is either recycled into fresh fuel, or retired to waste when its net reactivity worth is equal to or less than tails equivalence. Additional fissile material required as makeup is introduced to the system from the enrichment cascade only. Each closed system presented in this study simulates the production of 1000 MWe in steady-state operation. The findings of this preliminary study indicated that, at equilibrium, those closed cycles which employ DU-Th or HEU-Th as the primary fuel are more efficient with respect to resource consumption than those cycles where LEU is used as the primary fuel.

Boegel, A.J.; Merrill, E.T.; Newman, D.F.; Nolan, A.M.

1978-11-01T23:59:59.000Z

293

Tools for supercritical carbon dioxide cycle analysis and the cycle's applicability to sodium fast reactors  

E-Print Network (OSTI)

The Sodium-Cooled Fast Reactor (SFR) and the Supercritical Carbon Dioxide (S-C0?) Recompression cycle are two technologies that have the potential to impact the power generation landscape of the future. In order for their ...

Ludington, Alexander R. (Alexander Rockwell)

2009-01-01T23:59:59.000Z

294

CHEMICAL ASPECTS OF PELLET-CLADDING INTERACTION IN LIGHT WATER REACTOR FUEL ELEMENTS  

E-Print Network (OSTI)

ANS/ENS Topical Meeting on Reactor Safety Aspects of FuelINTERACTION IN LiaiT-WATER-REACTOR FUEL ELEMENTS by D. R.PCI) in light water reactor fuel elements, the chemical

Olander, D.R.

2010-01-01T23:59:59.000Z

295

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

E-Print Network (OSTI)

Anthony   V.   Guide  Nuclear  Reactors.   University   of  of   fuel   for   nuclear   reactors—create   wastes  Level  Waste   nuclear reactors, and subsequent utilization

Djokic, Denia

2013-01-01T23:59:59.000Z

296

The Advanced High-Temperature Reactor (AHTR) for Producing Hydrogen to Manufacture Liquid Fuels  

DOE Green Energy (OSTI)

Conventional world oil production is expected to peak within a decade. Shortfalls in production of liquid fuels (gasoline, diesel, and jet fuel) from conventional oil sources are expected to be offset by increased production of fuels from heavy oils and tar sands that are primarily located in the Western Hemisphere (Canada, Venezuela, the United States, and Mexico). Simultaneously, there is a renewed interest in liquid fuels from biomass, such as alcohol; but, biomass production requires fertilizer. Massive quantities of hydrogen (H2) are required (1) to convert heavy oils and tar sands to liquid fuels and (2) to produce fertilizer for production of biomass that can be converted to liquid fuels. If these liquid fuels are to be used while simultaneously minimizing greenhouse emissions, nonfossil methods for the production of H2 are required. Nuclear energy can be used to produce H2. The most efficient methods to produce H2 from nuclear energy involve thermochemical cycles in which high-temperature heat (700 to 850 C) and water are converted to H2 and oxygen. The peak nuclear reactor fuel and coolant temperatures must be significantly higher than the chemical process temperatures to transport heat from the reactor core to an intermediate heat transfer loop and from the intermediate heat transfer loop to the chemical plant. The reactor temperatures required for H2 production are at the limits of practical engineering materials. A new high-temperature reactor concept is being developed for H2 and electricity production: the Advanced High-Temperature Reactor (AHTR). The fuel is a graphite-matrix, coated-particle fuel, the same type that is used in modular high-temperature gas-cooled reactors (MHTGRs). The coolant is a clean molten fluoride salt with a boiling point near 1400 C. The use of a liquid coolant, rather than helium, reduces peak reactor fuel and coolant temperatures 100 to 200 C relative to those of a MHTGR. Liquids are better heat transfer fluids than gases and thus reduce three temperature losses in the system associated with (1) heat transfer from the fuel to the reactor coolant, (2) temperature rise across the reactor core, and (3) heat transfer across the heat exchangers between the reactor and H2 production plant. Lowering the peak reactor temperatures and thus reducing the high-temperature materials requirements may make the AHTR the enabling technology for low-cost nuclear hydrogen production.

Forsberg, C.W.; Peterson, P.F.; Ott, L.

2004-10-06T23:59:59.000Z

297

Extended Power Uprates and 2-yr Cycles for BWRs - Where Do We Go from Here?  

Science Conference Proceedings (OSTI)

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

Craig Brown; Ken Hartley; Jim Hulsman

298

PROCESSING OF MOLTEN SALT POWER REACTOR FUEL  

SciTech Connect

ABS> Fuel reprocessing methods are being investigated for molten salt nuclear reactors which use LiF--BeF/sub 2/ salt as a solvent for UF/sub 4/ and ThF/sub 4/. A liquid HF dissolution procedure coupled with fluorination has been developed for recovery of the uranium and LiF- BeF/sub 2/ solvent salt which is highly enriched in Li/sup 7/. The recovered salt is decontaminated in the process from the major reactor poisons; namely, rare earths and neptunium. A brief investigation of alternate methods, including oxide precipitation, partial freezing, and metal reduction, indicated that such methods may give some separation of the solvent salt from reactor poisons, but they do not appear to be sufficiently quantitative for a simple processing operation. Solubilities of LiF and BeF/sub 2/ in aqueous 70t0 100% HF are presented. The BeF/sub 2/ solubility is appreciably increased in the presence of water and large amounts of LiF. Salt solubilities of 150 g/liter are attainable. Tracer experiments indicate that rare earth solubilities, relative to LiF-- BeF/sub 2/ solvent salt solubility, increase from about 10/sup -4/ mole% in 98% HF to 0.003 mole% in 80% HF. Fluorination of uranium from LiF--BeF/sub 2/ salt was demonstrated. This appears feasible also for the recovery of the relatively small ccncentration of uranium produced in the LiF- BeF/sub 2/ThF/sub 4/ blanket. A proposed chemical flowsheet is presented on the basis of this exploratory work as applied to the semicontinuous processing of a 600 Mw power reactor. (auth)

Campbell, D.O.; Cathers, G.I.

1959-04-01T23:59:59.000Z

299

FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL  

Science Conference Proceedings (OSTI)

The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: 1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs 2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs 3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs 4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs.

Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

2009-03-10T23:59:59.000Z

300

Evaluation of Fuel Clad Corrosion Product Deposits and Circulating Corrosion Products in Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

Many pressurized water reactors (PWRs) have experienced negative consequences resulting from build-up of corrosion product deposits (crud) on fuel cladding. The negative consequences include unplanned shifts in core power (axial offset anomaly, or AOA), fuel cladding failure, anomalous shutdown chemistry, and elevated ex-core radiation fields. These problems have grown more common as PWRs have moved toward higher 235U enrichments and higher duty cores needed for extended cycle operation. This report expl...

2004-12-08T23:59:59.000Z

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While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
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301

2012 Fuel Cycle MPACT Working Group  

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

Site Site Hydrogen Research Center 301 Gateway Drive Aiken, SC 29803 Accommodations Country Inn & Suites Aiken 3270 Whiskey Road Aiken, SC 29803 (803) 649-4024 RESERVATIONS: The cut-off date for guest room block reservations is Wednesday, February 22, 2012. We have a block of rooms reserved at this hotel at the government per diem rate of $86.00 per night. Please reference DOE -NE Fuel Cycle MPACT Working Group Meeting when making your reservations to the get the government rate. Reservations will be by individual call-in, per your institutional protocol. Here is a listing of other hotels that offer government room rates. Please note that we do not have rooms reserved at the list locations, only Country Inn & Suites in Aiken. Maps Maps to SRNL from Columbia, Aiken, and Augusta

302

A Review of Thorium Utilization as an option for Advanced Fuel Cycle--Potential Option for Brazil in the Future  

SciTech Connect

Since the beginning of Nuclear Energy Development, Thorium was considered as a potential fuel, mainly due to the potential to produce fissile uranium 233. Several Th/U fuel cycles, using thermal and fast reactors were proposed, such as the Radkwoski once through fuel cycle for PWR and VVER, the thorium fuel cycles for CANDU Reactors, the utilization in Molten Salt Reactors, the utilization of thorium in thermal (AHWR), and fast reactors (FBTR) in India, and more recently in innovative reactors, mainly Accelerator Driven System, in a double strata fuel cycle. All these concepts besides the increase in natural nuclear resources are justified by non proliferation issues (plutonium constrain) and the waste radiological toxicity reduction. The paper intended to summarize these developments, with an emphasis in the Th/U double strata fuel cycle using ADS. Brazil has one of the biggest natural reserves of thorium, estimated in 1.2 millions of tons of ThO{sub 2}, as will be reviewed in this paper, and therefore R&D programs would be of strategically national interest. In fact, in the past there was some projects to utilize Thorium in Reactors, as the ''Instinto/Toruna'' Project, in cooperation with France, to utilize Thorium in Pressurized Heavy Water Reactor, in the mid of sixties to mid of seventies, and the thorium utilization in PWR, in cooperation with German, from 1979-1988. The paper will review these initiatives in Brazil, and will propose to continue in Brazil activities related with Th/U fuel cycle.

Maiorino, J.R.; Carluccio, T.

2004-10-03T23:59:59.000Z

303

Analysis of the ATW fuel cycle using the REBUS-3 code system.  

Science Conference Proceedings (OSTI)

Partitioning and transmutation strategies are under study in several countries as a means of reducing the long-term hazards of spent fuel and other high-level nuclear waste. Various reactor and accelerator-driven system concepts have been proposed to transmute the long-lived radioactive nuclei of waste into stable or short-lived species. Among these concepts, the accelerator-driven transmutation of waste (ATW) system has been proposed by LANL for rapid destruction of transuranic actinides and long-lived fission products ({sup 99}Tc and {sup 129}I).The current reference ATW concept employs a subcritical, liquid metal cooled, fast-spectrum nuclear subsystem. Because the discharged fuel is recycled, analysis of ATW nuclear performance requires modeling of the external cycle as well as the in-core fuel management. The fuel cycle analysis of ATW can be performed rigorously using Monte Carlo calculations coupled with detailed depletion calculations. However, the inefficiency of this approach makes it impractical, particularly in view of (a) the large number of fuel cycle calculations needed for design optimization and (b) the need to represent complex in-core and out-of-core fuel cycle operations. To meet the need for design-oriented capabilities, tools previously developed for fast reactor calculations are being adapted for application to ATW. Here we describe the extension and application of the REBUS-3 code to ATW fuel cycle analysis. This code has been extensively used for advanced liquid metal reactor design and analysis and validated against EBR-II irradiation data.

