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Note: This page contains sample records for the topic "advanced reactor concepts" from the National Library of EnergyBeta (NLEBeta).
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1

Advance Reactor Concepts Technical Review Panel Public Report  

Broader source: Energy.gov [DOE]

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

2

ASME Material Challenges for Advanced Reactor Concepts  

SciTech Connect (OSTI)

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

Piyush Sabharwall; Ali Siahpush

2013-07-01T23:59:59.000Z

3

Advanced Reactor Concepts Technical Review Panel Report | Department of  

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

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

4

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

SciTech Connect (OSTI)

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

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

1981-09-15T23:59:59.000Z

5

A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts  

SciTech Connect (OSTI)

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

Jacques V Hugo; David I Gertman; Jeffrey C Joe

2014-08-01T23:59:59.000Z

6

Generation -IV Reactor Concepts  

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

Generation-IV Reactor Concepts Generation-IV Reactor Concepts Thomas H. Fanning Argonne National Laboratory 9700 South Cass Avenue Argonne, Illinois 60439, USA The Generation-IV International Forum (GIF) is a multi-national research and development (R&D) collaboration. The GIF pursues the development of advanced, next generation reactor technology with goals to improve: a) sustainability (effective fuel utilization and minimization of waste) b) economics (competitiveness with respect to other energy sources) c) safety and reliability (e.g., no need for offsite emergency response), and d) proliferation resistance and physical protection The GIF Technology Roadmap exercise selected six generic systems for further study: the Gas- cooled Fast Reactor (GFR), the Lead-cooled Fast Reactor (LFR), the Molten Salt Reactor (MSR),

7

Advanced Reactor Technologies | Department of Energy  

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

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

8

Neutronic Analysis of an Advanced Fuel Design Concept for the High Flux Isotope Reactor  

SciTech Connect (OSTI)

This study presents the neutronic analysis of an advanced fuel design concept for the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) that could significantly extend the current fuel cycle length under the existing design and safety criteria. A key advantage of the fuel design herein proposed is that it would not require structural changes to the present HFIR core, in other words, maintaining the same rated power and fuel geometry (i.e., fuel plate thickness and coolant channel dimensions). Of particular practical importance, as well, is the fact that the proposed change could be justified within the bounds of the existing nuclear safety basis. The simulations herein reported employed transport theory-based and exposure-dependent eigenvalue characterization to help improve the prediction of key fuel cycle parameters. These parameters were estimated by coupling a benchmarked three-dimensional MCNP5 model of the HFIR core to the depletion code ORIGEN via the MONTEBURNS interface. The design of an advanced HFIR core with an improved fuel loading is an idea that evolved from early studies by R. D. Cheverton, formerly of ORNL. This study contrasts a modified and increased core loading of 12 kg of 235U against the current core loading of 9.4 kg. The simulations performed predict a cycle length of 39 days for the proposed fuel design, which represents a 50% increase in the cycle length in response to a 25% increase in fissile loading, with an average fuel burnup increase of {approx}23%. The results suggest that the excess reactivity can be controlled with the present design and arrangement of control elements throughout the core's life. Also, the new power distribution is comparable or even improved relative to the current power distribution, displaying lower peak to average fission rate densities across the inner fuel element's centerline and bottom cells. In fact, the fission rate density in the outer fuel element also decreased at these key locations for the proposed design. Overall, it is estimated that the advanced core design could increase the availability of the HFIR facility by {approx}50% and generate {approx}33% more neutrons annually, which is expected to yield sizeable savings during the remaining life of HFIR, currently expected to operate through 2014. This study emphasizes the neutronics evaluation of a new fuel design. Although a number of other performance parameters of the proposed design check favorably against the current design, and most of the core design features remain identical to the reference, it is acknowledged that additional evaluations would be required to fully justify the thermal-hydraulic and thermal-mechanical performance of a new fuel design, including checks for cladding corrosion performance as well as for industrial and economic feasibility.

Xoubi, Ned [ORNL; Primm, Trent [ORNL; Maldonado, G. Ivan [University of Tennessee, Knoxville (UTK)

2009-01-01T23:59:59.000Z

9

Advanced Reactor Technology Documents | Department of Energy  

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

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

10

ADVANCED SULFUR CONTROL CONCEPTS  

SciTech Connect (OSTI)

Conventional sulfur removal in integrated gasification combined cycle (IGCC) power plants involves numerous steps: COS (carbonyl sulfide) hydrolysis, amine scrubbing/regeneration, Claus process, and tail-gas treatment. Advanced sulfur removal in IGCC systems involves typically the use of zinc oxide-based sorbents. The sulfides sorbent is regenerated using dilute air to produce a dilute SO{sub 2} (sulfur dioxide) tail gas. Under previous contracts the highly effective first generation Direct Sulfur Recovery Process (DSRP) for catalytic reduction of this SO{sub 2} tail gas to elemental sulfur was developed. This process is currently undergoing field-testing. In this project, advanced concepts were evaluated to reduce the number of unit operations in sulfur removal and recovery. Substantial effort was directed towards developing sorbents that could be directly regenerated to elemental sulfur in an Advanced Hot Gas Process (AHGP). Development of this process has been described in detail in Appendices A-F. RTI began the development of the Single-step Sulfur Recovery Process (SSRP) to eliminate the use of sorbents and multiple reactors in sulfur removal and recovery. This process showed promising preliminary results and thus further process development of AHGP was abandoned in favor of SSRP. The SSRP is a direct Claus process that consists of injecting SO{sub 2} directly into the quenched coal gas from a coal gasifier, and reacting the H{sub 2}S-SO{sub 2} mixture over a selective catalyst to both remove and recover sulfur in a single step. The process is conducted at gasifier pressure and 125 to 160 C. The proposed commercial embodiment of the SSRP involves a liquid phase of molten sulfur with dispersed catalyst in a slurry bubble-column reactor (SBCR).

Apostolos A. Nikolopoulos; Santosh K. Gangwal; William J. McMichael; Jeffrey W. Portzer

2003-01-01T23:59:59.000Z

11

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

SciTech Connect (OSTI)

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

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

2010-01-01T23:59:59.000Z

12

Advanced Concepts Breakout Group  

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

Workshop Workshop Advanced Concepts Working Group Facilitator: John J. Petrovic Scribe: Sherry Marin Advanced Storage Techniques/ Approaches in Priority Order 1. Crystalline Nanoporous Materials (15) 2. Polymer Microspheres (12) Self-Assembled Nanocomposites (12) 3. Advanced Hydrides (11) Metals - Organic (11) 4. BN Nanotubes (5) Hydrogenated Amorphous Carbon (5) 5. Mesoporous materials (4) Bulk Amorphous Materials (BAMs) (4) 6. Iron Hydrolysis (3) 7. Nanosize powders (2) 8. Metallic Hydrogen (1) Hydride Alcoholysis (1) Overarching R&D Questions for All Advanced Materials * Maximum storage capacity - theoretical model * Energy balance / life cycle analysis * Hydrogen absorption / desorption kinetics * Preliminary cost analysis - potential for low cost, high

13

Advanced Test Reactor Tour  

SciTech Connect (OSTI)

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

Miley, Don

2011-01-01T23:59:59.000Z

14

Advanced Test Reactor Tour  

ScienceCinema (OSTI)

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

Miley, Don

2013-05-28T23:59:59.000Z

15

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

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

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

16

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

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

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

17

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

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

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

18

Advanced Reactor Research and Development Funding Opportunity Announcement  

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

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

19

CESAR: Center for Exascale Simulation of Advanced Reactors | Argonne  

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

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

20

Inherent safety concepts in nuclear power reactors  

Science Journals Connector (OSTI)

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

O M Pal Singh; R Shankar Singh

1989-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "advanced reactor concepts" 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 next-generation reactor concept: The Integral Fast Reactor (IFR)  

SciTech Connect (OSTI)

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

Chang, Y.I.

1992-01-01T23:59:59.000Z

22

A next-generation reactor concept: The Integral Fast Reactor (IFR)  

SciTech Connect (OSTI)

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

Chang, Y.I.

1992-07-01T23:59:59.000Z

23

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

SciTech Connect (OSTI)

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

Weiss, A.J. (comp.)

1988-02-01T23:59:59.000Z

24

Advanced Nuclear Research Reactor  

SciTech Connect (OSTI)

This report describes technical modifications implemented by INVAP to improve the safety of the Research Reactors the company designs and builds.

Lolich, J.V.

2004-10-06T23:59:59.000Z

25

Advanced Reactor Research and Development Funding Opportunity Announcement  

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

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

26

E-Print Network 3.0 - advanced lwr concept Sample Search Results  

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

ENABLING SUSTAINABLE NUCLEAR POWER Summary: and NRE Design Class., "Advances in the Subcritical, Gas-Cooled Fast Transmutation Reactor Concept", Nucl... uranium energy...

27

E-Print Network 3.0 - advanced fusion concepts Sample Search...  

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

10.3 Alternate Concepts Fusion Technology FY 1995 -- 372.6 12;... International Thermonuclear Experimental Reactor Plasma Technologies Fusion Technologies Advanced Materials......

28

Advanced Reactor Research and Development Funding Opportunity...  

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

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

29

Advanced fusion concepts: project summaries  

SciTech Connect (OSTI)

This report contains descriptions of the activities of all the projects supported by the Advanced Fusion Concepts Branch of the Office of Fusion Energy, US Department of Energy. These descriptions are project summaries of each of the individual projects, and contain the following: title, principle investigators, funding levels, purpose, approach, progress, plans, milestones, graduate students, graduates, other professional staff, and recent publications. Information is given for each of the following programs: (1) reverse-field pinch, (2) compact toroid, (3) alternate fuel/multipoles, (4) stellarator/torsatron, (5) linear magnetic fusion, (6) liners, and (7) Tormac. (MOW)

None

1980-12-01T23:59:59.000Z

30

Advanced high performance solid wall blanket concepts C.P.C. Wong a,  

E-Print Network [OSTI]

and desirable attributes for the reactor design. Future needs and directions on the development of advanced FW of first wall coating material selection, design of plasma stabilization coils, consideration of reactor concepts for comparison

Raffray, A. René

31

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

Energy Savers [EERE]

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

32

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

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

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

33

Advanced Nuclear Reactors | Department of Energy  

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

Advanced Nuclear Advanced Nuclear Reactors Advanced Nuclear Reactors Turbulent Flow of Coolant in an Advanced Nuclear Reactor Visualizing Coolant Flow in Sodium Reactor Subassemblies Sodium-cooled Fast Reactor (SFR) Coolant Flow At the heart of a nuclear power plant is the reactor. The fuel assembly is placed inside a reactor vessel where all the nuclear reactions occur to produce the heat and steam used for power generation. Nonetheless, an entire power plant consists of many other support components and key structures like coolant pipes; pumps and tanks including their surrounding steel framing; and concrete containment and support structures. The Reactors Product Line within NEAMS is concerned with modeling the reactor vessel as well as those components of a complete power plant that

34

Development of a system model for advanced small modular reactors.  

SciTech Connect (OSTI)

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

Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

2014-01-01T23:59:59.000Z

35

Solar Energy Grid Integration Systems-Advanced Concepts | Department...  

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

Systems Integration Solar Energy Grid Integration Systems-Advanced Concepts Solar Energy Grid Integration Systems-Advanced Concepts On September 1, 2011, DOE announced 25.9...

36

Advanced Materials and Concepts for Portable Power Fuel Cells...  

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

Advanced Materials and Concepts for Portable Power Fuel Cells Advanced Materials and Concepts for Portable Power Fuel Cells These slides were presented at the 2010 New Fuel Cell...

37

Introduction to DMFCs - Advanced Materials and Concepts for Portable...  

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

DMFCs - Advanced Materials and Concepts for Portable Power Fuel Cells Introduction to DMFCs - Advanced Materials and Concepts for Portable Power Fuel Cells Presentation by Piotr...

38

Advanced Accelerator Concepts Final Report  

SciTech Connect (OSTI)

A major focus of research supported by this Grant has been on the ALPHA antihydrogen trap. We first trapped antihydrogen in 2010 and soon thereafter demonstrated trapping for 1000s. We now have observed resonant quantum interactions with antihydrogen. These papers in Nature and Nature Physics report the major milestones in anti-atom trapping. The success was only achieved through careful work that advanced our understanding of collective dynamics in charged particle systems, the development of new cooling and diagnostics, and in- novation in understanding how to make physics measurements with small numbers of anti-atoms. This research included evaporative cooling, autoresonant excitation of longitudinal motion, and centrifugal separation. Antihydrogen trapping by ALPHA is progressing towards the point when a important theories believed by most to hold for all physical systems, such as CPT (Charge-Parity-Time) invariance and the Weak Equivalence Principle (matter and antimatter behaving the same way under the influence of gravity) can be directly tested in a new regime. One motivation for this test is that most accepted theories of the Big Bang predict that we should observe equal amounts of matter and antimatter. However astrophysicists have found very little antimatter in the universe. Our experiment will, if successful over the next seven years, provide a new test of these ideas. Many earlier detailed and beautiful tests have been made, but the trapping of neutral antimatter allows us to explore the possibility of direct, model-independent tests. Successful cooling of the anti atoms, careful limits on systematics and increased trapping rates, all planned for our follow-up experiment (ALPHA-II) will reach unrivaled precision. CPT invariance implies that the spectra of hydrogen and antihydrogen should be identical. Spectra can be measured in principle with great precision, and any di#11;erences we might observe would revolutionize fundamental physics. This is the physics motivation for our experiment, one that requires only a few dozen researchers but must effectively integrate plasma, accelerator, atomic, and fundamental physics, as well as combine numerous technologies in the control, manipulation, and measurement of neutral and non-neutral particles. The ELENA ring (to which we hope to contribute, should funding be provided) is expect, when completed, to significantly enhance the performance of antihydrogen trapping by increasing by a factor of 100 the number of antiprotons that can be successfully trapped and cooled. ELENA operation is scheduled to commence in 2017. In collaboration with LBNL scientists, we proposed a frictional cooling scheme. This is an alternative cooling method to that used by ELENA. It is less complicated, experimentally unproven, and produces a lower yield of cold antiprotons. Students and postdoctoral researchers work on the trapping, cooling, transport, and nonlinear dynamics of antiprotons bunches that are provided by the AD to ALPHA; they contribute to the operation of the experiment, to software development, and to the design and operation of experiments. Students are expected to spend at summers at CERN while taking courses; after completion of courses they typically reside at CERN for most of the half-year run. The Antiproton Decelerator [AD] at CERN, along with its experiments, is the only facility in the world where antiprotons can be trapped and cooled and combined with positrons to form cold antihydrogen, with the ultimate goal of studying CPT violation and, subsequently, gravitational interactions of antimatter. Beyond the ALPHA experiment, the group worked on beam physics problems including limits on the average current in a time-dependent period cathode and new methods to create longitudinally coherent high repetition rate soft x-ray sources and wide bandwidth mode locked x-ray lasers. We completed a detailed study of quantum mechanical effects in the transit time cooling of muons.