Khalil, H. S.; Yang, W. S.

1999-06-25T23:59:59.000Z

304

Fast power cycle for fusion reactors  

SciTech Connect

The unique, deep penetration capability of 14 MeV neutrons produced in DT fusion reactions allows the generation of very high temperature working fluid temperatures in a thermal power cycle. In the FAST (Fusion Augmented Steam Turbine) power cycle steam is directly superheated by the high temperature ceramic refractory interior of the blanket, after being generated by heat extracted from the relatively cool blanket structure. The steam is then passed to a high temperature gas turbine for power generation. Cycle studies have been carried out for a range of turbine inlet temperatures (1600/sup 0/F to 3000/sup 0/F (870 to 1650/sup 0/C)), number of reheats, turbine mechanical efficiency, recuperator effectiveness, and system pressure losses. Gross cycle efficiency is projected to be in the range of 55 to 60%, (fusion energy to electric power), depending on parameters selected. Turbine inlet temperatures above 2000/sup 0/F, while they do increase efficiency somewhat, are not necessarily for high cycle efficiency.

Powell, J.; Fillo, J.; Makowitz, H.

1978-01-01T23:59:59.000Z

305

Synfuels from fusion: producing hydrogen with the tandem mirror reactor and thermochemical cycles  

DOE Green Energy (OSTI)

This report examines, for technical merit, the combination of a fusion reactor driver and a thermochemical plant as a means for producing synthetic fuel in the basic form of hydrogen. We studied: (1) one reactor type - the Tandem Mirror Reactor - wishing to use to advantage its simple central cell geometry and its direct electrical output; (2) two reactor blanket module types - a liquid metal cauldron design and a flowing Li/sub 2/O solid microsphere pellet design so as to compare the technology, the thermal-hydraulics, neutronics and tritium control in a high-temperature operating mode (approx. 1200 K); (3) three thermochemical cycles - processes in which water is used as a feedstock along with a high-temperature heat source to produce H/sub 2/ and O/sub 2/.

Ribe, F.L.; Werner, R.W.

1981-01-21T23:59:59.000Z

306

Experience with non-fuel-bearing components in LWR (light-water reactor) fuel systems  

SciTech Connect

Many non-fuel-bearing components are so closely associated with the spent fuel assemblies that their integrity and behavior must be taken into consideration with the fuel assemblies, when handling spent fuel of planning waste management activities. Presented herein is some of the experience that has been gained over the past two decades from non-fuel-bearing components in light-water reactors (LWRs), both pressurized-water reactors (PWRs) and boiling-water reactors (BWRs). Among the most important of these components are the control rod systems, the absorber and burnable poison rods, and the fuel assembly channels. 15 refs., 5 figs., 2 tabs.

Bailey, W.J.; Berting, F.M.

1990-12-01T23:59:59.000Z

307

FUEL ELEMENT FOR A NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Duffy, J.G. Jr.

1961-05-30T23:59:59.000Z

308

FUEL ELEMENT FOR A NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Duffy, J.G. Jr.

1961-05-30T23:59:59.000Z

309

NFCSim: A Dynamic Fuel Burnup and Fuel Cycle Simulation Tool  

Science Conference Proceedings (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Core Physics and Fuel Management Methods, Analytical Tools, and Benchmarks

Erich A. Schneider; Charles G. Bathke; Michael R. James

310

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

SciTech Connect

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

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

2004-07-01T23:59:59.000Z

311

Environmental Emissions from Energy Technology Systems: The Total Fuel Cycle  

SciTech Connect

This is a summary report that compares emissions during the entire project life cycle for a number of fossil-fueled and renewable electric power systems, including geothermal steam (probably modeled after The Geysers). The life cycle is broken into Fuel Extraction, Construction, and Operation. The only emission covered is carbon dioxide.

San Martin, Robert L.

1989-01-01T23:59:59.000Z

312

Environmental Emissions From Energy Technology Systems: The Total Fuel Cycle  

SciTech Connect

This is a summary report that compares emissions during the entire project life cycle for a number of fossil-fueled and renewable electric power systems, including geothermal steam (probably modeled after The Geysers). The life cycle is broken into Fuel Extraction, Construction, and Operation. The only emission covered is carbon dioxide. (DJE 2005)

San Martin, Robert L.

1989-04-01T23:59:59.000Z

313

Davis-Besse Cycle 16 Fuel Deposit Analysis and Characterization  

Science Conference Proceedings (OSTI)

Fuel deposit samples were collected from Davis-Besse Unit 1 during the EOC16 outage. The impetus behind collecting crud samples came from the observation of unusual deposits on fuel during EOC15, as well as measured crud-induced power shape (CIPS) during Cycle 16. The purpose of EOC16 sample campaign therefore was to determine the nature of the fuel deposits. Samples were collected from two fuel assemblies, one after one cycle of exposure and the other after two cycles of exposure. Samples were collected...

2011-12-23T23:59:59.000Z

314

Modeling and design of a reload PWR core for a 48-month fuel cycle  

Science Conference Proceedings (OSTI)

The objective of this research was to use state-of-the-art nuclear and fuel performance packages to evaluate the feasibility and costs of a 48 calendar month core in existing pressurized water reactor (PWR) designs, considering the full range of practical design and economic considerations. The driving force behind this research is the desire to make nuclear power more economically competitive with fossil fuel options by expanding the scope for achievement of higher capacity factors. Using CASMO/SIMULATE, a core design with fuel enriched to 7{sup w}/{sub o} U{sup 235} for a single batch loaded, 48-month fuel cycle has been developed. This core achieves an ultra-long cycle length without exceeding current fuel burnup limits. The design uses two different types of burnable poisons. Gadolinium in the form of gadolinium oxide (Gd{sub 2}O{sub 3}) mixed with the UO{sub 2} of selected pins is sued to hold down initial reactivity and to control flux peaking throughout the life of the core. A zirconium di-boride (ZrB{sub 2}) integral fuel burnable absorber (IFBA) coating on the Gd{sub 2}O{sub 3}-UO{sub 2} fuel pellets is added to reduce the critical soluble boron concentration in the reactor coolant to within acceptable limits. Fuel performance issues of concern to this design are also outlined and areas which will require further research are highlighted.

McMahon, M.V.; Driscoll, M.J.; Todreas, N.E. [Massachusetts Inst. of Tech., Cambridge, MA (United States)

1997-05-01T23:59:59.000Z

315

Regulatory cross-cutting topics for fuel cycle facilities.  

Science Conference Proceedings (OSTI)

This report overviews crosscutting regulatory topics for nuclear fuel cycle facilities for use in the Fuel Cycle Research&Development Nuclear Fuel Cycle Evaluation and Screening study. In particular, the regulatory infrastructure and analysis capability is assessed for the following topical areas:Fire Regulations (i.e., how applicable are current Nuclear Regulatory Commission (NRC) and/or International Atomic Energy Agency (IAEA) fire regulations to advance fuel cycle facilities)Consequence Assessment (i.e., how applicable are current radionuclide transportation tools to support risk-informed regulations and Level 2 and/or 3 PRA) While not addressed in detail, the following regulatory topic is also discussed:Integrated Security, Safeguard and Safety Requirement (i.e., how applicable are current Nuclear Regulatory Commission (NRC) regulations to future fuel cycle facilities which will likely be required to balance the sometimes conflicting Material Accountability, Security, and Safety requirements.)

Denman, Matthew R.; Brown, Jason; Goldmann, Andrew Scott; Louie, David

2013-10-01T23:59:59.000Z

316

Preparation of high temperature gas-cooled reactor fuel element  

DOE Patents (OSTI)

This invention relates to a method for the preparation of high temperature gas-cooled reactor (HTGR) fuel elements wherein uncarbonized fuel rods are inserted in appropriate channels of an HTGR fuel element block and the entire block is inserted in an autoclave for in situ carbonization under high pressure. The method is particularly applicable to remote handling techniques.

Bradley, Ronnie A. (Oak Ridge, TN); Sease, John D. (Knoxville, TN)

1976-01-01T23:59:59.000Z

317

Experimental validation of the DARWIN2.3 package for fuel cycle applications  

Science Conference Proceedings (OSTI)

The DARWIN package, developed by the CEA and its French partners (AREVA and EDF) provides the required parameters for fuel cycle applications: fuel inventory, decay heat, activity, neutron, {gamma}, {alpha}, {beta} sources and spectrum, radiotoxicity. This paper presents the DARWIN2.3 experimental validation for fuel inventory and decay heat calculations on Pressurized Water Reactor (PWR). In order to validate this code system for spent fuel inventory a large program has been undertaken, based on spent fuel chemical assays. This paper deals with the experimental validation of DARWIN2.3 for the Pressurized Water Reactor (PWR) Uranium Oxide (UOX) and Mixed Oxide (MOX) fuel inventory calculation, focused on the isotopes involved in Burn-Up Credit (BUC) applications and decay heat computations. The calculation - experiment (C/E-1) discrepancies are calculated with the latest European evaluation file JEFF-3.1.1 associated with the SHEM energy mesh. An overview of the tendencies is obtained on a complete range of burn-up from 10 to 85 GWd/t (10 to 60 GWcVt for MOX fuel). The experimental validation of the DARWIN2.3 package for decay heat calculation is performed using calorimetric measurements carried out at the Swedish Interim Spent Fuel Storage Facility for Pressurized Water Reactor (PWR) assemblies, covering a large burn-up (20 to 50 GWd/t) and cooling time range (10 to 30 years). (authors)

San-Felice, L.; Eschbach, R.; Bourdot, P. [DEN, DER, CEA-Cadarache, F-13108 ST Paul-Lez-Durance (France); Tsilanizara, A.; Huynh, T. D. [DEN, DM2S, CEA-Saclay, F-91191 Gif sur Yvette (France); Ourly, H. [EDF, R and D, 1 av. General de Gaulle, F-92131 Clamart Cedex (France); Thro, J. F. [AREVA, Tour AREVA, F-92084 Paris la Defense (France)

2012-07-01T23:59:59.000Z

318

High Performance Fuel Desing for Next Generation Pressurized Water Reactors  

SciTech Connect

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

Mujid S. Kazimi; Pavel Hejzlar

2006-01-31T23:59:59.000Z

319

Low-Enriched Fuel Design Concept for the Prismatic Very High Temperature Reactor Core  

SciTech Connect

A new non-TRISO fuel and clad design concept is proposed for the prismatic, heliumcooled Very High Temperature Reactor core. The new concept could substantially reduce the current 10-20 wt% TRISO uranium enrichments down to 4-6 wt% for both initial and reload cores. The proposed fuel form would be a high-temperature, high-density uranium ceramic, for example UO2, configured into very small diameter cylindrical rods. The small diameter fuel rods significantly increase core reactivity through improved neutron moderation and fuel lumping. Although a high-temperature clad system for the concept remains to be developed, recent success in tube fabrication and preliminary irradiation testing of silicon carbide (SiC) cladding for light water reactor applications offers good potential for this application, and for future development of other carbide clad designs. A high-temperature ceramic fuel, together with a high-temperature clad material, could also lead to higher thermal safety margins during both normal and transient reactor conditions relative to TRISO fuel. The calculated neutronic results show that the lowenrichment, small diameter fuel rods and low thermal neutron absorbing clad retain the strong negative Doppler fuel temperature coefficient of reactivity that ensures inherent safe operation of the VHTR, and depletion studies demonstrate that an 18-month power cycle can be achieved with the lower enrichment fuel.

Sterbentz, James W

2007-05-01T23:59:59.000Z

320

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

SciTech Connect

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

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


321

Spent fuel utilization in a compact traveling wave reactor  

SciTech Connect

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

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

2012-06-06T23:59:59.000Z

322

Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume III. Resources and fuel cycle facilities  

SciTech Connect

Volume III explores resources and fuel cycle facilities. Chapters are devoted to: estimates of US uranium resources and supply; comparison of US uranium demands with US production capability forecasts; estimates of foreign uranium resources and supply; comparison of foreign uranium demands with foreign production capability forecasts; and world supply and demand for other resources and fuel cycle services. An appendix gives uranium, fissile material, and separative work requirements for selected reactors and fuel cycles.

1979-12-01T23:59:59.000Z

323

IAEA-TECDOC-1450 Thorium fuel cycle --Potential  

E-Print Network (OSTI)

1985 - 1989 Lingen, Germany BWR Irradiation-testing 60 MW(e) Test Fuel (Th,Pu)O2 pellets Terminated achieved a maximum burnup of 60 000 MWd/t without any fuel failure. In India, there has always beenIAEA-TECDOC-1450 Thorium fuel cycle -- Potential benefits and challenges May 2005 #12;IAEA

Laughlin, Robert B.

324

Greek research reactor performance characteristics after addition of beryllium reflector and LEU fuel  

SciTech Connect

The GRR-1 is a 5-MW pool-type, light-water-moderated and-cooled reactor fueled with MTR-type fuel elements. Recently received Be reflector blocks will soon be added to the core to add additional reactivity until fresh LEU fuel arrives. REBUS-3 xy fuel cycle analyses, using burnup dependent cross sections, were performed to assist in fuel management decisions for the water- and Be-reflected HEU nonequilibrium cores. Cross sections generated by EPRI-CELL have been benchmarked to identical VIM Monte Carlo models. The size of the Be-reflected LEU core has been reduced to 30 elements compared to 35 for the HEU water-reflected core, and an equilibrium cycle calculation has been performed.

Deen, J.R.; Snelgrove, J.L. (Argonne National Lab., IL (United States)); Papastergiou, C. (National Center for Scientific Research, Athens (Greece))

1992-01-01T23:59:59.000Z

325

Greek research reactor performance characteristics after addition of beryllium reflector and LEU fuel  

SciTech Connect

The GRR-1 is a 5-MW pool-type, light-water-moderated and-cooled reactor fueled with MTR-type fuel elements. Recently received Be reflector blocks will soon be added to the core to add additional reactivity until fresh LEU fuel arrives. REBUS-3 xy fuel cycle analyses, using burnup dependent cross sections, were performed to assist in fuel management decisions for the water- and Be-reflected HEU nonequilibrium cores. Cross sections generated by EPRI-CELL have been benchmarked to identical VIM Monte Carlo models. The size of the Be-reflected LEU core has been reduced to 30 elements compared to 35 for the HEU water-reflected core, and an equilibrium cycle calculation has been performed.

Deen, J.R.; Snelgrove, J.L. [Argonne National Lab., IL (United States); Papastergiou, C. [National Center for Scientific Research, Athens (Greece)

1992-12-31T23:59:59.000Z

326

Benefits and concerns of a closed nuclear fuel cycle  

Science Conference Proceedings (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

327

Configuration and performance of fuel cell-combined cycle options  

DOE Green Energy (OSTI)

The natural gas, indirect-fired, carbonate fuel-cell-bottomed, combined cycle (NG-IFCFC) and the topping natural-gas/solid-oxide fuel-cell combined cycle (NG-SOFCCC) are introduced as novel power-plant systems for the distributed power and on-site markets in the 20-200 mega-watt (MW) size range. The novel NG-IFCFC power-plant system configures the ambient pressure molten-carbonate fuel cell (MCFC) with a gas turbine, air compressor, combustor, and ceramic heat exchanger: The topping solid-oxide fuel-cell (SOFC) combined cycle is not new. The purpose of combining a gas turbine with a fuel cell was to inject pressurized air into a high-pressure fuel cell and to reduce the size, and thereby, to reduce the cost of the fuel cell. Today, the SOFC remains pressurized, but excess chemical energy is combusted and the thermal energy is utilized by the Carnot cycle heat engine to complete the system. ASPEN performance results indicate efficiencies and heat rates for the NG-IFCFC or NG-SOFCCC are better than conventional fuel cell or gas turbine steam-bottomed cycles, but with smaller and less expensive components. Fuel cell and gas turbine systems should not be viewed as competitors, but as an opportunity to expand to markets where neither gas turbines nor fuel cells alone would be commercially viable. Non-attainment areas are the most likely markets.

Rath, L.K.; Le, P.H.; Sudhoff, F.A.

1995-12-31T23:59:59.000Z

328

Thermonuclear inverse magnetic pumping power cycle for stellarator reactor  

DOE Patents (OSTI)

The plasma column in a stellarator is compressed and expanded alternatively in minor radius. First a plasma in thermal balance is compressed adiabatically. The volume of the compressed plasma is maintained until the plasma reaches a new thermal equilibrium. The plasma is then expanded to its original volume. As a result of the way a stellarator works, the plasma pressure during compression is less than the corresponding pressure during expansion. Therefore, negative work is done on the plasma over a complete cycle. This work manifests itself as a back-voltage in the toroidal field coils. Direct electrical energy is obtained from this voltage. Alternatively, after the compression step, the plasma can be expanded at constant pressure. The cycle can be made self-sustaining by operating a system of two stellarator reactors in tandem. Part of the energy derived from the expansion phase of a first stellarator reactor is used to compress the plasma in a second stellarator reactor.

Ho, Darwin D. (Pleasanton, CA); Kulsrud, Russell M. (Princeton, NJ)

1991-01-01T23:59:59.000Z

329

Hot Fuel Examination Facility's neutron radiography reactor  

SciTech Connect

Argonne National Laboratory-West is located near Idaho Falls, Idaho, and is operated by the University of Chicago for the United States Department of Energy in support of the Liquid Metal Fast Breeder Reactor Program, LMFBR. The Hot Fuel Examination Facility, HFEF, is one of several facilities located at the Argonne Site. HFEF comprises a large hot cell where both nondestructive and destructive examination of highly-irradiated reactor fuels are conducted in support of the LMFBR program. One of the nondestructive examination techniques utilized at HFEF is neutron radiography, which is provided by the NRAD reactor facility (a TRIGA type reactor) below the HFEF hot cell.

Pruett, D.P.; Richards, W.J.; Heidel, C.C.

1983-01-01T23:59:59.000Z

330

Nuclear fuel cycle facility accident analysis handbook  

Science Conference Proceedings (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

331

Fuel rod retention device for a nuclear reactor  

DOE Patents (OSTI)

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

Hylton, Charles L. (Madison Heights, VA)

1984-01-01T23:59:59.000Z

332

EXTENDING SODIUM FAST REACTOR DRIVER FUEL USE TO HIGHER TEMPERATURES  

Science Conference Proceedings (OSTI)

Calculations of potential sodium-cooled fast reactor fuel temperatures were performed to estimate the effects of increasing the outlet temperature of a given fast reactor design by increasing pin power, decreasing assembly flow, or increasing inlet temperature. Based upon experience in the U.S., both metal and mixed oxide (MOX) fuel types are discussed in terms of potential performance effects created by the increased operating temperatures. Assembly outlet temperatures of 600, 650 and 700 °C were used as goal temperatures. Fuel/cladding chemical interaction (FCCI) and fuel melting, as well as challenges to the mechanical integrity of the cladding material, were identified as the limiting phenomena. For example, starting with a recent 1000 MWth fast reactor design, raising the outlet temperature to 650 °C through pin power increase increased the MOX centerline temperature to more than 3300 °C and the metal fuel peak cladding temperature to more than 700 °C. These exceeded limitations to fuel performance; fuel melting was limiting for MOX and FCCI for metal fuel. Both could be alleviated by design ‘fixes’, such as using a barrier inside the cladding to minimize FCCI in the metal fuel, or using annular fuel in the case of MOX. Both would also require an advanced cladding material with improved stress rupture properties. While some of these are costly, the benefits of having a high-temperature reactor which can support hydrogen production, or other missions requiring high process heat may make the extra costs justified.

Douglas L. Porter

2011-02-01T23:59:59.000Z

333

The DOE Advanced Gas Reactor Fuel Development and Qualification Program  

Science Conference Proceedings (OSTI)

The high outlet temperatures and high thermal-energy conversion efficiency of modular High Temperature Gas-cooled Reactors (HTGRs) enable an efficient and cost effective integration of the reactor system with non-electricity generation applications, such as process heat and/or hydrogen production, for the many petrochemical and other industrial processes that require temperatures between 300°C and 900°C. The Department of Energy (DOE) has selected the HTGR concept for the Next Generation Nuclear Plant (NGNP) Project as a transformative application of nuclear energy that will demonstrate emissions-free nuclear-assisted electricity, process heat, and hydrogen production, thereby reducing greenhouse-gas emissions and enhancing energy security. The objective of the DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification program is to qualify tristructural isotropic (TRISO)-coated particle fuel for use in HTGRs. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission-product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete, fundamental understanding of the relationship between the fuel fabrication process and key fuel properties, the irradiation and accident safety performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. An overview of the program and recent progress is presented.