Wurtele, Jonathan S.

2014-05-13T23:59:59.000Z

39

Mirror Advanced Reactor Study (MARS) final report summary  

SciTech Connect (OSTI)

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

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

1983-12-08T23:59:59.000Z

40

Advanced Safeguards Approaches for New Fast Reactors  

SciTech Connect (OSTI)

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

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

2007-12-15T23:59:59.000Z

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

Tribal Renewable Energy Advanced Course: Project Financing Concepts  

Broader source: Energy.gov [DOE]

Watch the DOE Office of Indian Energy's advanced renewable energy course entitled "Tribal Renewable Energy Project Development: Advanced Financing Concepts" by clicking on the .swf link below. You...

42

Evolution of the core physics concept for the Canadian supercritical water reactor  

SciTech Connect (OSTI)

The supercritical water cooled reactor (SCWR) is one of the advanced reactor concepts chosen by the GEN-IV International Forum (GIF) for research and development efforts. Canada's contribution is the Canadian SCWR, a heavy water moderated, pressure tube supercritical light water cooled reactor. Recent developments in the SCWR lattice and core concepts, primarily the introduction of a large central flow tube filled with coolant combined with a two-ring fuel assembly, have enabled significant improvements compared to earlier concepts. These improvements include a reduction in coolant void reactivity (CVR) by more than 10 mk, and an almost 40% increase in fuel exit burnup, which is achieved via balanced power distribution between the fuel pins in the fuel assembly. In this paper the evolution of the physics concept is reviewed, and the present lattice and core physics concepts are presented.

Pencer, J.; Colton, A.; Wang, X.; Gaudet, M.; Hamilton, H.; Yetisir, M. [Atomic Energy of Canada, Ltd., Chalk River Laboratories, Chalk River, ON (Canada)

2013-07-01T23:59:59.000Z

43

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

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

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

44

Advanced fusion MHD power conversion using the CFAR (compact fusion advanced Rankine) cycle concept  

SciTech Connect (OSTI)

The CFAR (compact fusion advanced Rankine) cycle concept for a tokamak reactor involves the use of a high-temperature Rankine cycle in combination with microwave superheaters and nonequilibrium MHD disk generators to obtain a compact, low-capital-cost power conversion system which fits almost entirely within the reactor vault. The significant savings in the balance-of-plant costs are expected to result in much lower costs of electricity than previous concepts. This paper describes the unique features of the CFAR cycle and a high- temperature blanket designed to take advantage of it as well as the predicted performance of the MHD disk generators using mercury seeded with cesium. 40 refs., 8 figs., 3 tabs.

Hoffman, M.A.; Campbell, R.; Logan, B.G. (California Univ., Davis, CA (USA); Lawrence Livermore National Lab., CA (USA))

1988-10-01T23:59:59.000Z

45

Burnup concept for a long-life fast reactor core using MCNPX.  

SciTech Connect (OSTI)

This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional Light Water Reactors (LWRs) or other SMR designs. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.

Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

2013-02-01T23:59:59.000Z

46

Advanced Test Reactor National Scientific User Facility  

SciTech Connect (OSTI)

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

Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

2011-08-01T23:59:59.000Z

47

Advanced Gasifier Pilot Plant Concept Definition  

SciTech Connect (OSTI)

This report presents results from definition of a preferred commercial-scale advanced gasifier configuration and concept definition for a gasification pilot plant incorporating those preferred technologies. The preferred commercial gasifier configuration was established based on Cost Of Electricity estimates for an IGCC. Based on the gasifier configuration trade study results, a compact plug flow gasifier, with a dry solids pump, rapid-mix injector, CMC liner insert and partial quench system was selected as the preferred configuration. Preliminary systems analysis results indicate that this configuration could provide cost of product savings for electricity and hydrogen ranging from 15%-20% relative to existing gasifier technologies. This cost of product improvement draws upon the efficiency of the dry feed, rapid mix injector technology, low capital cost compact gasifier, and >99% gasifier availability due to long life injector and gasifier liner, with short replacement time. A pilot plant concept incorporating the technologies associated with the preferred configuration was defined, along with cost and schedule estimates for design, installation, and test operations. It was estimated that a 16,300 kg/day (18 TPD) pilot plant gasifier incorporating the advanced gasification technology and demonstrating 1,000 hours of hot-fire operation could be accomplished over a period of 33 months with a budget of $25.6 M.

Steve Fusselman; Alan Darby; Fred Widman

2005-08-31T23:59:59.000Z

48

Sterile Neutrino Search Using China Advanced Research Reactor  

E-Print Network [OSTI]

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

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

2013-06-18T23:59:59.000Z

49

Mirror Advanced Reactor Study interim design report  

SciTech Connect (OSTI)

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

Not Available

1983-04-01T23:59:59.000Z

50

Advanced Nuclear Reactor Systems An Indian Perspective  

Science Journals Connector (OSTI)

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

Ratan Kumar Sinha

2011-01-01T23:59:59.000Z

51

Prospects for the development of advanced reactors  

SciTech Connect (OSTI)

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

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

1992-12-31T23:59:59.000Z

52

Ultrafast Carrier RelaxationProcesses in Advanced Concept Solar Cells  

Science Journals Connector (OSTI)

We discuss short time carrier relaxation in advanced concept solar cells conditions using ensemble Monte Carlo (EMC) simulation coupled with rate equation and thermodynamic models, to...

Goodnick, Stephen M; Honsberg, Christiana; Zou, Yongjie

53

Advanced Combustion Concepts - Enabling Systems and Solutions...  

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

engine installed and vehicle available for application, emission and fuel economy optimization with advanced combustion modes. 4 Advanced combustion control strategy, capable of...

54

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

SciTech Connect (OSTI)

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

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

1991-09-01T23:59:59.000Z

55

High Flux Isotope Reactor cold neutron source reference design concept  

SciTech Connect (OSTI)

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

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

1998-05-01T23:59:59.000Z

56

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

SciTech Connect (OSTI)

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

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

1996-07-01T23:59:59.000Z

57

Tribal Renewable Energy Advanced Course: Project Development Concepts |  

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

Concepts Concepts Tribal Renewable Energy Advanced Course: Project Development Concepts Watch the DOE Office of Indian Energy renewable energy course entitled "Tribal Renewable Energy Project Development: Advanced Concept Topics" by clicking on the .swf file below. You can also download a PDF of the PowerPoint slides. This course provides in-depth information on project development concepts for renewable energy projects on tribal lands, including: Risk and uncertainty Tribal project roles Policies and incentives See the full list of DOE Office of Indian Energy educational webinars and provide your feedback on the National Training & Education Resource (NTER) website. Renewable Energy Project Development: Advanced Concept Topics repd_project_development_concepts_lowder.swf

58

Advanced Combustion Concepts - Enabling Systems and Solutions...  

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

engine * Integration of proposed air path and HCCI combustion control strategies into ECU software * Prototype level 2 updates and proof of combustion concept for vehicle readiness...

59

Advanced Combustion Concepts - Enabling Systems and Solutions...  

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

Fuel efficiency as key market driver Stringent emission requirements System cost of advanced combustion Targets 30% fuel efficiency improvement SULEV emissions...

60

Advanced thermionic reactor systems design code  

SciTech Connect (OSTI)

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

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

1991-01-01T23:59:59.000Z

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

STATEMENT OF CONSIDERATIONS Advance Test Reactor Class Waiver  

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

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

62

Design and analysis of megawatt-class heat-pipe reactor concepts  

SciTech Connect (OSTI)

There is growing interest in finding an alternative to diesel-powered systems at locations removed from a reliable electrical grid. One promising option is a 1- to 10-MW mobile reactor system, that could provide robust, self-contained, and long-term ({>=} 5 years) power in any environment. The reactor and required infrastructure could be transported to any location within one or a few standard transport containers. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than 'traditional' reactors that rely on pumped coolant through the core. This paper examines a heat pipe reactor that is fabricated and shipped as six identical core segments. Each core segment includes a heat-pipe-to-gas heat exchanger that is coupled to the condenser end of the heat pipes. The reference power conversion system is a CO{sub 2}-Brayton system. The segments by themselves are deeply subcritical during transport, and they would be locked into an operating configuration (with control inserted) at the final destination. Two design options are considered: a near-term option and an advanced option. The near-term option is a 5-MWt concept that uses uranium-dioxide fuel, a stainless-steel structure, and potassium as the heat-pipe working fluid. The advanced option is a 15-MWt concept that uses uranium-nitride fuel, a molybdenum/TZM structure, and sodium as the heat-pipe working fluid. The materials used in the advanced option allow for higher temperatures and power densities, and enhanced power throughput in the heat pipes. Higher powers can be obtained from both concepts by increasing the core size and the number of heat pipes. (authors)

Poston, D.; Kapernick, R. [Los Alamos National Laboratory, MS C921, Los Alamos, NM 87545 (United States)

2012-07-01T23:59:59.000Z

63

Idaho National Laboratory Advanced Test Reactor Probabilistic Risk  

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

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

64

Process Technology and Advanced Concepts: Organic Solar Cells (Fact Sheet)  

SciTech Connect (OSTI)

Capabilities fact sheet for the National Center for Photovoltaics: Process Technology and Advanced Concepts: Organic Solar Cell that includes scope, core competencies and capabilities, and contact/web information.

Not Available

2011-06-01T23:59:59.000Z

65

Renewable Energy Project Development: Advanced Concept Topics  

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

Concept Topics Concept Topics An Introduction to Risk, Tribal Roles, and Support Policies in the Renewable Energy Project Development Process Course Outline What we will cover...  About the DOE Office of Indian Energy Education Initiative  Concepts and Policies for Understanding Renewable Energy Projects on Tribal Lands - Risk and Uncertainty - Tribal Project Roles - Policies and Incentives  Additional Information and Resources 2 Introduction The U.S. Department of Energy (DOE) Office of Indian Energy Policy and Programs is responsible for assisting Tribes with energy planning and development, infrastructure, energy costs, and electrification of Indian lands and homes. As part of this commitment and on behalf of DOE, the Office of Indian Energy is leading education

66

Review of High Temperature Water and Steam Cooled Reactor Concepts  

SciTech Connect (OSTI)

This review summarizes design concepts of supercritical-pressure water cooled reactors (SCR), nuclear superheaters and steam cooled fast reactors from 1950's to the present time. It includes water moderated supercritical steam cooled reactor, SCOTT-R and SC-PWR of Westinghouse, heavy water moderated light water cooled SCR of GE, SCLWR and SCFR of the University of Tokyo, B-500SKDI of Kurchatov Institute, CANDU -X of AECL, nuclear superheaters of GE, subcritical-pressure steam cooled FBR of KFK and B and W, Supercritical-pressure steam cooled FBR of B and W, subcritical-pressure steam cooled high converter by Edlund and Schultz and subcritical-pressure water-steam cooled FBR by Alekseev. This paper is prepared based on the previous review of SCR2000 symposium, and some author's comments are added. (author)

Oka, Yoshiaki [Nuclear Engineering Research Laboratory, The University of Tokyo, 3-1, Hongo 7-Chome, Bunkyo-ku (Japan)

2002-07-01T23:59:59.000Z

67

Concept of an inherently-safe high temperature gas-cooled reactor  

SciTech Connect (OSTI)

As the challenge to ensure no harmful release of radioactive materials at the accidents by deterministic approach instead to satisfy acceptance criteria or safety goal for risk by probabilistic approach, new concept of advanced reactor, an inherently-safe high temperature gas-cooled reactor, is proposed based on the experience of the operation of the actual High Temperature Gas-cooled Reactor (HTGR) in Japan, High Temperature Engineering Test Reactor (HTTR), and the design of the commercial plant (GTHTR300), utilizing the inherent safety features of the HTGR (i.e., safety features based on physical phenomena). The safety design philosophy of the inherently-safe HTGR for the safety analysis of the radiological consequences is determined as the confinement of radioactive materials is assured by only inherent safety features without engineered safety features, AC power or prompt actions by plant personnel if the design extension conditions occur. Inherent safety features to prevent the loss or degradation of the confinement function are identified. It is proposed not to apply the probabilistic approach for the evaluation of the radiological consequences of the accidents in the safety analysis because no inherent safety features fail for the mitigation of the consequences of the accidents. Consequently, there are no event sequences to harmful release of radioactive materials if the design extension conditions occur in the inherently-safe HTGR concept. The concept and future R and D items for the inherently-safe HTGR are described in this paper.

Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro [Nuclear Hydrogen and Heat Application Research Center, Japan Atomic Energy Agency, Oarai-machi, Ibaraki-ken, 311-1394 (Japan)

2012-06-06T23:59:59.000Z

68

Advanced concepts for controlling energy surety microgrids.  

SciTech Connect (OSTI)

Today, researchers, engineers, and policy makers are seeking ways to meet the world's growing demand for energy while addressing critical issues such as energy security, reliability, and sustainability. Many believe that distributed generators operating within a microgrid have the potential to address most of these issues. Sandia National Laboratories has developed a concept called energy surety in which five of these 'surety elements' are simultaneously considered: energy security, reliability, sustainability, safety, and cost-effectiveness. The surety methodology leads to a new microgrid design that we call an energy surety microgrid (ESM). This paper discusses the unique control requirement needed to produce a microgrid system that has high levels of surety, describes the control system from the most fundamental level through a real-world example, and discusses our ideas and concepts for a complete system.