David Petti

2010-09-01T23:59:59.000Z

334

Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights October 2010  

Science Conference Proceedings (OSTI)

The DB Program monthly highlights report for September 2010, ORNL/TM-2010/252, was distributed to program participants by email on October 26. This report discusses: (1) Core and Fuel Analysis; (2) Spent Fuel Management; (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor); (4) TRU (transuranic elements) HTR Fuel Qualification; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle.

Snead, Lance Lewis [ORNL; Besmann, Theodore M [ORNL; Collins, Emory D [ORNL; Bell, Gary L [ORNL

2010-11-01T23:59:59.000Z

335

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

336

Space reactor fuel element testing in upgraded TREAT  

DOE Green Energy (OSTI)

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.

Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

1993-05-01T23:59:59.000Z

337

Space reactor fuel element testing in upgraded TREAT  

DOE Green Energy (OSTI)

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.

Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

1993-01-14T23:59:59.000Z

338

The damage function approach for estimating fuel cycle externalities  

DOE Green Energy (OSTI)

This paper discusses the methodology used in a study of fuel cycle externalities sponsored by the US Department of Energy and the Commission of the European Communities. The methodology is the damage function approach. This paper describes that approach and discusses its application and limitations. The fuel cycles addressed are those in which coal, biomass, oil, hydro, natural gas and uranium are used to generate electric power. The methodology is used to estimate the physical impacts of these fuel cycles on environmental resources and human health, and the external costs and benefits of these impacts.

Lee, R.

1993-10-01T23:59:59.000Z

339

Reactor physics assessment of thick silicon carbide clad PWR fuels  

E-Print Network (OSTI)

High temperature tolerance, chemical stability and low neutron affinity make silicon carbide (SiC) a potential fuel cladding material that may improve the economics and safety of light water reactors (LWRs). "Thick" SiC ...

Bloore, David A. (David Allan)

2013-01-01T23:59:59.000Z

340

Innovative fuel designs for high power density pressurized water reactor  

E-Print Network (OSTI)

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

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

2006-01-01T23:59:59.000Z

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


341

Advanced Nuclear Fuel Concepts for Minor Actinide Burning  

Science Conference Proceedings (OSTI)

Abstract Scope, New fuel cycle strategies entail advanced nuclear fuel concepts. This especially applies for the burning of minor actinides in a fast reactor cycle ...

342

The Framatome ANP Indirect-Cycle Very High Temperature Reactor  

SciTech Connect

Framatome ANP is developing a Very High Temperature Reactor (VHTR) design, relying on its previous experience with high temperature reactor designs, from its participation in the MODUL and the GT-MHR designs. The Framatome ANP VHTR design is based on an indirect cycle coupled to an 'off-the-shelf' combined cycle gas turbine. Although direct cycle HTR's are being promoted for their high efficiency, preliminary evaluations show that the Framatome ANP design efficiency is on par with a direct cycle while avoiding PGS (Power Generation System) developments and keeping the PGS contamination free. This concept was independently evaluated with sensitivity analysis by EDF. Moreover, the nuclear heat source of the indirect cycle could also be used to qualify the direct cycle components without risk of contamination behind the IHX, thus assisting in the preparation for the later introduction of that technology. Relying to the maximum extent on available technology, the Framatome ANP VHTR plant can demonstrate high-efficiency electricity generation and carbon-free hydrogen production. (authors)

Copsey, Bernie [Framatome ANP, Inc., 3315 Old Forest Road Lynchburg, VA (United States); Lecomte, Michel [Framatome ANP, SAS, Tour AREVA Paris, La Defense (France); Brinkmann, Gerd [Framatome ANP, GmbH, 49 (9131) 18-96630, Erlangen (Germany); Capitaine, Alain; Deberne, Nicolas [EDF/SEPTEN, Villeurbanne (France)

2004-07-01T23:59:59.000Z

343

Design and fuel management of PWR cores to optimize the once-through fuel cycle  

E-Print Network (OSTI)

The once-through fuel cycle has been analyzed to see if there are substantial prospects for improved uranium ore utilization in current

Fujita, Edward Kei

344

Supercritical CO2Brayton Cycle Control Strategy for Autonomous Liquid Metal-Cooled Reactors  

Science Conference Proceedings (OSTI)

This presentation discusses a supercritical carbon dioxide brayton cycle control strategy for autonomous liquid metal-cooled reactors.

Moisseytsev, A.; Sienicki, J.J.

2004-10-06T23:59:59.000Z

345

Fuel Cycle Research and Development Presentation Title  

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

- Irradiation studies - Fuel-clad interactions - Elastic property measurement - Thermal properties - Failure model analysis - Quench testing Technology development -...

346

Fuel Cycle Technologies | Department of Energy  

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

plant underscored the urgency behind enhancing accident tolerance of the existing reactor fleet. The United States must address these challenges in order to meet our goals for...

347

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 to change the long-standing U.S. policy of preventing the spread of nuclear weapons by denying nuclear technology transfer to non-NPT signatory states. The nuclear tests in 1998 have convinced the world community that India would never relinquish its nuclear arsenal. This has driven the desire to engage India through civilian nuclear cooperation. The cornerstone of any civilian nuclear technological support necessitates the separation of military and civilian facilities. A complete nuclear fuel cycle assessment of India emphasizes the entwinment of the military and civilian facilities and would aid in moving forward with the separation plan. To estimate the existing uranium reserves in India, a complete historical assessment of ore production, conversion, and processing capabilities was performed using open source information and compared to independent reports. Nuclear energy and plutonium production (reactor- and weapons-grade) was simulated using declared capacity factors and modern simulation tools. The three-stage nuclear power program entities and all the components of civilian and military significance were assembled into a flowsheet to allow for a macroscopic vision of the Indian fuel cycle. A detailed view of the nuclear fuel cycle opens avenues for technological collaboration. The fuel cycle that grows from this study exploits domestic thorium reserves with advanced international technology and optimized for the existing system. To utilize any appreciable fraction of the world’s supply of thorium, nuclear breeding is necessary. The two known possibilities for production of more fissionable material in the reactor than is consumed as fuel are fast breeders or thermal breeders. This dissertation analyzes a thermal breeder core concept involving the CANDU core design. The end-oflife fuel characteristics evolved from the designed fuel composition is proliferation resistant and economical in integrating this technology into the Indian nuclear fuel cycle. Furthermore, it is shown that the separation of the military and civilian components of the Indian fuel cycle can be facilitated through the implementation of such a system.

Woddi, Taraknath Venkat Krishna

2007-12-01T23:59:59.000Z

348

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 to change the long-standing U.S. policy of preventing the spread of nuclear weapons by denying nuclear technology transfer to non-NPT signatory states. The nuclear tests in 1998 have convinced the world community that India would never relinquish its nuclear arsenal. This has driven the desire to engage India through civilian nuclear cooperation. The cornerstone of any civilian nuclear technological support necessitates the separation of military and civilian facilities. A complete nuclear fuel cycle assessment of India emphasizes the entwinment of the military and civilian facilities and would aid in moving forward with the separation plan. To estimate the existing uranium reserves in India, a complete historical assessment of ore production, conversion, and processing capabilities was performed using open source information and compared to independent reports. Nuclear energy and plutonium production (reactor- and weapons-grade) was simulated using declared capacity factors and modern simulation tools. The three-stage nuclear power program entities and all the components of civilian and military significance were assembled into a flowsheet to allow for a macroscopic vision of the Indian fuel cycle. A detailed view of the nuclear fuel cycle opens avenues for technological collaboration. The fuel cycle that grows from this study exploits domestic thorium reserves with advanced international technology and optimized for the existing system. To utilize any appreciable fraction of the world's supply of thorium, nuclear breeding is necessary. The two known possibilities for production of more fissionable material in the reactor than is consumed as fuel are fast breeders or thermal breeders. This dissertation analyzes a thermal breeder core concept involving the CANDU core design. The end-oflife fuel characteristics evolved from the designed fuel composition is proliferation resistant and economical in integrating this technology into the Indian nuclear fuel cycle. Furthermore, it is shown that the separation of the military and civilian components of the Indian fuel cycle can be facilitated through the implementation of such a system.

Woddi, Taraknath Venkat Krishna

2007-12-01T23:59:59.000Z

349

Yttrium and rare earth stabilized fast reactor metal fuel  

DOE Patents (OSTI)

To increase the operating temperature of a reactor, the melting point and mechanical properties of the fuel must be increased. For an actinide-rich fuel, yttrium, lanthanum and/or rare earth elements can be added, as stabilizers, to uranium and plutonium and/or a mixture of other actinides to raise the melting point of the fuel and improve its mechanical properties. Since only about 1% of the actinide fuel may be yttrium, lanthanum, or a rare earth element, the neutron penalty is low, the reactor core size can be reduced, the fuel can be burned efficiently, reprocessing requirements are reduced, and the nuclear waste disposal volumes reduced. A further advantage occurs when yttrium, lanthanum, and/or other rare earth elements are exposed to radiation in a reactor, they produce only short half life radioisotopes, which reduce nuclear waste disposal problems through much shorter assured-isolation requirements.

Guon, Jerold (Woodland Hills, CA); Grantham, LeRoy F. (Calabasas, CA); Specht, Eugene R. (Simi Valley, CA)

1992-01-01T23:59:59.000Z

350

Spent nuclear fuel discharges from US reactors 1993  

SciTech Connect

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

Not Available

1995-02-01T23:59:59.000Z

351

VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS  

DOE Patents (OSTI)

A hollow, porous-walled fuel element filled with fissionable fuel and provided with an outlet port through its wall is described. In operation in a gas-cooled reactor, the element is connected, through its outlet port, to the vacuum side of a pump that causes a portion of the coolant gas flowing over the exterior surface of the element to be drawn through the porous walls thereof and out through the outlet port. This continuous purging gas flow sweeps away gaseous fission products as they are released by the fissioning fuel. (AEC) A fuel element for a nuclear reactor incorporating a body of metal of melting point lower than the temperature of operation of the reactor and a nuclear fuel in finely divided form dispersed in the body of metal as a settled slurry is presented. (AEC)

Furgerson, W.T.