Menicucci, David F.; Ortiz-Moyet, Juan

2011-05-01T23:59:59.000Z

69

E-Print Network 3.0 - advanced control concept Sample Search...  

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

concept Search Powered by Explorit Topic List Advanced Search Sample search results for: advanced control concept Page: << < 1 2 3 4 5 > >> 1 MIT LINCOLN LABORATORY ORGANIZATION OF...

70

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

SciTech Connect (OSTI)

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

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

1993-01-15T23:59:59.000Z

71

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

Office of Legacy Management (LM)

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

72

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

SciTech Connect (OSTI)

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

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

1994-06-01T23:59:59.000Z

73

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

SciTech Connect (OSTI)

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

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

2009-09-01T23:59:59.000Z

74

Argonne's pyroprocessing and advanced reactor research featured on WGN  

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

Argonne's pyroprocessing and advanced reactor research featured on WGN Argonne's pyroprocessing and advanced reactor research featured on WGN radio Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library About Nuclear Energy Nuclear Reactors Designed by Argonne Argonne's Nuclear Science and Technology Legacy Opportunities within NE Division Visit Argonne Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Celebrating the 70th Anniversary of Chicago Pile 1 (CP-1) Argonne OutLoud on Nuclear Energy Argonne Energy Showcase 2012 Highlights Bookmark and Share Argonne's pyroprocessing and advanced reactor research featured on WGN radio Uranium dendrites These tiny branches, or "dendrites," of pure uranium form when engineers

75

Baseline Concept Description of a Small Modular High Temperature Reactor  

SciTech Connect (OSTI)

The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both small or medium-sized and modular by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOEs ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities (Figure 1). The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.

Hans Gougar

2014-05-01T23:59:59.000Z

76

ECE533 Advanced MOS Concepts and VLSI Design Spring 2012  

E-Print Network [OSTI]

ECE533 Advanced MOS Concepts and VLSI Design Spring 2012 S.K. Islam, 504 Min Kao Building, 974, Prentice Hall/Pearson 2003, ISBN 0-13-090996-3 CMOS Digital Integrated Circuits Analysis and Design, Sung Circuit Design, Layout, and Simulation, IEEE Press, 1998. · Ken Martin, Digital Integrated Circuit Design

Tennessee, University of

77

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

SciTech Connect (OSTI)

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

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

2004-07-01T23:59:59.000Z

78

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

SciTech Connect (OSTI)

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

Sivill, R.L.

1990-03-01T23:59:59.000Z

79

Advanced Reactors Thermal Energy Transport for Process Industries  

SciTech Connect (OSTI)

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

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

2014-07-01T23:59:59.000Z

80

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

SciTech Connect (OSTI)

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

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

2005-10-01T23:59:59.000Z

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


81

DEVELOPMENT OF OPERATIONAL CONCEPTS FOR ADVANCED SMRs: THE ROLE OF COGNITIVE SYSTEMS ENGINEERING  

SciTech Connect (OSTI)

Advanced small modular reactors (AdvSMRs) will use advanced digital instrumentation and control systems, and make greater use of automation. These advances not only pose technical and operational challenges, but will inevitably have an effect on the operating and maintenance (O&M) cost of new plants. However, there is much uncertainty about the impact of AdvSMR designs on operational and human factors considerations, such as workload, situation awareness, human reliability, staffing levels, and the appropriate allocation of functions between the crew and various automated plant systems. Existing human factors and systems engineering design standards and methodologies are not current in terms of human interaction requirements for dynamic automated systems and are no longer suitable for the analysis of evolving operational concepts. New models and guidance for operational concepts for complex socio-technical systems need to adopt a state-of-the-art approach such as Cognitive Systems Engineering (CSE) that gives due consideration to the role of personnel. This approach we report on helps to identify and evaluate human challenges related to non-traditional concepts of operations. A framework - defining operational strategies was developed based on the operational analysis of Argonne National Laboratorys Experimental Breeder Reactor-II (EBR-II), a small (20MWe) sodium-cooled reactor that was successfully operated for thirty years. Insights from the application of the systematic application of the methodology and its utility are reviewed and arguments for the formal adoption of CSE as a value-added part of the Systems Engineering process are presented.

Jacques Hugo; David Gertman

2014-04-01T23:59:59.000Z

82

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor  

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

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

83

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor  

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

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

84

Partnerships Help Advance Small Modular Reactor Technology | Department of  

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

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

85

Development of environmentally advanced hydropower turbine system design concepts  

SciTech Connect (OSTI)

A team worked together on the development of environmentally advanced hydro turbine design concepts to reduce hydropower`s impact on the environment, and to improve the understanding of the technical and environmental issues involved, in particular, with fish survival as a result of their passage through hydro power sites. This approach brought together a turbine design and manufacturing company, biologists, a utility, a consulting engineering firm and a university research facility, in order to benefit from the synergy of diverse disciplines. Through a combination of advanced technology and engineering analyses, innovative design concepts adaptable to both new and existing hydro facilities were developed and are presented. The project was divided into 4 tasks. Task 1 investigated a broad range of environmental issues and how the issues differed throughout the country. Task 2 addressed fish physiology and turbine physics. Task 3 investigated individual design elements needed for the refinement of the three concept families defined in Task 1. Advanced numerical tools for flow simulation in turbines are used to quantify characteristics of flow and pressure fields within turbine water passageways. The issues associated with dissolved oxygen enhancement using turbine aeration are presented. The state of the art and recent advancements of this technology are reviewed. Key elements for applying turbine aeration to improve aquatic habitat are discussed and a review of the procedures for testing of aerating turbines is presented. In Task 4, the results of the Tasks were assembled into three families of design concepts to address the most significant issues defined in Task 1. The results of the work conclude that significant improvements in fish passage survival are achievable.

Franke, G.F.; Webb, D.R.; Fisher, R.K. Jr. [Voith Hydro, Inc. (United States)] [and others

1997-08-01T23:59:59.000Z

86

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

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

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

87

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

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

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

88

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

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

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

89

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

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

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

90

Testing of Passive Safety System Performance for Higher Power Advanced Reactors  

SciTech Connect (OSTI)

This report describes the results of NERI research on the testing of advanced passive safety performance for the Westinghouse AP1000 design. The objectives of this research were: (a) to assess the AP1000 passive safety system core cooling performance under high decay power conditions for a spectrum of breaks located at a variety of locations, (b) to compare advanced thermal hydraulic computer code predictions to the APEX high decay power test data and (c) to develop new passive safety system concepts that could be used for Generation IV higher power reactors.

brian G. Woods; Jose Reyes, Jr.; John Woods; John Groome; Richard Wright

2004-12-31T23:59:59.000Z

91

Evaluation of Concepts for Mulitiple Application Thermal Reactor for Irradiation eXperiments (MATRIX)  

SciTech Connect (OSTI)

The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Originally operated primarily in support of the Offcie of Naval Reactors (NR), the mission has gradually expanded to cater to other customers, such as the DOE Office of Nuclear Energy (NE), private industry, and universities. Unforeseen circumstances may lead to the decommissioning of ATR, thus leaving the U.S. Government without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. This work can be viewed as an update to a project from the 1990s called the Broad Application Test Reactor (BATR). In FY 2012, a survey of anticipated customer needs was performed, followed by analysis of the original BATR concepts with fuel changed to low-enriched uranium. Departing from these original BATR designs, four concepts were identified for further analysis in FY2013. The project informally adopted the acronym MATRIX (Multiple-Application Thermal Reactor for Irradiation eXperiments). This report discusses analysis of the four MATRIX concepts along with a number of variations on these main concepts. Designs were evaluated based on their satisfaction of anticipated customer requirements and the Cylindrical variant was selected for further analysis of options. This downselection should be considered preliminary and the backup alternatives should include the other three main designs. The baseline Cylindrical MATRIX design is expected to be capable of higher burnup than the ATR (or longer cycle length given a particular batch scheme). The volume of test space in IPTs is larger in MATRIX than in ATR with comparable magnitude of neutron flux. In addition to the IPTs, the Cylindrical MATRIX concept features test spaces at the centers of fuel assemblies where very high fast flux can be achieved. This magnitude of fast flux is similar to that achieved in the ATR A-positions, however, the available volume having these conditions is greater in the MATRIX design than in the ATR. From the analyses performed in this work, it appears that the Cylindrical MATRIX design can be designed to meet the anticipated needs of the ATR replacement reactor. However, this statement must be qualified by acknowledging that this design is quite immature, and therefore any requirements currently met must be re-evaluated as the design matures. Also, some of the requirements were not strictly met, but are believed to be achievable once features to be added later are designed.

Michael A. Pope; Hans D. Gougar; John M. Ryskamp

2013-09-01T23:59:59.000Z

92

E-Print Network 3.0 - advanced accelerator concepts Sample Search...  

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

Summary: of an advanced exotic beam facility evolved from the Rare Isotope Accelerator (RIA) concept. The OMB and the DOE... to the advance of the accelerator physics...

93

Advanced Neutron Source Reactor thermal analysis of fuel plate defects  

SciTech Connect (OSTI)

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

Giles, G.E.

1995-08-01T23:59:59.000Z

94

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

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

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

95

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

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

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

96

Draft Function Allocation Framework and Preliminary Technical Basis for Advanced SMR Concepts of Operations  

SciTech Connect (OSTI)

This report presents preliminary research results from the investigation into the development of new models and guidance for Concepts of Operations in advanced small modular reactor (AdvSMR) designs. AdvSMRs are nuclear power plants (NPPs), but unlike conventional large NPPs that are constructed on site, AdvSMRs systems and components will be fabricated in a factory and then assembled on site. AdvSMRs will also use advanced digital instrumentation and control systems, and make greater use of automation. Some AdvSMR designs also propose to be operated in a multi-unit configuration with a single central control room as a way to be more cost-competitive with existing NPPs. These differences from conventional NPPs not only pose technical and operational challenges, but they will undoubtedly also have regulatory compliance implications, especially with respect to staffing requirements and safety standards.

Jacques Hugo; John Forester; David Gertman; Jeffrey Joe; Heather Medema; Julius Persensky; April Whaley

2013-08-01T23:59:59.000Z

97

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

Science Journals Connector (OSTI)

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

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

2010-01-01T23:59:59.000Z

98

Integrated intelligent systems in advanced reactor control rooms  

SciTech Connect (OSTI)

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

Beckmeyer, R.R.

1989-01-01T23:59:59.000Z

99

Generic Repository Concepts and Thermal Analysis for Advanced Fuel Cycles  

SciTech Connect (OSTI)

The current posture of the used nuclear fuel management program in the U.S. following termination of the Yucca Mountain Project, is to pursue research and development (R&D) of generic (i.e., non-site specific) technologies for storage, transportation and disposal. Disposal R&D is directed toward understanding and demonstrating the performance of reference geologic disposal concepts selected to represent the current state-of-the-art in geologic disposal. One of the principal constraints on waste packaging and emplacement in a geologic repository is management of the waste-generated heat. This paper describes the selection of reference disposal concepts, and thermal management strategies for waste from advanced fuel cycles. A geologic disposal concept for spent nuclear fuel (SNF) or high-level waste (HLW) consists of three components: waste inventory, geologic setting, and concept of operations. A set of reference geologic disposal concepts has been developed by the U.S. Department of Energy (DOE) Used Fuel Disposition Campaign, for crystalline rock, clay/shale, bedded salt, and deep borehole (crystalline basement) geologic settings. We performed thermal analysis of these concepts using waste inventory cases representing a range of advanced fuel cycles. Concepts of operation consisting of emplacement mode, repository layout, and engineered barrier descriptions, were selected based on international progress and previous experience in the U.S. repository program. All of the disposal concepts selected for this study use enclosed emplacement modes, whereby waste packages are in direct contact with encapsulating engineered or natural materials. The encapsulating materials (typically clay-based or rock salt) have low intrinsic permeability and plastic rheology that closes voids so that low permeability is maintained. Uniformly low permeability also contributes to chemically reducing conditions common in soft clay, shale, and salt formations. Enclosed modes are associated with temperature constraints that limit changes to the encapsulating materials, and they generally have less capacity to dissipate heat from the waste package and its immediate surroundings than open modes such as that proposed for a repository at Yucca Mountain, Nevada. Open emplacement modes can be ventilated for many years prior to permanent closure of the repository, limiting peak temperatures both before and after closure, and combining storage and disposal functions in the same facility. Open emplacement modes may be practically limited to unsaturated host formations, unless emplacement tunnels are effectively sealed everywhere prior to repository closure. Thermal analysis of disposal concepts and waste inventory cases has identified important relationships between waste package size and capacity, and the duration of surface decay storage needed to meet temperature constraints. For example, the choice of salt as the host medium expedites the schedule for geologic disposal by approximately 50 yr (other factors held constant) thereby reducing future reliance on surface decay storage. Rock salt has greater thermal conductivity and stability at higher temperatures than other media considered. Alternatively, the choice of salt permits the use of significantly larger waste packages for SNF. The following sections describe the selection of reference waste inventories, geologic settings, and concepts of operation, and summarize the results from the thermal analysis.

Hardin, Ernest [Sandia National Laboratories (SNL)] [Sandia National Laboratories (SNL); Blink, James [Lawrence Livermore National Laboratory (LLNL)] [Lawrence Livermore National Laboratory (LLNL); Carter, Joe [Savannah River National Laboratory (SRNL)] [Savannah River National Laboratory (SRNL); Massimiliano, Fratoni [Lawrence Livermore National Laboratory (LLNL)] [Lawrence Livermore National Laboratory (LLNL); Greenberg, Harris [Lawrence Livermore National Laboratory (LLNL)] [Lawrence Livermore National Laboratory (LLNL); Howard, Rob L [ORNL] [ORNL

2011-01-01T23:59:59.000Z

100

A Blueprint for GNEP Advanced Burner Reactor Startup Fuel Fabrication Facility  

SciTech Connect (OSTI)

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

S. Khericha

2010-12-01T23:59:59.000Z

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

Distillation and Dehydro Reactors Advanced Process Conrol Freeport Texas PLant  

E-Print Network [OSTI]

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

Eisele, D.

2014-01-01T23:59:59.000Z

102

Concept development of rotating bed chemical looping combustion reactor:.  

E-Print Network [OSTI]

??In this research a new rotary chemical looping combustion (CLC) reactor was developed which is suitable for larger scales and solves some of the issues (more)

Hermans, C.W.M.

2013-01-01T23:59:59.000Z

103

Development of advanced strain diagnostic techniques for reactor environments.  

SciTech Connect (OSTI)

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

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

2013-02-01T23:59:59.000Z

104

Advanced Test Reactor National Scientific User Facility Partnerships  

SciTech Connect (OSTI)

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

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

2012-03-01T23:59:59.000Z

105

Advanced Materials and Concepts for Portable Power Fuel Cells  

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

1 1 1 DOE Kick-off Meeting, Washington, DC September 28, 2010 Fuel Cell Projects Kick-off Meeting Washington, DC - September 28, 2010 Advanced Materials and Concepts for Portable Power Fuel Cells for Portable Power Fuel Cells Piotr Zelenay Los Alamos National Laboratory Los Alamos National Laboratory Los Alamos, New Mexico 87545 This presentation does not contain any proprietary, confidential, or otherwise restricted information - t t Overview Timeline * Start date: September 2010 * End date: Four-year duration Budget Budget * Total funding estimate: - DOE share: $3,825K Contractor share: $342K $342K - Contractor share: * FY10 funding received: $250K * FY11 funding estimate: $1,000K Barriers * A. Durability (catalyst; electrode) (catalyst; electrode)

106

The Advanced Test Reactor National Scientific User Facility  

SciTech Connect (OSTI)

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

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

2011-08-01T23:59:59.000Z

107

Enhanced in-pile instrumentation at the advanced test reactor  

SciTech Connect (OSTI)

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

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

2011-07-01T23:59:59.000Z

108

Evaluation of the parfait blanket concept for fast breeder reactors  

E-Print Network [OSTI]

An evaluation of the neutronic, thermal-hydraulic, mechanical and economic characteristics of fast breeder reactor configurations containing an internal blanket has been performed. This design, called the parfait blanket ...

Ducat, Glenn Alexander

1974-01-01T23:59:59.000Z

109

Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics  

SciTech Connect (OSTI)

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

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

2014-02-01T23:59:59.000Z

110

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

SciTech Connect (OSTI)

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

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

2002-07-01T23:59:59.000Z

111

Advanced High-Temperature, High-Pressure Transport Reactor Gasification  

SciTech Connect (OSTI)

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

Michael Swanson; Daniel Laudal

2008-03-31T23:59:59.000Z

112

Thermo-Economic Assessment of Advanced,High-Temperature CANDU Reactors  

SciTech Connect (OSTI)

Research underway on the advanced CANDU examines new, innovative, reactor concepts with the aim of significant cost reduction and resource sustainability through improved thermodynamic efficiency and plant simplification. The so-called CANDU-X concept retains the key elements of the current CANDU designs, including heavy-water moderator that provides a passive heat sink and horizontal pressure tubes. Improvement in thermodynamic efficiency is sought via substantial increases in both pressure and temperature of the reactor coolant. Following on from the new Next Generation (NG) CANDU, which is ready for markets in 2005 and beyond, the reactor coolant is chosen to be light water but at supercritical operating conditions. Two different temperature regimes are being studied, Mark 1 and Mark 2, based respectively on continued use of zirconium or on stainless-steel-based fuel cladding. Three distinct cycle options have been proposed for Mark 1: the High-Pressure Steam Generator (HPSG) cycle, the Dual cycle, and the Direct cycle. For Mark 2, the focus is on simplification via a Direct cycle. This paper presents comparative thermo-economic assessments of the CANDU-X cycle options, with the ultimate goal of ascertaining which particular cycle option is the best overall in terms of thermodynamics and economics. A similar assessment was already performed for the NG CANDU. The economic analyses entail obtaining cost estimates of major plant components, such as heat exchangers, turbines and pumps. (authors)

Spinks, Norman J.; Pontikakis, Nikos; Duffey, Romney B. [Atomic Energy of Canada Limited, Chalk River, ON KOJ 1J0 (Canada)

2002-07-01T23:59:59.000Z

113

Suppression of thermal stratification in gravity driven water pool of an advanced reactor using shrouds  

Science Journals Connector (OSTI)

Abstract Advanced and innovative reactor systems are considering the use of large pools as heat sink for various safety functions like decay heat removal or containment cooling. These designs generally have heat exchangers immersed in the pool. For enhanced safety and reliability, preferred heat transfer mode is considered to be passive so that heat sink availability is maintained even in failure of power supply and active components. However, heat transfer by natural convection in large pools poses a problem of thermal stratification. As a result of natural convection, hot layers of water may accumulate over the relatively cold one and in turn inhibit the natural convection itself. Not only the heat transfer performance may get deteriorated but some structural parts of the pool like concrete wall may be subjected to high temperature which is not desirable. In this paper, a new concept of employing shrouds around the heat source is proposed. These shrouds provide multiple natural circulation loops around the heat source, thereby facilitating mixing of hot and cold fluid, which eliminate stratification. The concept has been applied to the Gravity Driven Water Pool (GDWP) of Advance Heavy Water Reactor (AHWR) in which Isolation Condensers (ICs) tubes are submerged for decay heat removal of AHWR using ICS and thermal stratification phenomenon was predicted without and with ICS. Results indicate that the shrouds have application in elimination of thermal stratification in GDWP.

P.K. Verma; A.K. Nayak; Vikas Jain; P.K. Vijayan; K.K. Vaze

2013-01-01T23:59:59.000Z

114

Metal fires and their implications for advanced reactors.  

SciTech Connect (OSTI)

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

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

2010-10-01T23:59:59.000Z

115

Concept of nuclear reactor pumped laser for ICF  

SciTech Connect (OSTI)

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

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

1996-05-01T23:59:59.000Z

116

Effects of an Advanced Reactors Design, Use of Automation, and Mission on Human Operators  

SciTech Connect (OSTI)

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

Jeffrey C. Joe; Johanna H. Oxstrand

2014-06-01T23:59:59.000Z

117

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

SciTech Connect (OSTI)

The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Programs understanding of the cost drivers that will determine nuclear powers 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

118

Fabrication development for the Advanced Neutron Source Reactor  

SciTech Connect (OSTI)

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

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

1995-08-01T23:59:59.000Z

119

Feasibility Investigation of the Decay Heat Removal Capability Using the Concept of a Thermosyphon in the Liquid Metal Reactor  

SciTech Connect (OSTI)

A new design concept for a decay heat removal system in a liquid metal reactor is proposed. The new design utilizes a thermosyphon to enhance the heat removal capacity and its heat transfer characteristics are analyzed against the current PSDRS (Passive Safety Decay heat Removal System) in the KALIMER (Korea Advanced Liquid Metal Reactor) design. The preliminary analysis results show that the new design with a thermosyphon yields substantial increase of 20{approx}40% in the decay heat removal capacity compared to the current design that do not have the thermosyphon. The new design reduces the temperature rise in the cooling air of the system and helps the surrounding structure in maintaining its mechanical integrity for long term operation at an accident. Also the analysis revealed the characteristics of the interactions among various heat transfer modes in the new design. (authors)

Yeon-Sik Kim; Yoon-Sub Sim; Eui-Kwang Kim [Korea Atomic Energy Research Institute, 150, Dukjin-Dong, Yusong-Gu, Taejon 305-353 (Korea, Republic of)

2002-07-01T23:59:59.000Z

120

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

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

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

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

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

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

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

122

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

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

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

123

Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation  

E-Print Network [OSTI]

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

Heidet, Florent

2010-01-01T23:59:59.000Z

124

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

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

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

125

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

SciTech Connect (OSTI)

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

Curtis Smith; Steven Prescott; Tony Koonce

2014-04-01T23:59:59.000Z

126

Reference modular High Temperature Gas-Cooled Reactor Plant: Concept description report  

SciTech Connect (OSTI)

This report provides a summary description of the Modular High Temperature Gas-Cooled Reactor (MHTGR) concept and interim results of assessments of costs, safety, constructibility, operability, maintainability, and availability. Conceptual design of this concept was initiated in October 1985 and is scheduled for completion in 1987. Participating industrial contractors are Bechtel National, Inc. (BNI), Stone and Webster Engineering Corporation (SWEC), GA Technologies, Inc. (GA), General Electric Co. (GE), and Combustion Engineering, Inc. (C-E).

Not Available

1986-10-01T23:59:59.000Z

127

Safety Assurance for Irradiating Experiments in the Advanced Test Reactor  

SciTech Connect (OSTI)

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

T. A. Tomberlin; S. B. Grover

2004-11-01T23:59:59.000Z

128

Thermochemical Solar Energy Storage Via Redox Oxides: Materials and Reactor/Heat Exchanger Concepts  

Science Journals Connector (OSTI)

Abstract Thermochemical Storage of solar heat exploits the heat effects of reversible chemical reactions for the storage of solar energy. Among the possible reversible gas-solid chemical reactions, the utilization of a pair of redox reactions of multivalent solid oxides can be directly coupled to CSP plants employing air as the heat transfer fluid bypassing the need for a separate heat exchanger. The present work concerns the development of thermochemical storage systems based on such oxide-based redox materials and in particular on cobalt oxide; in the one hand by tailoring their heat storage/release capability and on the other hand via their incorporation in proper reactor/heat exchanger devices. In this respect the first stage of the work involved parametric testing of cobalt oxide compositions via Thermo-Gravimetric Analysis to comparatively investigate the temperature range for cyclic oxidation-reduction and optimize the cycle conditions for maximum reduction and re-oxidation extent. Subsequently, two reactor concepts for the coupling of solar energy to the redox reactions have been implemented and tested. These reactor concepts include in one hand structured ceramic reactors/heat exchangers based on redox-oxide-coated honeycombs and on the other hand powder-fed, solar-heated, rotary kiln reactors. The two reactor concepts were tested within non-solar-aided lab-scale and solar- aided campaigns, respectively. The feasibility of both concepts was shown and good chemical conversions were achieved. The experiments pointed out the challenging points related to the manufacture of pilot-scale reactors/heat exchangers with enhanced heat storage capacity. A numerical model using commercial CFD software is developed to define optimal geometrical characteristics and operating conditions and refine the pilot scale design in order to achieve efficient, long-term off-sun operation.

S. Tescari; C. Agrafiotis; S. Breuer; L. de Oliveira; M. Neises-von Puttkamer; M. Roeb; C. Sattler

2014-01-01T23:59:59.000Z

129

Advanced neutron source reactor probabilistic flow blockage assessment  

SciTech Connect (OSTI)

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

Ramsey, C.T.

1995-08-01T23:59:59.000Z

130

Fuel qualification plan for the Advanced Neutron Source Reactor  

SciTech Connect (OSTI)

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

Copeland, G.L.

1995-07-01T23:59:59.000Z

131

Advanced Reactor Innovation Evaluation Study (ARIES) Properties Archive  

DOE Data Explorer [Office of Scientific and Technical Information (OSTI)]

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

132

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

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

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

133

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

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

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

134

Advanced HEV/PHEV Concepts | Department of Energy  

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

Technologies Office Merit Review 2014: In-Vehicle Evaluation of Lower-Energy Energy Storage System (LEESS) Devices Overview and Progress of United States Advanced Battery...

135

Reactor prospect of spheromak concept by electrostatic helicity injection  

Science Journals Connector (OSTI)

The highest performing spheromaks in the laboratory are formed by electrostatic helicity injection. Discharges with up to 1 MA plasma current and core electron temperature as high as 500 eV have been recently obtained. For such a scheme to scale-up to a reactor however a much higher current multiplication factor (plasma current over injector current) must be achieved. It is shown here that spheromak current multiplication is linearly proportional to flux amplification (ratio of poloidal fluxes inside and outside the separatrix of the mean field). Hence spheromak optimization is centered around achieving high flux amplification which is provided by linear or nonlinear resonant coupling between helicity injector and the spheromak force-free eigenmode. The nonlinear resonant field amplification is the most promising route to high flux amplification in a realistic plasma that often significantly deviates from the Taylor state. Accessing such nonlinear resonant field amplification can be facilitated by auxiliary current drive around the magnetic axis and auxiliary heating to break the electron temperature constraint on flux amplification.

X. Z. Tang; A. H. Boozer

2008-01-01T23:59:59.000Z

136

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

SciTech Connect (OSTI)

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

Not Available

1992-04-01T23:59:59.000Z

137

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

SciTech Connect (OSTI)

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

Shannon Bragg-Sitton

2014-02-01T23:59:59.000Z

138

Work Domain Analysis of a Predecessor Sodium-cooled Reactor as Baseline for AdvSMR Operational Concepts  

SciTech Connect (OSTI)

This report presents the results of the Work Domain Analysis for the Experimental Breeder Reactor (EBR-II). This is part of the phase of the research designed to incorporate Cognitive Work Analysis in the development of a framework for the formalization of an Operational Concept (OpsCon) for Advanced Small Modular Reactors (AdvSMRs). For a new AdvSMR design, information obtained through Cognitive Work Analysis, combined with human performance criteria, can and should be used in during the operational phase of a plant to assess the crew performance aspects associated with identified AdvSMR operational concepts. The main objective of this phase was to develop an analytical and descriptive framework that will help systems and human factors engineers to understand the design and operational requirements of the emerging generation of small, advanced, multi-modular reactors. Using EBR-II as a predecessor to emerging sodium-cooled reactor designs required the application of a method suitable to the structured and systematic analysis of the plant to assist in identifying key features of the work associated with it and to clarify the operational and other constraints. The analysis included the identification and description of operating scenarios that were considered characteristic of this type of nuclear power plant. This is an invaluable aspect of Operational Concept development since it typically reveals aspects of future plant configurations that will have an impact on operations. These include, for example, the effect of core design, different coolants, reactor-to-power conversion unit ratios, modular plant layout, modular versus central control rooms, plant siting, and many more. Multi-modular plants in particular are expected to have a significant impact on overall OpsCon in general, and human performance in particular. To support unconventional modes of operation, the modern control room of a multi-module plant would typically require advanced HSIs that would provide sophisticated operational information visualization, coupled with adaptive automation schemes and operator support systems to reduce complexity. These all have to be mapped at some point to human performance requirements. The EBR-II results will be used as a baseline that will be extrapolated in the extended Cognitive Work Analysis phase to the analysis of a selected advanced sodium-cooled SMR design as a way to establish non-conventional operational concepts. The Work Domain Analysis results achieved during this phase have not only established an organizing and analytical framework for describing existing sociotechnical systems, but have also indicated that the method is particularly suited to the analysis of prospective and immature designs. The results of the EBR-II Work Domain Analysis have indicated that the methodology is scientifically sound and generalizable to any operating environment.