1963-12-17T23:59:59.000Z

352

High efficiency carbonate fuel cell/turbine hybrid power cycle  

Science Conference Proceedings (OSTI)

The hybrid power cycle studies were conducted to identify a high efficiency, economically competitive system. A hybrid power cycle which generates power at an LHV efficiency > 70% was identified that includes an atmospheric pressure direct carbonate fuel cell, a gas turbine, and a steam cycle. In this cycle, natural gas fuel is mixed with recycled fuel cell anode exhaust, providing water for reforming fuel. The mixed gas then flows to a direct carbonate fuel cell which generates about 70% of the power. The portion of the anode exhaust which is not recycled is burned and heat transferred through a heat exchanger (HX) to the compressed air from a gas turbine. The heated compressed air is then heated further in the gas turbine burner and expands through the turbine generating 15% of the power. Half the exhaust from the turbine provides air for the anode exhaust burner. All of the turbine exhaust eventually flows through the fuel cell cathodes providing the O2 and CO2 needed in the electrochemical reaction. Exhaust from the cathodes flows to a steam system (heat recovery steam generator, staged steam turbine generating 15% of the cycle power). Simulation of a 200 MW plant with a hybrid power cycle had an LHV efficiency of 72.6%. Power output and efficiency are insensitive to ambient temperature, compared to a gas turbine combined cycle; NOx emissions are 75% lower. Estimated cost of electricity for 200 MW is 46 mills/kWh, which is competitive with combined cycle where fuel cost is > $5.8/MMBTU. Key requirement is HX; in the 200 MW plant studies, a HX operating at 1094 C using high temperature HX technology currently under development by METC for coal gassifiers was assumed. A study of a near term (20 MW) high efficiency direct carbonate fuel cell/turbine hybrid power cycle has also been completed.

Steinfeld, G.; Maru, H.C. [Energy Research Corp., Danbury, CT (United States); Sanderson, R.A. [Sanderson (Robert) and Associates, Wethersfield, CT (United States)

1996-07-01T23:59:59.000Z

353

HTGR Technology Family Assessment for a Range of Fuel Cycle Missions  

SciTech Connect

This report examines how the HTGR technology family can provide options for the once through, modified open cycle (MOC), or full recycle fuel cycle strategies. The HTGR can serve all the fuel cycle missions that an LWR can; both are thermal reactors. Additional analyses are warranted to determine if HTGR “full recycle” service could provide improved consumption of transuranic (TRU) material than LWRs (as expected), to analyze the unique proliferation resistance issues associated with the “pebble bed” approach, and to further test and analyze methods to separate TRISO-coated fuel particles from graphite and/or to separate used HTGR fuel meat from its TRISO coating. The feasibility of these two separation issues is not in doubt, but further R&D could clarify and reduce the cost and enable options not adequately explored at present. The analyses here and the now-demonstrated higher fuel burnup tests (after the illustrative designs studied here) should enable future MOC and full recycle HTGR concepts to more rapidly consume TRU, thereby offering waste management advantages. Interest in “limited separation” or “minimum fuel treatment” separation approaches motivates study of impurity-tolerant fuel fabrication.

Steven J. Piet; Samuel E. Bays; Nick R. Soelberg

2010-11-01T23:59:59.000Z

354

Implications of high efficiency power cycles for fusion reactor design  

SciTech Connect

The implications of the High Efficiency Power Cycle for fusion reactors are examined. The proposed cycle converts most all of the high grade CTR heat input to electricity. A low grade thermal input (T approximately 100$sup 0$C) is also required, and this can be supplied at low cost geothermal energy at many locations in the U. S. Approximately 3 KW of low grade heat is required per KW of electrical output. The thermodynamics and process features of the proposed cycle are discussed. Its advantages for CTR's are that low Q machines (e.g. driven Tokamaks, mirrors) can operate with a high (approximately 80 percent) conversion of CTR fusion energy to electricity, where with conventional power cycles no plant output could be achieved with such low Q operation. (auth)

Powell, J.R.; Usher, J.; Salzano, F.J.

1975-01-01T23:59:59.000Z

355

BWR Fuel Deposit Sample Evaluation: River Bend Cycle 11 Crud Flakes (Part 1)  

Science Conference Proceedings (OSTI)

The River Bend boiling water reactor (BWR) experienced fuel defects due to heavy crud deposition during Cycle 11. This report describes the use of a new analytical methodology to examine crud samples from failed rods from this plant. The methodology uses a special scraping tool to obtain clearly defined flake samples that can then be examined by traditional analytical techniques. This new analytical methodology can provide preliminary data for root cause assessment in a matter of months rather than the y...

2004-09-24T23:59:59.000Z

356

Assessment of innovative fuel designs for high performance light water reactors  

E-Print Network (OSTI)

To increase the power density and maximum allowable fuel burnup in light water reactors, new fuel rod designs are investigated. Such fuel is desirable for improving the economic performance light water reactors loaded with ...

Carpenter, David Michael

2006-01-01T23:59:59.000Z

357

Gel-sphere-pac reactor fuel fabrication and its application to a variety of fuels  

SciTech Connect

The gel-sphere-pac fuel fabrication option was evaluated for its possible application to commercial scale fuel fabrication for 19 fuel element designs that use oxide fuel in metal clad rods. The dry gel spheres are prepared at the reprocessing plant and are then calcined, sintered, inspected, and loaded into fuel rods and packed by low-energy vibration. A fuel smear density of 83 to 88% theoretical can be obtained. All fuel fabrication process steps were defined and evaluated from fuel receiving to finished fuel element shipping. The evaluation also covers the feasibility of the process, the current status of technology, estimates of the required time and cost to develop the technology to commercial status, and the safety and licensability of commercial scale plants. The primary evaluation was for a Light-Water Reactor fuel element containing (U,Pu)O/sub 2/ fuel. The other 18 fuel element types - 3 for Light-Water Reactors, 1 for a Heavy-Water Reactor, 1 for a Gas-Cooled Fast Reactor, 7 for Liquid-Metal-Cooled Fast Breeder Reactors, and 3 pairs for Light-Water Prebreeder and Breeder Reactors - were compared with the Light-Water Reactor. The gel-sphere-pac option was found applicable to 17 of the 19 element types; the characteristics of a commercial scale plant were defined for these for making cost estimates for such plants. The evaluation clearly shows the gel-sphere-pac process to be a viable fuel fabrication option. Estimates indicate a significant potential fabrication cost advantage for the gel-sphere-pac process if a remotely operated and remotely maintained fuel fabrication plant is required.

Olsen, A.R.; Judkins, R.R. (comps.)

1979-12-01T23:59:59.000Z

358

Pilot Application to Nuclear Fuel Cycle Options | Department of Energy  

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

Pilot Application to Nuclear Fuel Cycle Options Pilot Application to Nuclear Fuel Cycle Options Pilot Application to Nuclear Fuel Cycle Options A Screening Method for Guiding R&D Decisions: Pilot Application to Screen Nuclear Fuel Cycle Options The Department of Energy's Office of Nuclear Energy (DOE-NE) invests in research and development (R&D) to ensure that the United States will maintain its domestic nuclear energy capability and scientific and technical leadership in the international community of nuclear power nations in the years ahead. The 2010 Nuclear Energy Research and Development Roadmap presents a high-level vision and framework for R&D activities that are needed to keep the nuclear energy option viable in the near term and to expand its use in the decades ahead. The roadmap identifies the development

359

Nuclear Fuel Cycle and Waste Management Technologies - Nuclear Engineering  

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

Nuclear Fuel Cycle and Nuclear Fuel Cycle and Waste Management Technologies Nuclear Fuel Cycle and Waste Management Technologies Overview Modeling and analysis Unit Process Modeling Mass Tracking System Software Waste Form Performance Modeling Safety Analysis, Hazard and Risk Evaluations Development, Design, Operation Overview Systems and Components Development Expertise System Engineering Design Other Major Programs Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE Division on Flickr Nuclear Fuel Cycle and Waste Management Technologies Overview Bookmark and Share Much of the NE Division's research is directed toward developing software and performing analyses, system engineering design, and experiments to support the demonstration and optimization of the electrometallurgical

360

Summary and recommendations: Total fuel cycle assessment workshop  

SciTech Connect

This report summarizes the activities of the Total Fuel Cycle Assessment Workshop held in Austin, Texas, during October 6--7, 1994. It also contains the proceedings from that workshop.

NONE

1995-08-01T23:59:59.000Z

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


361

Nuclear fuel cycle risk assessment: survey and computer compilation of risk-related literature. [Once-through Cycle and Plutonium Recycle  

Science Conference Proceedings (OSTI)

The US Nuclear Regulatory Commission has initiated the Fuel Cycle Risk Assessment Program to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. Both the once-through cycle and plutonium recycle are being considered. A previous report generated by this program defines and describes fuel cycle facilities, or elements, considered in the program. This report, the second from the program, describes the survey and computer compilation of fuel cycle risk-related literature. Sources of available information on the design, safety, and risk associated with the defined set of fuel cycle elements were searched and documents obtained were catalogued and characterized with respect to fuel cycle elements and specific risk/safety information. Both US and foreign surveys were conducted. Battelle's computer-based BASIS information management system was used to facilitate the establishment of the literature compilation. A complete listing of the literature compilation and several useful indexes are included. Future updates of the literature compilation will be published periodically. 760 annotated citations are included.

Yates, K.R.; Schreiber, A.M.; Rudolph, A.W.

1982-10-01T23:59:59.000Z

362

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.

363

Pressurized Water Reactor Fuel Cleaning Using Advanced Ultrasonics  

Science Conference Proceedings (OSTI)

EPRI Ultrasonic Fuel Cleaning Technology (patent pending) was successfully qualified and demonstrated in the field at AmerenUE Callaway Plant under joint sponsorship of the EPRI Robust Fuel Program, Working Group 1 Fuel/Water Chemistry, and an AmerenUE Tailored Collaboration. In October 1999, the project team cleaned sixteen reload assemblies, which are currently undergoing re-irradiation in Cycle 11 at Callaway Plant. The assemblies show no evidence to date of any adverse fuel performance as a consequen...

2000-11-17T23:59:59.000Z

364

NEUTRON REACTOR FUEL ELEMENT UTILIZING ZIRCONIUM-BASE ALLOYS  

DOE Patents (OSTI)

This patent relates to clad fuel elements for use in neutronic reactors and is drawn to such a fuel element which consists of a core of fissionable material, comprised of an alloy of zirconium and U/sup 235/ enriched uranium, encased in a jacket of a binary zirconium-tin alloy in which the tin content ranges between 1 and 15% by weight.

Saller, H.A.; Keeler, J.R.; Szumachowski, E.R.

1957-11-12T23:59:59.000Z

365

Mixed oxide fuels testing in the advanced test reactor to support plutonium disposition  

Science Conference Proceedings (OSTI)

An intense worldwide effort is now under way to find means of reducing the stockpile of weapons-grade plutonium. One of the most attractive solutions would be to use WGPu as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PUO{sub 2}) mixed with urania (UO{sub 2}). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification, (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania, (3) The effects of WGPu isotopic composition, (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight, (5) The effects of americium and gallium in WGPu, (6) Fission gas release from MOX fuel pellets made from WGPu, (7) Fuel/cladding gap closure, (8) The effects of power cycling and off-normal events on fuel integrity, (9) Development of radial distributions of burnup and fission products, (10) Power spiking near the interfaces of MOX and urania fuel assemblies, and (11) Fuel performance code validation. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified.

Ryskamp, J.M.; Sterbentz, J.W.; Chang, G.S. [and others

1995-09-01T23:59:59.000Z

366

Lessons Learned From Dynamic Simulations of Advanced Fuel Cycles  

SciTech Connect

Years of performing dynamic simulations of advanced nuclear fuel cycle options provide insights into how they could work and how one might transition from the current once-through fuel cycle. This paper summarizes those insights from the context of the 2005 objectives and goals of the Advanced Fuel Cycle Initiative (AFCI). Our intent is not to compare options, assess options versus those objectives and goals, nor recommend changes to those objectives and goals. Rather, we organize what we have learned from dynamic simulations in the context of the AFCI objectives for waste management, proliferation resistance, uranium utilization, and economics. Thus, we do not merely describe “lessons learned” from dynamic simulations but attempt to answer the “so what” question by using this context. The analyses have been performed using the Verifiable Fuel Cycle Simulation of Nuclear Fuel Cycle Dynamics (VISION). We observe that the 2005 objectives and goals do not address many of the inherently dynamic discriminators among advanced fuel cycle options and transitions thereof.

Steven J. Piet; Brent W. Dixon; Jacob J. Jacobson; Gretchen E. Matthern; David E. Shropshire

2009-04-01T23:59:59.000Z

367

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

368

Fuel Cycle Comparison for Distributed Power Technologies  

Fuel Cell Technologies Publication and Product Library (EERE)

This report examines backup power and prime power systems and addresses the potential energy and environmental effects of substituting fuel cells for existing combustion technologies based on microtur

369

Report to the American Physical Society by the study group on nuclear fuel cycles and waste management  

SciTech Connect

Utilization of nuclear fuels and management of nuclear wastes have become major topics of public discussion. Under the auspices of the American Physical Society this study was undertaken as an independent evaluation of technical issues in the use of fissionable materials in nuclear fuel cycles, together with their principal economic, environmental, health and safety implications. Reprocessing and recycling in light water reactors were examined, along with technical measures proposed as possible safeguards; advanced reactor fuel cycles were also studied for their resource and safeguards implications. Much of the work of the group centered on the principal alternatives for disposal of radioactive wastes and control of effluents. The group examined the research and development programs sponsored by government agencies along with associated relationships among agencies and between government and private industry. Available information was also considered on nuclear fuel resources, and on important economic and environmental aspects of the various fuel cycles in order to strive for a balanced comparative study. The report presents many conclusions on various aspects of the nuclear fuel cycles and also provides recommendations concerning present utilization and future improvement of fuel cycle technology.

APS Study Group Participants; Hebel, L.C. Chairman; Christensen, E.L.; Donath, F.A.; Falconer, W.E.; Lidofsky, L.J.; Moniz, E.J.; Moss, T.H.; Pigford, R.L.; Pigford, T.H.; Rochlin, G.I.; Silsbee, R.H.; Wrenn, M.E.

1978-01-01T23:59:59.000Z

370

Deep Burn Fuel Cycle Integration: Evaluation of Two-Tier Scenarios  

Science Conference Proceedings (OSTI)

The use of a deep burn strategy using VHTRs (or DB-MHR), as a means of burning transuranics produced by LWRs, was compared to performing this task with LWR MOX. The spent DB-MHR fuel was recycled for ultimate final recycle in fast reactors (ARRs). This report summarizes the preliminary findings of the support ratio (in terms of MWth installed) between LWRs, DB-MHRs and ARRs in an equilibrium “two-tier” fuel cycle scenario. Values from literature were used to represent the LWR and DB-MHR isotopic compositions. A reactor physics simulation of the ARR was analyzed to determine the effect that the DB-MHR spent fuel cooling time on the ARR transuranic consumption rate. These results suggest that the cooling time has some but not a significant impact on the ARRs conversion ratio and transuranic consumption rate. This is attributed to fissile worth being derived from non-fissile or “threshold-fissioning” isotopes in the ARR’s fast spectrum. The fraction of installed thermal capacity of each reactor in the DB-MHR 2-tier fuel cycle was compared with that of an equivalent MOX 2-tier fuel cycle, assuming fuel supply and demand are in equilibrium. The use of DB-MHRs in the 1st-tier allows for a 10% increase in the fraction of fleet installed capacity of UO2-fueled LWRs compared to using a MOX 1st-tier. Also, it was found that because the DB-MHR derives more power per unit mass of transuranics charged to the fresh fuel, the “front-end” reprocessing demand is less than MOX. Therefore, more fleet installed capacity of DB-MHR would be required to support a given fleet of UO2 LWRs than would be required of MOX plants. However, the transuranic deep burn achieved by DB-MHRs reduces the number of fast reactors in the 2nd-tier to support the DB-MHRs “back-end” transuranic output than if MOX plants were used. Further analysis of the relative costs of these various types of reactors is required before a comparative study of these options could be considered complete.

S. Bays; H. Zhang; M. Pope

2009-05-01T23:59:59.000Z

371

Nuclear power generation and fuel cycle report 1996  

SciTech Connect

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

372

Fuel-Cycle Assessment of Selected Bioethanol Production Pathways  

E-Print Network (OSTI)

Fuel-Cycle Assessment of Selected Bioethanol Production Pathways in the United States ANL/ESD/06-Cycle Assessment of Selected Bioethanol Production Pathways in the United States ANL/ESD/06-7 by M. Wu, M. Wang ................................................................................ 6 2 Simplified Process Flow Diagram of Biochemical Conversion of Corn Stover to Ethanol with Steam

Argonne National Laboratory

373

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

SciTech Connect

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

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

2011-07-18T23:59:59.000Z

374

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

SciTech Connect

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

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

2010-09-01T23:59:59.000Z

375

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

SciTech Connect

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

376

Core design study of a supercritical light water reactor with double row fuel rods  

SciTech Connect

An equilibrium core for supercritical light water reactor has been designed. A novel type of fuel assembly with dual rows of fuel rods between water rods is chosen and optimized to get more uniform assembly power distributions. Stainless steel is used for fuel rod cladding and structural material. Honeycomb structure filled with thermal isolation is introduced to reduce the usage of stainless steel and to keep moderator temperature below the pseudo critical temperature. Water flow scheme with ascending coolant flow in inner regions is carried out to achieve high outlet temperature. In order to enhance coolant outlet temperature, the radial power distributions needs to be as flat as possible through operation cycle. Fuel loading pattern and control rod pattern are optimized to flatten power distribution at inner regions. Axial fuel enrichment is divided into three parts to control axial power peak, which affects maximum cladding surface temperature. (authors)

Zhao, C.; Wu, H.; Cao, L.; Zheng, Y. [School of Nuclear Science and Technology, Xi'an Jiaotong Univ., No. 28, Xianning West Road, Xi'an, ShannXi, 710049 (China); Yang, J.; Zhang, Y. [China Nuclear Power Technology Research Inst., Yitian Road, ShenZhen, GuangDong, 518026 (China)

2012-07-01T23:59:59.000Z

377

Fusion fuel cycle: material requirements and potential effluents  

SciTech Connect

Environmental effluents that may be associated with the fusion fuel cycle are identified. Existing standards for controlling their release are summarized and anticipated regulatory changes are identified. The ability of existing and planned environmental control technology to limit effluent releases to acceptable levels is evaluated. Reference tokamak fusion system concepts are described and the principal materials required of the associated fuel cycle are analyzed. These materials include the fusion fuels deuterium and tritium; helium, which is used as a coolant for both the blanket and superconducting magnets; lithium and beryllium used in the blanket; and niobium used in the magnets. The chemical and physical processes used to prepare these materials are also described.

Teofilo, V.L.; Bickford, W.E.; Long, L.W.; Price, B.A.; Mellinger, P.J.; Willingham, C.E.; Young, J.K.

1980-10-01T23:59:59.000Z

378

World nuclear capacity and fuel cycle requirements, November 1993  

SciTech Connect

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

379

Introduction to Nuclear Fuel Cycle and Advanced Nuclear Fuels: Jon ...  

Science Conference Proceedings (OSTI)

Mar 1, 2012 ... Increased use of fossil fuel will result in. • Resource shortfalls and regional conflicts,. • Serious environmental impact. • Worldwide expansion of ...

380

Nuclear Power Generation and Fuel Cycle Report 1997  

Gasoline and Diesel Fuel Update (EIA)

7) 7) Distribution Category UC-950 Nuclear Power Generation and Fuel Cycle Report 1997 September 1997 Energy Information Administration Office of Coal, Nuclear, Electric and Alternate Fuels U.S. Department of Energy Washington, DC 20585 This report was prepared by the Energy Information Administration, the independent statistical and analytical agency within the Department of Energy. The information contained herein should not be construed as advocating or reflecting any policy position of the Department of Energy or of any other organization. Contacts Energy Information Administration/ Nuclear Power Generation and Fuel Cycle Report 1997 ii The Nuclear Power Generation and Fuel Cycle Report is prepared by the U.S. Department of Energy's Energy Information Administration. Questions and comments concerning the contents of the report may be directed to:

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


381

2012 Fuel Cycle Technologies Annual Review Meeting Transaction Report |  

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

Fuel Cycle Technologies Annual Review Meeting Transaction Fuel Cycle Technologies Annual Review Meeting Transaction Report 2012 Fuel Cycle Technologies Annual Review Meeting Transaction Report The United States must continue to ensure improvements and access to this technology so we can meet our economic, environmental and energy security goals. We rely on nuclear energy because it provides a consistent, reliable and stable source of base load electricity with an excellent safety record in the United States. In order to continue or expand the role for nuclear power in our long- term energy platform, the United States must: Continually improve the safety and security of nuclear energy and its associated technologies worldwide. Develop solutions for the transportation, storage, and long-term disposal of used nuclear fuel and associated wastes.