Ronald Farris; David Gertman; Jacques Hugo

2014-03-01T23:59:59.000Z

139

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

SciTech Connect (OSTI)

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

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

1995-12-31T23:59:59.000Z

140

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

SciTech Connect (OSTI)

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

John Jackson; Todd Allen; Frances Marshall; Jim Cole

2013-03-01T23:59:59.000Z

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


141

Advances in Metallic Nuclear Fuel  

Science Journals Connector (OSTI)

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

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

1987-04-01T23:59:59.000Z

142

Advanced Fusion Reactors for Space Propulsion and Power Systems  

SciTech Connect (OSTI)

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

Chapman, John J.

2011-06-15T23:59:59.000Z

143

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

SciTech Connect (OSTI)

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

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

2006-07-01T23:59:59.000Z

144

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

SciTech Connect (OSTI)

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

Weiss, A.J. (comp.)

1987-02-01T23:59:59.000Z

145

Study of plutonium disposition using existing GE advanced Boiling Water Reactors  

SciTech Connect (OSTI)

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

Not Available

1994-06-01T23:59:59.000Z

146

Advanced Underground Gas Storage Concepts Refrigerated-Mined Cavern Storage  

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

UNDERGROUND GAS STORAGE CONCEPTS UNDERGROUND GAS STORAGE CONCEPTS REFRIGERATED-MINED CAVERN STORAGE FINAL REPORT DOE CONTRACT NUMBER DE-AC26-97FT34349 SUBMITTED BY: PB-KBB INC. 11757 KATY FREEWAY, SUITE 600 HOUSTON, TX 77079 SEPTEMBER 1998 Disclaimer This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily

147

Technical Needs for Prototypic Prognostic Technique Demonstration for Advanced Small Modular Reactor Passive Components  

SciTech Connect (OSTI)

This report identifies a number of requirements for prognostics health management of passive systems in AdvSMRs, documents technical gaps in establishing a prototypical prognostic methodology for this purpose, and describes a preliminary research plan for addressing these technical gaps. AdvSMRs span multiple concepts; therefore a technology- and design-neutral approach is taken, with the focus being on characteristics that are likely to be common to all or several AdvSMR concepts. An evaluation of available literature is used to identify proposed concepts for AdvSMRs along with likely operational characteristics. Available operating experience of advanced reactors is used in identifying passive components that may be subject to degradation, materials likely to be used for these components, and potential modes of degradation of these components. This information helps in assessing measurement needs for PHM systems, as well as defining functional requirements of PHM systems. An assessment of current state-of-the-art approaches to measurements, sensors and instrumentation, diagnostics and prognostics is also documented. This state-of-the-art evaluation, combined with the requirements, may be used to identify technical gaps and research needs in the development, evaluation, and deployment of PHM systems for AdvSMRs. A preliminary research plan to address high-priority research needs for the deployment of PHM systems to AdvSMRs is described, with the objective being the demonstration of prototypic prognostics technology for passive components in AdvSMRs. Greater efficiency in achieving this objective can be gained through judicious selection of materials and degradation modes that are relevant to proposed AdvSMR concepts, and for which significant knowledge already exists. These selections were made based on multiple constraints including the analysis performed in this document, ready access to laboratory-scale facilities for materials testing and measurement, and potential synergies with other national laboratory and university partners.

Meyer, Ryan M.; Coble, Jamie B.; Hirt, Evelyn H.; Ramuhalli, Pradeep; Mitchell, Mark R.; Wootan, David W.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

2013-05-17T23:59:59.000Z

148

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

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

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

149

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

Office of Energy Efficiency and Renewable Energy (EERE)

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

150

The Advanced Candu reactor annunciation system - Compliance with IEC standard and US NRC guidelines  

SciTech Connect (OSTI)

Annunciation is a key plant information system that alerts Operations staff to important changes in plant processes and systems. Operational experience at nuclear stations worldwide has shown that many annunciation implementations do not provide the support needed by Operations staff in all plant situations. To address utility needs for annunciation improvement in Candu plants, AECL in partnership with Canadian Candu utilities, undertook an annunciation improvement program in the early nineties. The outcome of the research and engineering development program was the development and simulator validation of alarm processing, display, and information presentation techniques that provide practical and effective solutions to key operational deficiencies with earlier annunciation implementations. The improved annunciation capabilities consist of a series of detection, information processing and presentation functions called the Candu Annunciation Message List System (CAMLS). The CAMLS concepts embody alarm processing, presentation and interaction techniques, and strategies and methods for annunciation system configuration to ensure improved annunciation support for all plant situations, especially in upset situations where the alarm generation rate is high. The Advanced Candu Reactor (ACR) project will employ the CAMLS annunciation concepts as the basis for primary annunciation implementations. The primary annunciation systems will be implemented from CAMLS applications hosted on AECL Advanced Control Centre Information System (ACCIS) computing technology. The ACR project has also chosen to implement main control room annunciation aspects in conformance with the following international standard and regulatory review guide for control room annunciation practice: International Electrotechnical Commission (IEC) 62241 - Main Control Room, Alarm Function and Presentation (International standard) US NRC NUREG-0700 - Human-System Interface Design Review Guidelines, Section 4 - Alarm System (Regulatory review guide) In order for the ACR annunciation system to comply with IEC-62241 and NUREG-0700, the CAMLS concepts must satisfy their respective requirements and guidelines. This paper will provide a summary of how CAMLS complies with IEC-62241 and NUREG-0700. Overall, the CAMLS-based annunciation requirements conform to the intent of the recommended annunciation practices included in both these documents. However, there are minor differences in requirements interpretation and annunciation practice between CAMLS implementation concepts and the recommended practice in each document. Generally the differences with IEC-62241 revolve around definitions, the display properties of new alarms, and video display unit (VDU) alarm display organization. The differences with NUREG-0700 revolve around segregation of status alarms, the use of spatially dedicated, continuously visible displays, simultaneous display of high-priority alarms, audible signals for alarm states and reminder signals, blank lines in alarm message lists and manual silencing. These minor differences are discussed and rationales for the CAMLS implementation concepts are provided. (authors)

Leger, R.; Malcolm, S. [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ont. L5K 1B2 (Canada); Davey, E. [Crew Systems Solutions, Deep River, Ont. K0J 1P0 (Canada)

2006-07-01T23:59:59.000Z

151

ADVANCED DIRECT LIQUEFACTION CONCEPTS FOR PETC GENERIC UNITS  

SciTech Connect (OSTI)

The results of Laboratory and Bench-Scale experiments and supporting technical and economic assessments conducted under DOE Contract No. DE-AC22-91PC91040 is reported for the period April 1, 1998 to June 30, 1998. This contract is with the University of kentucky Research Foundation, which supports work with the University of Kentucky Center for Applied Energy Researc, CONSOL, Inc., LDP Associates, and Hydrocarbon Technologies, Inc. This work involves the introduction into the basic two-stage liquefaction process several novel concepts, which include dispersed lower-cost catalysts, coal cleaning by oil agglomeration, and distillate hydrotreating and dewaxing.

Adam J. Berkovich

2000-03-01T23:59:59.000Z

152

Assessment of Silicon Carbide Composites for Advanced Salt-Cooled Reactors  

SciTech Connect (OSTI)

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

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

2007-09-01T23:59:59.000Z

153

Range of blanket concepts from near term solutions to advanced concepts  

E-Print Network [OSTI]

a considerable development risk. Therefore, the selection of blanket concepts depends on the overall strategy which can be operated safely, with minimum impact on the environment, and at an acceptable cost to the environment, and; Á/ can generate electricity at acceptable costs. The evaluation of whether or not and how

Raffray, A. René

154

New Reactor Concepts and New Nuclear Data Needed to Develop Them  

Science Journals Connector (OSTI)

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

S. Ganesan

2005-01-01T23:59:59.000Z

155

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

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

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

156

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

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

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

157

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

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

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

158

Advanced atomization concept for CWF burning in small combustors  

SciTech Connect (OSTI)

The present project involves the second phase of research on a new concept in coal-water fuel (CWF) atomization that is applicable to burning in small combustors. It is intended to address the most important problem associated with CWF combustion; i.e., production of small spray droplets in an efficient manner by an atomization device. Phase 1 of this work was successfully completed with the development of an opposed-jet atomizer that met the goals of the first contract. Performance as a function of operating conditions was measured, and the technical feasibility of the device established in the Atlantic Research Atomization Test Facility employing a Malvern Particle Size Analyzer. Testing then proceeded to a combustion stage in a test furnace at a firing rate of 0.5 to 1.5 MMBtu/H.

Heaton, H.; McHale, E.

1991-01-01T23:59:59.000Z

159

ADVANCED DIRECT LIQUEFACTION CONCEPTS FOR PETC GENERIC UNITS  

SciTech Connect (OSTI)

The results of Laboratory and Bench-Scale experiments and supporting technical and economic assessments conducted under DOE Contract No. DE-AC22-91PC91040 is reported for the period July 1, 1998 to September 30, 1998. This contract is with the University of kentucky Research Foundation, which supports work with the University of Kentucky Center for Applied Energy Researc, CONSOL, Inc., LDP Associates, and Hydrocarbon Technologies, Inc. This work involves the introduction into the basic two-stage liquefaction process several novel concepts, which include dispersed lower-cost catalysts, coal cleaning by oil agglomeration, and distillate hydrotreating and dewaxing. This project has been modified to include an investigation into the production of value added materials from coal using liquefaction based technologies.

Adam J. Berkovich

2000-02-01T23:59:59.000Z

160

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

SciTech Connect (OSTI)

The nuclear industry interest in advanced fuel and reactor design often drives towards fuel with uranium enrichments greater than 5 wt% 235U. Unfortunately, little data exists, in the form of reactor physics and criticality benchmarks, for uranium enrichments ranging between 5 and 10 wt% 235U. The primary purpose of this project is to provide benchmarks for fuel similar to what may be required for advanced light water reactors (LWRs). These experiments will ultimately provide additional information for application to the criticality-safety bases for commercial fuel facilities handling greater than 5 wt% 235U fuel.

William Anderson; James Tulenko; Bradley Rearden; Gary Harms

2008-09-11T23:59:59.000Z

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

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

E-Print Network [OSTI]

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

De Cindio, Fiorella

162

Uncertainty analysis of LBLOCA for Advanced Heavy Water Reactor  

Science Journals Connector (OSTI)

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

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

2008-01-01T23:59:59.000Z

163

Advanced Direct Liquefaction Concepts for PETC Generic Units - Phase II  

SciTech Connect (OSTI)

The results of Laboratory and Bench-Scale experiments and supporting technical and economic assessments conducted under DOE Contract No. DE-AC22-91PC91040 are reported for the period July 1, 1997 to September 30, 1997. This contract is with the University of Kentucky Research Foundation which supports work with the University of Kentucky Center for Applied Energy Research, CONSOL, Inc., LDP Associates, and Hydrocarbon Technologies, Inc. This work involves the introduction into the basic two stage liquefaction process several novel concepts which include dispersed lower-cost catalysts, coal cleaning by oil agglomeration, and distillate hydrotreating and dewaxing. Results are reported from experiments in which various methods were tested to activate dispersed Mo precursors. Several oxothiomolybdates precursors having S/Mo ratios from two to six were prepared. Another having a S/Mo ratio of eleven was also prepared that contained an excess of sulfur. In the catalyst screening test, none of these precursors exhibited an activity enhancement that might suggest that adding sulfur into the structure of the Mo precursors would be beneficial to the process. In another series of experiments, AHM impregnated coal slurried in the reaction mixture was pretreated withH S/H under pressure and successively heated for 30 min at 120, 250 2 2 and 360 C. THF conversions in the catalyst screening test were not affected while resid conversions o increased such that pretreated coals impregnated with 100 ppm Mo gave conversions equivalent to untreated coals impregnated with 300 ppm fresh Mo. Cobalt, nickel and potassium phosphomolybdates were prepared and tested as bimetallic precursors. The thermal stability of these compounds was evaluated in TG/MS to determine whether the presence of the added metal would stabilize the Keggin structure at reaction temperature. Coals impregnated with these salts showed the Ni and Co salts gave the same THF conversion as PMA while the Ni salt gave higher resid conversion than the other salts and untreated PMA. To activate PMA, a series of sulfided PMA materials was prepared by subjecting the crystalline acid to H S/H at 125-450 C. The chemistries 2 2 o of these partially sulfided materials are reported as well as the reactivity of several impregnated coals. None of the coals impregnated with these sulfided PMA materials gave conversions that exceeded PMA. Reports covering work by the subcontractors for this reporting period have not been received. A report from CONSOL covering a previous reporting period is included.

None

1997-12-01T23:59:59.000Z

164

Development and Evaluation of a Safeguards System Concept for a Pebble-Fueled High Temperature Gas-cooled Reactor  

E-Print Network [OSTI]

............................................................................................... 24 8 Flow diagram for an Advanced CANDU reactor ..................................... 27 9 Implementation of safeguards measures at a CANDU facility using video surveillance and radiation monitors... ................................................ 28 10 Implementation of safeguards measures at a CANDU facility using core discharge monitor.............................................................................. 28 11 Primary safeguards measures at MONJU Fast Reactor in Japan...

Gitau, Ernest Travis Ngure

2012-10-19T23:59:59.000Z

165

Vehicle Technologies Office Merit Review 2014: Advanced Combustion Concepts- Enabling Systems and Solutions (ACCESS) for High Efficiency Light Duty Vehicles  

Broader source: Energy.gov [DOE]

Presentation given by Robert Bosch at 2014 DOE Hydrogen and Fuel Cells Program and Vehicle Technologies Office Annual Merit Review and Peer Evaluation Meeting about advanced combustion concepts -...

166

The hybrid reactor project based on the straight field line mirror concept  

SciTech Connect (OSTI)

The straight field line mirror (SFLM) concept is aiming towards a steady-state compact fusion neutron source. Besides the possibility for steady state operation for a year or more, the geometry is chosen to avoid high loads on materials and plasma facing components. A comparatively small fusion hybrid device with 'semi-poor' plasma confinement (with a low fusion Q factor) may be developed for industrial transmutation and energy production from spent nuclear fuel. This opportunity arises from a large fission to fusion energy multiplication ratio, Q{sub r} = P{sub fis}/P{sub fus}>>1. The upper bound on Q{sub r} is primarily determined by geometry and reactor safety. For the SFLM, the upper bound is Q{sub r} Almost-Equal-To 150, corresponding to a neutron multiplicity of k{sub eff}=0.97. Power production in a mirror hybrid is predicted for a substantially lower electron temperature than the requirement T{sub e} Almost-Equal-To 10 keV for a fusion reactor. Power production in the SFLM seems possible with Q Almost-Equal-To 0.15, which is 10 times lower than typically anticipated for hybrids (and 100 times smaller than required for a fusion reactor). This relaxes plasma confinement demands, and broadens the range for use of plasmas with supra-thermal ions in hybrid reactors. The SFLM concept is based on a mirror machine stabilized by qudrupolar magnetic fields and large expander tanks beyond the confinement region. The purpose of the expander tanks is to distribute axial plasma loss flow over a sufficiently large area so that the receiving plates can withstand the heat. Plasma stability is not relying on a plasma flow into the expander regions. With a suppressed plasma flow into the expander tanks, a possibility arise for higher electron temperature. A brief presentation will be given on basic theory for the SFLM with plasma stability and electron temperature issues, RF heating computations with sloshing ion formation, neutron transport computations with reactor safety margins and material load estimates, magnetic coil designs as well as a discussion on the implications of the geometry for possible diagnostics. Reactor safety issues are addressed and a vertical orientation of the device could assist passive coolant circulation. Specific attention is put to a device with a 25 m long confinement region and 40 cm plasma radius in the mid-plane. In an optimal case (k{sub eff}= 0.97) with a fusion power of only 10 MW, such a device may be capable of producing a power of 1.5 GW{sub th}.

Agren, O.; Noack, K.; Moiseenko, V. E.; Hagnestal, A.; Kaellne, J.; Anglart, H. [Uppsala University, Angstroem Laboratory, Uppsala University, Box 534, SE-751 21 Uppsala (Sweden); Institute of Plasma Physics, National Science Center 'Kharkiv Institute of Physics and Technology', 61108 Kharkiv (Ukraine); Uppsala University, Angstroem Laboratory, Uppsala University, Box 534, SE-751 21 Uppsala (Sweden); Royal Institute of Technology, Nuclear Reactor Technology, SE 100 44 Stockholm (Sweden)

2012-06-19T23:59:59.000Z

167

Results from the DOE Advanced Gas Reactor Fuel Development and Qualification Program  

SciTech Connect (OSTI)

Modular HTGR designs were developed to provide natural safety, which prevents core damage under all design basis accidents and presently envisioned severe accidents. The principle that guides their design concepts is to passively maintain core temperatures below fission product release thresholds under all accident scenarios. This level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude and allows potential elimination of the need for evacuation and sheltering beyond a small exclusion area. This level, however, is predicated on exceptionally high fuel fabrication quality and performance under normal operation and accident conditions. Germany produced and demonstrated high quality fuel for their pebble bed HTGRs in the 1980s, but no U.S. manufactured fuel had exhibited equivalent performance prior to the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The design goal of the modular HTGRs is to allow elimination of an exclusion zone and an emergency planning zone outside the plant boundary fence, typically interpreted as being about 400 meters from the reactor. To achieve this, the reactor design concepts require a level of fuel integrity that is better than that claimed for all prior US manufactured TRISO fuel, by a few orders of magnitude. The improved performance level is about a factor of three better than qualified for German TRISO fuel in the 1980s. At the start of the AGR program, without a reactor design concept selected, the AGR fuel program selected to qualify fuel to an operating envelope that would bound both pebble bed and prismatic options. This resulted in needing a fuel form that could survive at peak fuel temperatures of 1250C on a time-averaged basis and high burnups in the range of 150 to 200 GWd/MTHM (metric tons of heavy metal) or 16.4 to 21.8% fissions per initial metal atom (FIMA). Although Germany has demonstrated excellent performance of TRISO-coated UO2 particle fuel up to about 10% FIMA and 1150C, UO2 fuel is known to have limitations because of CO formation and kernel migration at the high burnups, power densities, temperatures, and temperature gradients that may be encountered in the prismatic modular HTGRs. With uranium oxycarbide (UCO) fuel, the kernel composition is engineered to prevent CO formation and kernel migration, which are key threats to fuel integrity at higher burnups, temperatures, and temperature gradients. Furthermore, the recent poor fuel performance of UO2 TRISO fuel pebbles measured in Chinese irradiation testing in Russia and in German pebbles irradiated at 1250C, and historic data on poorer fuel performance in safety testing of German pebbles that experienced burnups in excess of 10% FIMA [1] have each raised concern about the use of UO2 TRISO above 10% FIMA and 1150C and the degree of margin available in the fuel system. This continues to be an active area of study internationally.

David Petti

2014-06-01T23:59:59.000Z

168

ADVANCED REACTOR SAFETY PROGRAM STAKEHOLDER INTERACTION AND FEEDBACK  

SciTech Connect (OSTI)

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

Benjamin W. Spencer; Hai Huang

2014-08-01T23:59:59.000Z

169

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

SciTech Connect (OSTI)

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

Noel Duckwitz

2011-06-01T23:59:59.000Z

170

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

SciTech Connect (OSTI)

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

Noel Duckwitz

2011-06-01T23:59:59.000Z

171

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

SciTech Connect (OSTI)

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

Noel Duckwitz

2011-06-01T23:59:59.000Z

172

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

SciTech Connect (OSTI)

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

Not Available

1992-04-01T23:59:59.000Z

173

Proceedings of the 1992 topical meeting on advances in reactor physics  

SciTech Connect (OSTI)

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

Not Available

1992-01-01T23:59:59.000Z

174

Piping benchmark problems for the General Electric Advanced Boiling Water Reactor  

SciTech Connect (OSTI)

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

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

1993-08-01T23:59:59.000Z

175

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

SciTech Connect (OSTI)

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

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

1994-08-01T23:59:59.000Z

176

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

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

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

177

A feasibility study of reactor-based deep-burn concepts.  

SciTech Connect (OSTI)

A systematic assessment of the General Atomics (GA) proposed Deep-Burn concept based on the Modular Helium-Cooled Reactor design (DB-MHR) has been performed. Preliminary benchmarking of deterministic physics codes was done by comparing code results to those from MONTEBURNS (MCNP-ORIGEN) calculations. Detailed fuel cycle analyses were performed in order to provide an independent evaluation of the physics and transmutation performance of the one-pass and two-pass concepts. Key performance parameters such as transuranic consumption, reactor performance, and spent fuel characteristics were analyzed. This effort has been undertaken in close collaborations with the General Atomics design team and Brookhaven National Laboratory evaluation team. The study was performed primarily for a 600 MWt reference DB-MHR design having a power density of 4.7 MW/m{sup 3}. Based on parametric and sensitivity study, it was determined that the maximum burnup (TRU consumption) can be obtained using optimum values of 200 {micro}m and 20% for the fuel kernel diameter and fuel packing fraction, respectively. These values were retained for most of the one-pass and two-pass design calculations; variation to the packing fraction was necessary for the second stage of the two-pass concept. Using a four-batch fuel management scheme for the one-pass DB-MHR core, it was possible to obtain a TRU consumption of 58% and a cycle length of 286 EFPD. By increasing the core power to 800 MWt and the power density to 6.2 MW/m{sup 3}, it was possible to increase the TRU consumption to 60%, although the cycle length decreased by {approx}64 days. The higher TRU consumption (burnup) is due to the reduction of the in-core decay of fissile Pu-241 to Am-241 relative to fission, arising from the higher power density (specific power), which made the fuel more reactivity over time. It was also found that the TRU consumption can be improved by utilizing axial fuel shuffling or by operating with lower material temperatures (colder core). Results also showed that the transmutation performance of the one-pass deep-burn concept is sensitive to the initial TRU vector, primarily because longer cooling time reduces the fissile content (Pu-241 specifically.) With a cooling time of 5 years, the TRU consumption increases to 67%, while conversely, with 20-year cooling the TRU consumption is about 58%. For the two-pass DB-MHR (TRU recycling option), a fuel packing fraction of about 30% is required in the second pass (the recycled TRU). It was found that using a heterogeneous core (homogeneous fuel element) concept, the TRU consumption is dependent on the cooling interval before the 2nd pass, again due to Pu-241 decay during the time lag between the first pass fuel discharge and the second pass fuel charge. With a cooling interval of 7 years (5 and 2 years before and after reprocessing) a TRU consumption of 55% is obtained. With an assumed ''no cooling'' interval, the TRU consumption is 63%. By using a cylindrical core to reduce neutron leakage, TRU consumption of the case with 7-year cooling interval increases to 58%. For a two-pass concept using a heterogeneous fuel element (and homogeneous core) with first and second pass volume ratio of 2:1, the TRU consumption is 62.4%. Finally, the repository loading benefits arising from the deep-burn and Inert Matrix Fuel (IMF) concepts were estimated and compared, for the same initial TRU vector. The DB-MHR concept resulted in slightly higher TRU consumption and repository loading benefit compared to the IMF concept (58.1% versus 55.1% for TRU consumption and 2.0 versus 1.6 for estimated repository loading benefit).

Kim, T. K.; Taiwo, T. A.; Hill, R. N.; Yang, W. S.

2005-09-16T23:59:59.000Z

178

Safeguards and security concept for the Secure Automated Fabrication (SAF) and Liquid Metal Reactor (LMR) fuel cycle, SAF line technical support  

SciTech Connect (OSTI)

This report is a safeguards and security concept system review for the secure automated fabrication (SAF) and national liquid metal reactor (LMR) fuel programs.

Schaubert, V.J.; Remley, M.E.; Grantham, L.F.

1986-02-21T23:59:59.000Z

179

Advanced nuclear reactors and tritium impacts. Modeling the aquatic pathway  

Science Journals Connector (OSTI)

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

Francisco Fernando Lamego Simes Filho; Abner Duarte Soares; Andr da Silva Aguiar; Celso Marcelo Franklin Lapa; Antonio Carlos Ferreira Guimares

2013-01-01T23:59:59.000Z

180

Statistical Methods Handbook for Advanced Gas Reactor Fuel Materials  

SciTech Connect (OSTI)

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

J. J. Einerson

2005-05-01T23:59:59.000Z

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

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

SciTech Connect (OSTI)

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

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

2012-07-01T23:59:59.000Z

182

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

SciTech Connect (OSTI)

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

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

1993-03-01T23:59:59.000Z

183

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

SciTech Connect (OSTI)

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

Mike Lewis

2014-02-01T23:59:59.000Z

184

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

Science Journals Connector (OSTI)

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

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

2006-01-01T23:59:59.000Z

185

E-Print Network 3.0 - assisted reactor concept Sample Search...  

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

Technology Through Flexibility: The Case of a Demonstration Accelerator-Driven Subcritical Reactor Park EPRG... for Accelerator- Driven Subcritical Reactors (ADSRs). It shows...

186

E-Print Network 3.0 - army gas-cooled reactor Sample Search Results  

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

ENABLING SUSTAINABLE NUCLEAR POWER Summary: and NRE Design Class., "Advances in the Subcritical, Gas-Cooled Fast Transmutation Reactor Concept", Nucl... . Tedder, J. Lackey, J....

187

Testing of an advanced thermochemical conversion reactor system  

SciTech Connect (OSTI)

This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective of this project was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions. 6 refs., 59 figs., 29 tabs.

Not Available

1990-01-01T23:59:59.000Z

188

Secondary heat exchanger design and comparison for advanced high temperature reactor  

SciTech Connect (OSTI)

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

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

2012-07-01T23:59:59.000Z

189

Advanced Scintillator Detector Concept (ASDC): A Concept Paper on the Physics Potential of Water-Based Liquid Scintillator  

E-Print Network [OSTI]

The recent development of Water-based Liquid Scintillator (WbLS), and the concurrent development of high-efficiency and high-precision-timing light sensors, has opened up the possibility for a new kind of large-scale detector capable of a very broad program of physics. The program would include determination of the neutrino mass hierarchy and observation of CP violation with long-baseline neutrinos, searches for proton decay, ultra-precise solar neutrino measurements, geo- and supernova neutrinos including diff?use supernova antineutrinos, and neutrinoless double beta decay. We outline here the basic requirements of the Advanced Scintillation Detector Concept (ASDC), which combines the use of WbLS, doping with a number of potential isotopes for a range of physics goals, high efficiency and ultra-fast timing photosensors, and a deep underground location. We are considering such a detector at the Long Baseline Neutrino Facility (LBNF) far site, where the ASDC could operate in conjunction with the liquid argon t...