382

Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies  

SciTech Connect

This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO{sub 2} assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the {sup 239}Pu and {ge}90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

Chodak, P. III

1996-05-01T23:59:59.000Z

383

Novel Reactor Design for Solid Fuel Chemical Looping Combustion  

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

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

384

The Stirred Tank Reactor Polymer Electrolyte Membrane Fuel Cell  

E-Print Network (OSTI)

The design and operation of a differential Polymer Electrolyte Membrane (PEM) fuel cell is described. The fuel cell design is based on coupled Stirred Tank Reactors (STR); the gas phase in each reactor compartment was well mixed. The characteristic times for reactant flow, gas phase diffusion and reaction were chosen so that the gas compositions at both the anode and cathode are uniform. The STR PEM fuel cell is one-dimensional; the only spatial gradients are transverse to the membrane. The STR PEM fuel cell was employed to examine fuel cell start- up, and its dynamic responses to changes in load, temperature and reactant flow rates. Multiple time scales in systems response are found to correspond to water absorption by the membrane, water transport through the membrane and stress-related mechanical changes of the membrane.

Benziger, J; Karnas, E; Moxley, J; Teuscher, C; Kevrekidis, Yu G; Benziger, Jay

2003-01-01T23:59:59.000Z

385

Fuel Cycle Technologies | Department of Energy  

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

in the fossil fuel supply. As the only large-scale source of nearly greenhouse gas-free energy, nuclear power is an essential part of our energy mix, generating about 20...

386

Fuel cycle comparison of distributed power generation technologies.  

DOE Green Energy (OSTI)

The fuel-cycle energy use and greenhouse gas (GHG) emissions associated with the application of fuel cells to distributed power generation were evaluated and compared with the combustion technologies of microturbines and internal combustion engines, as well as the various technologies associated with grid-electricity generation in the United States and California. The results were primarily impacted by the net electrical efficiency of the power generation technologies and the type of employed fuels. The energy use and GHG emissions associated with the electric power generation represented the majority of the total energy use of the fuel cycle and emissions for all generation pathways. Fuel cell technologies exhibited lower GHG emissions than those associated with the U.S. grid electricity and other combustion technologies. The higher-efficiency fuel cells, such as the solid oxide fuel cell (SOFC) and molten carbonate fuel cell (MCFC), exhibited lower energy requirements than those for combustion generators. The dependence of all natural-gas-based technologies on petroleum oil was lower than that of internal combustion engines using petroleum fuels. Most fuel cell technologies approaching or exceeding the DOE target efficiency of 40% offered significant reduction in energy use and GHG emissions.

Elgowainy, A.; Wang, M. Q.; Energy Systems

2008-12-08T23:59:59.000Z

387

SAFEGUARDS EXPERIENCE ON THE DUPIC FUEL CYCLE PROCESS  

SciTech Connect

Safeguards have been applied to the R and D process for directly fabricating CANDU fuel with PWR spent fuel material. Safeguards issues to be resolved were identified in the areas such as international cooperation on handling foreign origin nuclear material, technology development of operator's measurement system of the bulk handling process of spent fuel material, and a built-in C/S system for independent verification of material flow. The fuel cycle concept (Direct Use of PWR spent fuel in CANDU, DUPIC) was developed in consideration of reutilization of over-flowing spent fuel resources at PWR sites and a reduction of generated high level wastes. All those safeguards issues have been finally resolved, and the first batch of PWR spent fuel material was successfully introduced in the DUPIC lab facility and has been in use for routine process development.

J. HONG; H. KIM; ET AL

2001-02-01T23:59:59.000Z

388

High efficiency carbonate fuel cell/turbine hybrid power cycles  

SciTech Connect

Carbonate fuel cells developed in commercial 2.85 MW size, have an efficiency of 57.9%. Studies of higher efficiency hybrid power cycles were conducted to identify an economically competitive system and an efficiency over 65%. A hybrid power cycle was identified that includes a direct carbonate fuel cell, a gas turbine, and a steam cycle, which generates power at a LHV efficiency over 70%; it is called a Tandem Technology Cycle (TTC). In a TTC operating on natural gas fuel, 95% of the fuel is mixed with recycled fuel cell anode exhaust, providing water for reforming the fuel, and flows to a direct carbonate fuel cell system which generates 72% of the power. The portion of fuel cell anode exhaust not recycled, is burned and heat is transferred to compressed air from a gas turbine, heating it to 1800 F. The stream is then heated to 2000 F in gas turbine burner and expands through the turbine generating 13% of the power. Half the gas turbine exhaust flows to anode exhaust burner and the rest flows to the fuel cell cathodes providing the O2 and CO2 needed in the electrochemical reaction. Studies of the TTC for 200 and 20 MW size plants quantified performance, emissions and cost-of-electricity, and compared the TTC to gas turbine combined cycles. A 200-MW TTC plant has an efficiency of 72.6%; estimated cost of electricity is 45.8 mills/kWhr. A 20-MW TTC plant has an efficiency of 65.2% and a cost of electricity of 50 mills/kWhr.

Steinfeld, G.

1996-12-31T23:59:59.000Z

389

SOLVENT EXTRACTION RESEARCH AND DEVELOPMENT IN THE U.S. FUEL CYCLE PROGRAM  

SciTech Connect

Treatment or processing of used nuclear fuel to recycle uranium and plutonium has historically been accomplished using the well known PUREX process. The PUREX process has been used on an industrial scale for over 60 years in the nuclear industry. Research is underway to develop advanced separation methods for the recovery of other used fuel components, such as the minor actinides (Np, Am, Cm) for possible transmutation in fast spectrum reactors, or other constituents (e.g. Cs, Sr, transition metals, lanthanides) to help facilitate effective waste management options. This paper will provide an overview of new solvent extraction processes developed for advanced nuclear fuel cycles, and summarize recent experimental results. This will include the utilization of new extractants for selective separation of target metals and new processes developed to selectively recover one or more elements from used fuel.

Terry A. Todd

2011-10-01T23:59:59.000Z

390

Thermonuclear inverse magnetic pumping power cycle for stellarator reactor  

DOE Patents (OSTI)

The plasma column in a stellarator is compressed and expanded alternatively in minor radius. First a plasma in thermal balance is compressed adiabatically. The volume of the compressed plasma is maintained until the plasma reaches a new thermal equilibrium. The plasma is then expanded to its original volume. As a result of the way a stellarator works, the plasma pressure during compression is less than the corresponding pressure during expansion. Therefore, negative work is done on the plasma over a complete cycle. This work manifests itself as a back-voltage in the toroidal field coils. Direct electrical energy is obtained from this voltage. Alternatively, after the compression step, the plasma can be expanded at constant pressure. The cycle can be made self-sustaining by operating a system of two stellarator reactors in tandem. Part of the energy derived from the expansion phase of a first stellarator reactor is used to compress the plasma in a second stellarator reactor. 9 figs., 4 tabs.

Ho, D.D.M.; Kulsrud, R.M.

1986-09-25T23:59:59.000Z

391

Reactor physics behavior of transuranic-bearing TRISO-particle fuel in a pressurized water reactor  

SciTech Connect

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) - only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is retained. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint. (authors)

Pope, M. A.; Sen, R. S.; Ougouag, A. M.; Youinou, G. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Boer, B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); SCK-CEN, Boertang 200, BE-2400 Mol (Belgium)

2012-07-01T23:59:59.000Z

392

Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor  

SciTech Connect

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag; Gilles Youinou; Brian Boer

2012-04-01T23:59:59.000Z

393

The FIT Model - Fuel-cycle Integration and Tradeoffs  

Science Conference Proceedings (OSTI)

All mass streams from fuel separation and fabrication are products that must meet some set of product criteria – fuel feedstock impurity limits, waste acceptance criteria (WAC), material storage (if any), or recycle material purity requirements such as zirconium for cladding or lanthanides for industrial use. These must be considered in a systematic and comprehensive way. The FIT model and the “system losses study” team that developed it [Shropshire2009, Piet2010] are an initial step by the FCR&D program toward a global analysis that accounts for the requirements and capabilities of each component, as well as major material flows within an integrated fuel cycle. This will help the program identify near-term R&D needs and set longer-term goals. The question originally posed to the “system losses study” was the cost of separation, fuel fabrication, waste management, etc. versus the separation efficiency. In other words, are the costs associated with marginal reductions in separations losses (or improvements in product recovery) justified by the gains in the performance of other systems? We have learned that that is the wrong question. The right question is: how does one adjust the compositions and quantities of all mass streams, given uncertain product criteria, to balance competing objectives including cost? FIT is a method to analyze different fuel cycles using common bases to determine how chemical performance changes in one part of a fuel cycle (say used fuel cooling times or separation efficiencies) affect other parts of the fuel cycle. FIT estimates impurities in fuel and waste via a rough estimate of physics and mass balance for a set of technologies. If feasibility is an issue for a set, as it is for “minimum fuel treatment” approaches such as melt refining and AIROX, it can help to make an estimate of how performances would have to change to achieve feasibility.

Steven J. Piet; Nick R. Soelberg; Samuel E. Bays; Candido Pereira; Layne F. Pincock; Eric L. Shaber; Meliisa C Teague; Gregory M Teske; Kurt G Vedros

2010-09-01T23:59:59.000Z

394

THE NUCLEAR FUEL CYCLE: PROSPECTS FOR REDUCING ITS COST  

SciTech Connect

Nuclear fuel cost of 1.25 mills/kwh would make nuclear power competitive with conventional power in lowcost coal areas if capital and operating costs can be brought to within about 10 percent of those of coal-fired plants. Substantial decreases in fuel fabrication cost are anticipated by 1970: other costs in the fuel cycle are expccted to remain about the same as at present. Unit costs and irradiation levels that would be needed to give a fuel cost of 1.25 mills/kwh are believed to be attainable by 1970. (auth)

Albrecht, W.L.