Alonso, J R; Bergevin, M; Bernstein, A; Bignell, L; Blucher, E; Calaprice, F; Conrad, J M; Descamps, F B; Diwan, M V; Dwyer, D A; Dye, S T; Elagin, A; Feng, P; Grant, C; Grullon, S; Hans, S; Jaffe, D E; Kettell, S H; Klein, J R; Lande, K; Learned, J G; Luk, K B; Maricic, J; Marleau, P; Mastbaum, A; McDonough, W F; Oberauer, L; Gann, G D Orebi; Rosero, R; Rountree, S D; Sanchez, M C; Shaevitz, M H; Shokair, T M; Smy, M B; Strait, M; Svoboda, R; Tolich, N; Vagins, M R; van Bibber, K A; Viren, B; Vogelaar, R B; Wetstein, M J; Winslow, L; Wonsak, B; Worcester, E T; Wurm, M; Yeh, M; Zhang, C

2014-01-01T23:59:59.000Z

190

Advanced Scintillator Detector Concept (ASDC): A Concept Paper on the Physics Potential of Water-Based Liquid Scintillator  

E-Print Network [OSTI]

The recent development of Water-based Liquid Scintillator (WbLS), and the concurrent development of high-efficiency and high-precision-timing light sensors, has opened up the possibility for a new kind of large-scale detector capable of a very broad program of physics. The program would include determination of the neutrino mass hierarchy and observation of CP violation with long-baseline neutrinos, searches for proton decay, ultra-precise solar neutrino measurements, geo- and supernova neutrinos including diffuse supernova antineutrinos, and neutrinoless double beta decay. We outline here the basic requirements of the Advanced Scintillation Detector Concept (ASDC), which combines the use of WbLS, doping with a number of potential isotopes for a range of physics goals, high efficiency and ultra-fast timing photosensors, and a deep underground location. We are considering such a detector at the Long Baseline Neutrino Facility (LBNF) far site, where the ASDC could operate in conjunction with the liquid argon tracking detector proposed by the LBNE collaboration. The goal is the deployment of a 30-100 kiloton-scale detector, the basic elements of which are being developed now in experiments such as WATCHMAN, ANNIE, SNO+, and EGADS.

J. R. Alonso; N. Barros; M. Bergevin; A. Bernstein; L. Bignell; E. Blucher; F. Calaprice; J. M. Conrad; F. B. Descamps; M. V. Diwan; D. A. Dwyer; S. T. Dye; A. Elagin; P. Feng; C. Grant; S. Grullon; S. Hans; D. E. Jaffe; S. H. Kettell; J. R. Klein; K. Lande; J. G. Learned; K. B. Luk; J. Maricic; P. Marleau; A. Mastbaum; W. F. McDonough; L. Oberauer; G. D. Orebi Gann; R. Rosero; S. D. Rountree; M. C. Sanchez; M. H. Shaevitz; T. M. Shokair; M. B. Smy; A. Stahl; M. Strait; R. Svoboda; N. Tolich; M. R. Vagins; K. A. van Bibber; B. Viren; R. B. Vogelaar; M. J. Wetstein; L. Winslow; B. Wonsak; E. T. Worcester; M. Wurm; M. Yeh; C. Zhang

2014-10-24T23:59:59.000Z

191

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

SciTech Connect (OSTI)

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

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

2014-09-01T23:59:59.000Z

192

Roadmap for development of an advanced head-end reactor  

SciTech Connect (OSTI)

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

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

2013-07-01T23:59:59.000Z

193

Development of Advanced Concept for Shortening Construction Period of ABWR Plant  

SciTech Connect (OSTI)

Construction of a nuclear power plant (NPP) requires a very long period because of large amount of construction materials and many issues for negotiation among multiple sections. Shortening the construction period advances the date of return on an investment, and can also result in reduced construction cost. Therefore, the study of this subject has a very high priority for utilities. We achieved a construction period of 37 months from the first concrete work to fuel loading (F/L) (51.5 months from the inspection of the foundation (I/F) to the start of commercial operation (C/O)) at the Kashiwazaki-Kariwa NPPs No. 6 and 7 (KK-6/7), which are the first ABWR plants in the world. At TEPCO's next plant, we think that a construction period of less than 36 months (45 months from I/F to C/O) can be realized based on conventional methods such as early start of equipment installation and blocking of equipment to be brought in advance. Furthermore, we are studying the feasibility of a 21.5-month construction period (30 months from I/F to C/O) with advanced ideas and methods. The important concepts for a 21.5-month construction period are adoption of a new building structure that is the steel plate reinforced concrete (SC) structure and promotion of extensive modularization of equipment and building structure. With introducing these new concepts, we are planning the master schedule (M/S) and finding solutions to conflicts in the schedule of area release from building construction work to equipment installation work (schedule-conflicts.) In this report, we present the shortest construction period and an effective method to put it into practice for the conventional general arrangement (GA) of ABWR. In the future, we will continue the study on the improvement of building configuration and arrangements, and make clear of the concept for large composite modules of building structures and equipment. (authors)

Hiroshi Ijichi; Toshio Yamashita; Masahiro Tsutagawa; Hiroya Mori [Toshiba Corporation (Japan); Nobuaki Ooshima; Jun Miura [Hitachi Ltd. (Japan); Minoru Kanechika [Kajima Corporation (Japan); Nobuaki Miura [Shimizu Corporation (Japan)

2002-07-01T23:59:59.000Z

194

Advanced direct liquefaction concepts for PETC generic units. Final report, Phase I  

SciTech Connect (OSTI)

The Advanced Concepts for Direct Coal Liquefaction program was initiated by the Department of Energy in 1991 to develop technologies that could significantly reduce the cost of producing liquid fuels by the direct liquefaction of coal. The advanced 2-stage liquefaction technology that was developed at Wilsonville over the past 10 years has contributed significantly toward decreasing the cost of producing liquids from coal to about $33/bbl. It remains, however, the objective of DOE to further reduce this cost to a level more competitive with petroleum based products. This project, among others, was initiated to investigate various alternative approaches to develop technologies that might ultimately lead to a 25 % reduction in cost of product. In this project a number of novel concepts were investigated, either individually or in a coupled configuration that had the potential to contribute toward meeting the DOE goal. The concepts included mature technologies or ones closely related to them, such as coal cleaning by oil agglomeration, fluid coking and distillate hydrotreating and dewaxing. Other approaches that were either embryonic or less developed were chemical pretreatment of coal to remove oxygen, and dispersed catalyst development for application in the 2-stage liquefaction process. This report presents the results of this project. It is arranged in four sections which were prepared by participating organizations responsible for that phase of the project. A summary of the overall project and the principal results are given in this section. First, however, an overview of the process economics and the process concepts that were developed during the course of this program is presented.

NONE

1995-03-01T23:59:59.000Z

195

Advanced High-Temperature, High-Pressure Transport Reactor Gasification  

SciTech Connect (OSTI)

The transport reactor development unit (TRDU) was modified to accommodate oxygen-blown operation in support of a Vision 21-type energy plex that could produce power, chemicals, and fuel. These modifications consisted of changing the loop seal design from a J-leg to an L-valve configuration, thereby increasing the mixing zone length and residence time. In addition, the standpipe, dipleg, and L-valve diameters were increased to reduce slugging caused by bubble formation in the lightly fluidized sections of the solid return legs. A seal pot was added to the bottom of the dipleg so that the level of solids in the standpipe could be operated independently of the dipleg return leg. A separate coal feed nozzle was added that could inject the coal upward into the outlet of the mixing zone, thereby precluding any chance of the fresh coal feed back-mixing into the oxidizing zone of the mixing zone; however, difficulties with this coal feed configuration led to a switch back to the original downward configuration. Instrumentation to measure and control the flow of oxygen and steam to the burner and mix zone ports was added to allow the TRDU to be operated under full oxygen-blown conditions. In total, ten test campaigns have been conducted under enriched-air or full oxygen-blown conditions. During these tests, 1515 hours of coal feed with 660 hours of air-blown gasification and 720 hours of enriched-air or oxygen-blown coal gasification were completed under this particular contract. During these tests, approximately 366 hours of operation with Wyodak, 123 hours with Navajo sub-bituminous coal, 143 hours with Illinois No. 6, 106 hours with SUFCo, 110 hours with Prater Creek, 48 hours with Calumet, and 134 hours with a Pittsburgh No. 8 bituminous coal were completed. In addition, 331 hours of operation on low-rank coals such as North Dakota lignite, Australian brown coal, and a 90:10 wt% mixture of lignite and wood waste were completed. Also included in these test campaigns was 50 hours of gasification on a petroleum coke from the Hunt Oil Refinery and an additional 73 hours of operation on a high-ash coal from India. Data from these tests indicate that while acceptable fuel gas heating value was achieved with these fuels, the transport gasifier performs better on the lower-rank feedstocks because of their higher char reactivity. Comparable carbon conversions have been achieved at similar oxygen/coal ratios for both air-blown and oxygen-blown operation for each fuel; however, carbon conversion was lower for the less reactive feedstocks. While separation of fines from the feed coals is not needed with this technology, some testing has suggested that feedstocks with higher levels of fines have resulted in reduced carbon conversion, presumably due to the inability of the finer carbon particles to be captured by the cyclones. These data show that these low-rank feedstocks provided similar fuel gas heating values; however, even among the high-reactivity low-rank coals, the carbon conversion did appear to be lower for the fuels (brown coal in particular) that contained a significant amount of fines. The fuel gas under oxygen-blown operation has been higher in hydrogen and carbon dioxide concentration since the higher steam injection rate promotes the water-gas shift reaction to produce more CO{sub 2} and H{sub 2} at the expense of the CO and water vapor. However, the high water and CO{sub 2} partial pressures have also significantly reduced the reaction of (Abstract truncated)

Michael L. Swanson

2005-08-30T23:59:59.000Z

196

In-Situ Creep Testing Capability for the Advanced Test Reactor  

SciTech Connect (OSTI)

An instrumented creep testing capability is being developed for specimens irradiated in Pressurized Water Reactor (PWR) coolant conditions at the Advanced Test Reactor (ATR). The test rig has been developed such that samples will be subjected to stresses ranging from 92 to 350 MPa at temperatures between 290 and 370 C up to at least 2 dpa (displacement per atom). The status of Idaho National Laboratory (INL) efforts to develop the test rig in-situ creep testing capability for the ATR is described. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper reports efforts by INL to evaluate a prototype test rig in an autoclave at INLs High Temperature Test Laboratory (HTTL). Initial data from autoclave tests with 304 stainless steel (304 SS) specimens are reported.

B. G. Kim; J. L. Rempe; D. L. Knudson; K. G. Condie; B. H. Sencer

2012-09-01T23:59:59.000Z

197

TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis  

SciTech Connect (OSTI)

The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided.

Liles, D.R.; Mahaffy, J.H.

1984-02-01T23:59:59.000Z

198

Design and Status of RERTR Irradiation Tests in the Advanced Test Reactor  

SciTech Connect (OSTI)

Irradiation testing of U-Mo based fuels is the central component of the Reduced Enrichment for Research and Test Reactors (RERTR) program fuel qualification plan. Several RERTR tests have recently been completed or are planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory in Idaho Falls, ID. Four mini-plate experiments in various stages of completion are described in detail, including the irradiation test design, objectives, and irradiation conditions. Observations made during and after the in-reactor RERTR-7A experiment breach are summarized. The irradiation experiment design and planned irradiation conditions for full-size plate test are described. Progress toward element testing will be reviewed.

Daniel M. Wachs; Richard G. Ambrosek; Gray Chang; Mitchell K. Meyer

2006-10-01T23:59:59.000Z

199

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

Science Journals Connector (OSTI)

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

Aly Karameldin; Samer S. Mekhemar

2004-01-01T23:59:59.000Z

200

Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs  

SciTech Connect (OSTI)

This report documents the work performed during the first phase of the National Aeronautics and Space Administration (NASA), National Research Announcement (NRA) Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs. The document includes an optimization of both 100-kW{sub e} and 250-kW{sub e} (at the propulsion unit) Rankine cycle power conversion systems. In order to perform the mass optimization of these systems, several parametric evaluations of different design options were investigated. These options included feed and reheat, vapor superheat levels entering the turbine, three different material types, and multiple heat rejection system designs. The overall masses of these Nb-1%Zr systems are approximately 3100 kg and 6300 kg for the 100- kW{sub e} and 250-kW{sub e} systems, respectively, each with two totally redundant power conversion units, including the mass of the single reactor and shield. Initial conceptual designs for each of the components were developed in order to estimate component masses. In addition, an overall system concept was presented that was designed to fit within the launch envelope of a heavy lift vehicle. A technology development plan is presented in the report that describes the major efforts that are required to reach a technology readiness level of 6. A 10-year development plan was proposed.

Yoder, G.L.

2005-10-03T23:59:59.000Z

Note: This page contains sample records for the topic "advanced reactor concepts" from the National Library of EnergyBeta (NLEBeta).
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201

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

E-Print Network [OSTI]

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

Tsvetkova, Galina Valeryevna

2004-09-30T23:59:59.000Z

202

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

Science Journals Connector (OSTI)

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

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

2014-01-01T23:59:59.000Z

203

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

SciTech Connect (OSTI)

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

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

2011-05-31T23:59:59.000Z

204

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

SciTech Connect (OSTI)

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

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

2012-08-01T23:59:59.000Z

205

Axial effects of xenon-samarium poisoning in the advanced test reactor  

SciTech Connect (OSTI)

The paper details an analytical study of the time-dependent behavior in the spatial distributions of xenon and samarium fission product poisons in the Advanced Test Reactor (ATR) during operation and after shutdown. The results of this study provide insight into the behavior and significance of the changing spatial distributions of fission product poisons with respect to the prediction of shim positions at critical for reactor restart after a xenon shutdown. The study was performed with the PDQ neutron diffusion theory code and ENDF/B-V cross sections using a one-dimensional radial model of an ATR lobe and a two-dimensional radial-axial (RZ) model of an ATR lobe. The PDQ results were supported by a review of the basic differential equations, which describe the buildup and decay of the xenon and samarium fission product poisons and precursors. The ATR is a 250-MW, uranium-aluminum-fueled reactor used to study the effects of irradiation on reactor materials. Forty highly enriched uranium fuel elements are arranged in a serpentine configuration within the compact core resulting in a very high power density of (1.0 MW/[ell] of core).