1959-02-20T23:59:59.000Z

395

Combined cycle phosphoric acid fuel cell electric power system  

DOE Green Energy (OSTI)

By arranging two or more electric power generation cycles in series, combined cycle systems are able to produce electric power more efficiently than conventional single cycle plants. The high fuel to electricity conversion efficiency results in lower plant operating costs, better environmental performance, and in some cases even lower capital costs. Despite these advantages, combined cycle systems for the 1 - 10 megawatt (MW) industrial market are rare. This paper presents a low noise, low (oxides of nitrogen) NOx, combined cycle alternative for the small industrial user. By combining a commercially available phosphoric acid fuel cell (PAFC) with a low-temperature Rankine cycle (similar to those used in geothermal applications), electric conversion efficiencies between 45 and 47 percent are predicted. While the simple cycle PAFC is competitive on a cost of energy basis with gas turbines and diesel generators in the 1 to 2 MW market, the combined cycle PAFC is competitive, on a cost of energy basis, with simple cycle diesel generators in the 4 to 25 MW market. In addition, the efficiency and low-temperature operation of the combined cycle PAFC results in a significant reduction in carbon dioxide emissions with NO{sub x} concentration on the order of 1 parts per million (per weight) (ppmw).

Mollot, D.J.; Micheli, P.L.

1995-12-31T23:59:59.000Z

396

Electrometallurgical treatment of degraded N-reactor fuel  

Science Conference Proceedings (OSTI)

N-Reactor fuel constitutes almost 80% of the entire mass of the US Department of Energy's (DOE's) spent fuel inventory. The current plan for disposition of this fuel calls for interim dry storage, followed by direct repository disposal. However, this approach may not be viable for the entire inventory of N-Reactor fuel. The physical condition and chemical composition of much of the fuel have changed during the period that it has been in storage. The cladding of many of the fuel elements has been breached, allowing the metallic uranium fuel to react with water in the storage pools producing uranium oxides (U{sub x}O{sub y}) and uranium hydride (UH{sub 3}). Even if the breached fuel is placed in dry storage, it may continue to undergo significant changes caused by the reaction of exposed uranium with any remaining water in the container. Uranium oxides, uranium hydride, and hydrogen gas are expected to form as a result of this reaction. The presence of potentially explosive hydrogen and uranium hydride, which under certain conditions is pyrophoric, raises technical concerns that will need to be addressed. The electrometallurgical treatment process developed by Argonne National Laboratory (ANL) has potential for conditioning degraded N-Reactor fuel for long-term storage or disposal. The first step in evaluating the applicability of this process is the preparation of degraded fuel that is similar to the actual degraded N-Reactor fuel. Subsequently, the simulated degraded fuel can be introduced into an electrorefiner to examine the effect of corrosion products on the electrorefining process. Some of the technical issues to be resolved include the viability of direct electrorefining without a head-end reduction step, the effect of adherent corrosion products on the electrorefining kinetics, and the recovery and treatment of loose corrosion products that pull away from the degraded fuel. This paper presents results from an experimental study of the preparation, characterization, and subsequent electrometallurgical treatment of samples of simulated degraded N-Reactor fuel.

Gourishankar, K. V.; Karell, E. J.; Everhart, R. E.; Indacochea, E.

2000-03-03T23:59:59.000Z

397

Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights September 2010  

SciTech Connect

The DB Program monthly highlights report for August 2010, ORNL/TM-2010/184, was distributed to program participants by email on September 17. This report discusses: (1) Core and Fuel Analysis - (a) Core Design Optimization in the HTR (high temperature helium-cooled reactor) Prismatic Design (Logos), (b) Core Design Optimization in the HTR Pebble Bed Design (INL), (c) Microfuel analysis for the DB HTR (INL, GA, Logos); (2) Spent Fuel Management - (a) TRISO (tri-structural isotropic) repository behavior (UNLV), (b) Repository performance of TRISO fuel (UCB); (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor) - Synergy with other reactor fuel cycles (GA, Logos); (4) TRU (transuranic elements) HTR Fuel Qualification - (a) Thermochemical Modeling, (b) Actinide and Fission Product Transport, (c) Radiation Damage and Properties; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle - (a) Graphite Recycle (ORNL), (b) Aqueous Reprocessing, (c) Pyrochemical Reprocessing METROX (metal recovery from oxide fuel) Process Development (ANL).

Snead, Lance Lewis [ORNL; Besmann, Theodore M [ORNL; Collins, Emory D [ORNL; Bell, Gary L [ORNL

2010-10-01T23:59:59.000Z

398

Fuel provision for nonbreeding deuterium-tritium fusion reactors  

SciTech Connect

Nonbreeding D-T reactors have decisive advantages in minimum size, unit cost, variety of applications, and ease of heat removal over reactors using any other fusion cycle, and significant advantages in environmental and safety characteristics over breeding D-T reactors. Considerations of relative energy production demonstrate that the most favorable source of tritium for a widely deployed system of nonbreeding D-T reactors is the very large (approx. 10 GW thermal) semi-catalyzed-deuterium (SCD), or sub-SCD reactor, where none of the escaping /sup 3/He (> 95%) or tritium (< 25%) is reinjected for burn-up. Feasibility of the ignited SCD tokamak reactor requires spatially averaged betas of 15 to 20% with a magnetic field at the TF coils of 12 to 13 Tesla.

Jassby, D.L.; Katsurai, M.

1980-01-01T23:59:59.000Z

399

System for fuel rod removal from a reactor module  

DOE Patents (OSTI)

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

Matchett, Richard L. (Bethel Park, PA); Roof, David R. (North Huntingdon, PA); Kikta, Thomas J. (Pittsburgh, PA); Wilczynski, Rosemarie (McKees Rocks, PA); Nilsen, Roy J. (Pittsburgh, PA); Bacvinskas, William S. (Bethel Park, PA); Fodor, George (Pittsburgh, PA)

1990-01-01T23:59:59.000Z

400

System for fuel rod removal from a reactor module  

DOE Patents (OSTI)

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

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

1988-07-28T23:59:59.000Z

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401

High Density Fuel Development for Research Reactors  

SciTech Connect

An international effort to develop, qualify, and license high and very high density fuels has been underway for several years within the framework of multi-national RERTR programs. The current development status is the result of significant contributions from many laboratories, specifically CNEA in Argentina, AECL in Canada, CEA in France, TUM in Germany, KAERI in Korea, VNIIM, RDIPE, IPPE, NCCP and RIARR in Russia, INL, ANL and Y-12 in USA. These programs are mainly engaged with UMo dispersion fuels with densities from 6 to 8 gU/cm3 (high density fuel) and UMo monolithic fuel with density as high as 16 gU/cm3 (very high density fuel). This paper, mainly focused on the French & US programs, gives the status of high density UMo fuel development and perspectives on their qualification.

Daniel Wachs; Dennis Keiser; Mitchell Meyer; Douglas Burkes; Curtis Clark; Glenn Moore; Jan-Fong Jue; Totju Totev; Gerard Hofman; Tom Wiencek; Yeon So Kim; Jim Snelgrove

2007-09-01T23:59:59.000Z

402

Regulatory cross-cutting topics for fuel cycle facilities.  

SciTech Connect

This report overviews crosscutting regulatory topics for nuclear fuel cycle facilities for use in the Fuel Cycle Research&Development Nuclear Fuel Cycle Evaluation and Screening study. In particular, the regulatory infrastructure and analysis capability is assessed for the following topical areas:Fire Regulations (i.e., how applicable are current Nuclear Regulatory Commission (NRC) and/or International Atomic Energy Agency (IAEA) fire regulations to advance fuel cycle facilities)Consequence Assessment (i.e., how applicable are current radionuclide transportation tools to support risk-informed regulations and Level 2 and/or 3 PRA) While not addressed in detail, the following regulatory topic is also discussed:Integrated Security, Safeguard and Safety Requirement (i.e., how applicable are current Nuclear Regulatory Commission (NRC) regulations to future fuel cycle facilities which will likely be required to balance the sometimes conflicting Material Accountability, Security, and Safety requirements.)

Denman, Matthew R.; Brown, Jason; Goldmann, Andrew Scott; Louie, David

2013-10-01T23:59:59.000Z

403

Criticality safety strategy for the Fuel Cycle Facility electrorefiner at Argonne National Laboratory, West  

Science Conference Proceedings (OSTI)

The Integral Fast Reactor being developed by Argonne National Laboratory (ANL) combines the advantages of metal-fueled, liquid-metal-cooled reactors and a closed fuel cycle. Presently, the Fuel Cycle Facility (FCF) at ANL-West in Idaho Falls, Idaho is being modified to recycle spent metallic fuel from Experimental Breeder Reactor II as part of a demonstration project sponsored by the Department of Energy. A key component of the FCF is the electrorefiner (ER) in which the actinides are separated from the fission products. In the electrorefining process, the metal fuel is anodically dissolved into a high-temperature molten salt and refined uranium or uranium/plutonium products are deposited at cathodes. In this report, the criticality safety strategy for the FCF ER is summarized. FCF ER operations and processes formed the basis for evaluating criticality safety and control during actinide metal fuel refining. In order to show criticality safety for the FCF ER, the reference operating conditions for the ER had to be defined. Normal operating envelopes (NOES) were then defined to bracket the important operating conditions. To keep the operating conditions within their NOES, process controls were identified that can be used to regulate the actinide forms and content within the ER. A series of operational checks were developed for each operation that wig verify the extent or success of an operation. The criticality analysis considered the ER operating conditions at their NOE values as the point of departure for credible and incredible failure modes. As a result of the analysis, FCF ER operations were found to be safe with respect to criticality.

Mariani, R.D.; Benedict, R.W. [Argonne National Lab., Idaho Falls, ID (United States); Lell, R.M.; Turski, R.B.; Fujita, E.K. [Argonne National Lab., IL (United States)

1993-09-01T23:59:59.000Z

404

Impact of Nuclear Energy Futures on Advanced Fuel Cycle Options  

SciTech Connect

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

405

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

DOE Patents (OSTI)

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