Auslander, D.J.; Smith, A.C.; McCracken, R.T. (Idaho National Engineering Lab., Idaho Falls (United States))

1990-01-01T23:59:59.000Z

206

Advanced Circulating Pressurized Fluidized Bed Combustion (APFBC) Repowering Concept Assessment at Duke Energy's Dan River Station  

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

Wolfmeyer et al. APFBC Repowering Assessment at Duke Energy's Dan River Station Wolfmeyer et al. APFBC Repowering Assessment at Duke Energy's Dan River Station paper 970561 Page 1 of 36 Advanced Circulating Pressurized Fluidized Bed Combustion (APFBC) Repowering Concept Assessment at Duke Energy's Dan River Station John C. Wolfmeyer, P.E., and Cal Jowers, P.E. Duke Energy / Charlotte, North Carolina Richard E. Weinstein, P.E., Harvey N. Goldstein, P.E., and Jay S. White Parsons Power Group Inc. / Reading, Pennsylvania Robert W. Travers, P.E. U.S. Department of Energy Office of Fossil Energy / Germantown, Maryland electronic mail addresses/phone no. electronic mail addresses/phone no. Wolfmeyer { JCWolfme@Duke-Energy.COM 704 / 382-4017 Goldstein { Harvey_N_Goldstein@Parsons.COM 610 / 855-3281 Jowers { -- 704 / 382-9577 White { Jay_S_White@Parsons.COM

207

AICD -- Advanced Industrial Concepts Division Biological and Chemical Technologies Research Program. 1993 Annual summary report  

SciTech Connect (OSTI)

The annual summary report presents the fiscal year (FY) 1993 research activities and accomplishments for the United States Department of Energy (DOE) Biological and Chemical Technologies Research (BCTR) Program of the Advanced Industrial Concepts Division (AICD). This AICD program resides within the Office of Industrial Technologies (OIT) of the Office of Energy Efficiency and Renewable Energy (EE). The annual summary report for 1993 (ASR 93) contains the following: A program description (including BCTR program mission statement, historical background, relevance, goals and objectives), program structure and organization, selected technical and programmatic highlights for 1993, detailed descriptions of individual projects, a listing of program output, including a bibliography of published work, patents, and awards arising from work supported by BCTR.

Petersen, G.; Bair, K.; Ross, J. [eds.

1994-03-01T23:59:59.000Z

208

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

SciTech Connect (OSTI)

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

Gantt, D.A.

1998-09-17T23:59:59.000Z

209

Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment  

SciTech Connect (OSTI)

The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/Bs) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

Dawn M. Scates; John (Jack) K Hartwell; John B. Walter

2008-09-01T23:59:59.000Z

210

Computed phase equilibria for burnable neutron absorbing materials for advanced pressurized heavy water reactors  

Science Journals Connector (OSTI)

Burnable neutron absorbing materials are expected to be an integral part of the new fuel design for the Advanced CANDU[CANDU is as a registered trademark of Atomic Energy of Canada Limited.] Reactor. The neutron absorbing material is composed of gadolinia and dysprosia dissolved in an inert cubic-fluorite yttria-stabilized zirconia matrix. A thermodynamic model based on Gibbs energy minimization has been created to provide estimated phase equilibria as a function of composition and temperature. This work includes some supporting experimental studies involving X-ray diffraction.

E.C. Corcoran; B.J. Lewis; W.T. Thompson; J. Hood; F. Akbari; Z. He; P. Reid

2009-01-01T23:59:59.000Z

211

ORIGEN-ARP Cross-Section Libraries for Magnox, Advanced Gas-Cooled, and VVER Reactor Designs  

SciTech Connect (OSTI)

Cross-section libraries for the ORIGEN-ARP system were extended to include four non-U.S. reactor types: the Magnox reactor, the Advanced Gas-Cooled Reactor, the VVER-440, and the VVER-1000. Typical design and operational parameters for these four reactor types were determined by an examination of a variety of published information sources. Burnup simulation models of the reactors were then developed using the SAS2H sequence from the Oak Ridge National Laboratory SCALE code system. In turn, these models were used to prepare the burnup-dependent cross-section libraries suitable for use with ORIGEN-ARP. The reactor designs together with the development of the SAS2H models are described, and a small number of validation results using spent-fuel assay data are reported.

Murphy, BD

2004-03-10T23:59:59.000Z

212

Recovery of Carbon Dioxide in Advanced Fossil Energy Conversion Processes Using a Membrane Reactor  

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

Carbon Dioxide in Advanced Fossil Energy Conversion Processes Carbon Dioxide in Advanced Fossil Energy Conversion Processes Using a Membrane Reactor Ashok S. Damle * Research Triangle Institute P.O. Box 12194 Research Triangle Park, NC 27709 Phone: (919) 541-6146 Fax: (919) 541-6965 E-mail: adamle@rti.org Thomas P. Dorchak National Energy Technology Laboratory P.O. Box 880, Mail Stop C04 Morgantown, WV 26507-0880 Phone: (304) 285-4305 E-mail: tdorch@netl.doe.gov Abstract Increased awareness of the global warming trend has led to worldwide concerns regarding "greenhouse gas" emissions, with CO 2 being the single greatest contributor to global warming. Fossil fuels (i.e., coal, oil, and natural gas) currently supply over 85% of the world's energy needs, and their utilization is the major source of the anthropogenic greenhouse gas emissions of

213

Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Technical Exchange Meeting  

SciTech Connect (OSTI)

During FY13, the INL developed an advanced SMR PRA framework which has been described in the report Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Detailed Technical Framework Specification, INL/EXT-13-28974 (April 2013). In this framework, the various areas are considered: Probabilistic models to provide information specific to advanced SMRs Representation of specific SMR design issues such as having co-located modules and passive safety features Use of modern open-source and readily available analysis methods Internal and external events resulting in impacts to safety All-hazards considerations Methods to support the identification of design vulnerabilities Mechanistic and probabilistic data needs to support modeling and tools In order to describe this framework more fully and obtain feedback on the proposed approaches, the INL hosted a technical exchange meeting during August 2013. This report describes the outcomes of that meeting.

Curtis Smith

2013-09-01T23:59:59.000Z

214

Spacetime kinetics modeling of Advanced Heavy Water Reactor for control studies  

Science Journals Connector (OSTI)

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

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

2010-01-01T23:59:59.000Z

215

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

Science Journals Connector (OSTI)

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

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

2015-01-01T23:59:59.000Z

216

Waste Heat Recovery from the Advanced Test Reactor Secondary Coolant Loop  

SciTech Connect (OSTI)

This study investigated the feasibility of using a waste heat recovery system (WHRS) to recover heat from the Advanced Test Reactor (ATR) secondary coolant system (SCS). This heat would be used to preheat air for space heating of the reactor building, thus reducing energy consumption, carbon footprint, and energy costs. Currently, the waste heat from the reactor is rejected to the atmosphere via a four-cell, induced-draft cooling tower. Potential energy and cost savings are 929 kW and $285K/yr. The WHRS would extract a tertiary coolant stream from the SCS loop and pump it to a new plate and frame heat exchanger, from which the heat would be transferred to a glycol loop for preheating outdoor air supplied to the heating and ventilation system. The use of glycol was proposed to avoid the freezing issues that plagued and ultimately caused the failure of a WHRS installed at the ATR in the 1980s. This study assessed the potential installation of a new WHRS for technical, logistical, and economic feasibility.

Donna Post Guillen

2012-11-01T23:59:59.000Z

217

Numerical Study on Crossflow Printed Circuit Heat Exchanger for Advanced Small Modular Reactors  

SciTech Connect (OSTI)

Various fluids such as water, gases (helium), molten salts (FLiNaK, FLiBe) and liquid metal (sodium) are used as a coolant of advanced small modular reactors (SMRs). The printed circuit heat exchanger (PCHE) has been adopted as the intermediate and/or secondary heat exchanger of SMR systems because this heat exchanger is compact and effective. The size and cost of PCHE can be changed by the coolant type of each SMR. In this study, the crossflow PCHE analysis code for advanced small modular reactor has been developed for the thermal design and cost estimation of the heat exchanger. The analytical solution of single pass, both unmixed fluids crossflow heat exchanger model was employed to calculate a two dimensional temperature profile of a crossflow PCHE. The analytical solution of crossflow heat exchanger was simply implemented by using built in function of the MATLAB program. The effect of fluid property uncertainty on the calculation results was evaluated. In addition, the effect of heat transfer correlations on the calculated temperature profile was analyzed by taking into account possible combinations of primary and secondary coolants in the SMR systems. Size and cost of heat exchanger were evaluated for the given temperature requirement of each SMR.

Su-Jong Yoon [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Piyush Sabharwall [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Eung-Soo Kim [Seoul National Univ., Seoul (Korea, Republic of)

2014-03-01T23:59:59.000Z

218

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

SciTech Connect (OSTI)

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

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

2014-01-01T23:59:59.000Z

219

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

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

fission reactors, which release energy by splitting atoms... ) International Thermonuclear Experimental Reactor (ITER), which will be ... Source: Fusiongnition Research...

220

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

SciTech Connect (OSTI)

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

David Petti; Hans Gougar; Gary Bell

2005-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "advanced reactor concepts" from the National Library of EnergyBeta (NLEBeta).
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221

Introduction to DMFCs - Advanced Materials and Concepts for Portable Power Fuel Cells  

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

Technologies Program Webinar Technologies Program Webinar July 17, 2012 1 Introduction to DMFCs Advanced Materials and Concepts for Portable Power Fuel Cells Piotr Zelenay Los Alamos National Laboratory, Los Alamos, New Mexico 87545, U.S.A. Fuel Cell Technologies Program Webinar - July 17, 2012 - The Fuel Choice P. Piela and P. Zelenay, Fuel Cell Review, 1, 17, 2004 Fuel Cell Technologies Program Webinar - July 17, 2012 2 Direct Methanol Fuel Cell Anode: Pt-Ru Cathode: Pt Membrane: e.g. Nafion ® 115 e - CH 3 OH H + H 2 O CH 3 OH Electroosmotic drag MEMBRANE 1.5 O 2 (air) H 2 O CO 2 + 3 H 2 O 6 H + + 6 e - ANODE CATHODE CH 3 OH (l) + 1.5 O 2  2 H 2 O (l) + CO 2  V = 1.21 V; G° = 6.1 kWh kg -1 = 4.8 kWh L -1 Fuel Cell Technologies Program Webinar - July 17, 2012 3 ______________________ O 2 H 

222

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

SciTech Connect (OSTI)

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

Not Available

1992-08-01T23:59:59.000Z

223

2012_AdvReactor_Factsheet.indd  

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

nuclear.gov nuclear.gov February 15, 2011 A ADVANCED REACTOR CONCEPTS DVANCED REACTOR CONCEPTS The U.S. Department of Energy's Offi ce of Nuclear Energy T he Advanced Reactor Concepts (ARC) program, an expanded version of the Generation IV research, development and demonstration (RD&D) program, sponsors research, development and deployment activities leading to further safety, technical, economical, and environmental advancements of innovative nuclear energy technologies. The Office of Nuclear Energy (NE) will pursue these advancements through RD&D activities at the Department of Energy (DOE) national laboratories and U.S. universities, as well as through collaboration with nuclear industry and international partners. These activities will focus on advancing

224

U.S. Department Of Energy Advanced Small Modular Reactor R&D Program: Instrumentation, Controls, and Human-Machine Interface (ICHMI) Pathway  

SciTech Connect (OSTI)

Instrumentation, controls, and human-machine interfaces (ICHMI) are essential enabling technologies that strongly influence nuclear power plant performance and operational costs. The nuclear power industry is currently engaged in a transition from traditional analog-based instrumentation, controls, and human-machine interface systems to implementations employing digital technologies. This transition has primarily occurred in an ad hoc fashion through individual system upgrades at existing plants and has been constrained by licenseability concerns. Although the recent progress in constructing new plants has spurred design of more fully digital plant-wide ICHMI systems, the experience base in the nuclear power application domain is limited. Additionally, development of advanced reactor concepts, such as Generation IV designs and small modular reactors, introduces different plant conditions (e.g., higher temperatures, different coolants, etc.) and unique plant configurations (e.g., multiunit plants with shared systems, balance of plant architectures with reconfigurable co-generation options) that increase the need for enhanced ICHMI capabilities to fully achieve industry goals related to economic competitiveness, safety and reliability, sustainability, and proliferation resistance and physical protection. As a result, significant challenges remain to be addressed to enable the nuclear power industry to complete the transition to safe and comprehensive use of modern ICHMI technology. The U.S. Department of Energy (DOE) has recognized that ICHMI research, development, and demonstration (RD&D) is needed to resolve the technical challenges that may compromise the effective and efficient utilization of modern ICHMI technology and consequently inhibit realization of the benefits offered by expanded utilization of nuclear power. Consequently, several DOE programs have substantial ICHMI RD&D elements within their respective research portfolios. This paper describes current ICHMI research in support of advanced small modular reactors. The objectives that can be achieved through execution of the defined RD&D are to provide optimal technical solutions to critical ICHMI issues, resolve technology gaps arising from the unique measurement and control characteristics of advanced reactor concepts, provide demonstration of needed technologies and methodologies in the nuclear power application domain, mature emerging technologies to facilitate commercialization, and establish necessary technical evidence and application experience to enable timely and predictable licensing. 1 Introduction Instrumentation, controls, and human-machine interfaces are essential enabling technologies that strongly influence nuclear power plant performance and operational costs. The nuclear power industry is currently engaged in a transition from traditional analog-based instrumentation, controls, and human-machine interface (ICHMI) systems to implementations employing digital technologies. This transition has primarily occurred in an ad hoc fashion through individual system upgrades at existing plants and has been constrained by licenseability concerns. Although the recent progress in constructing new plants has spurred design of more fully digital plant-wide ICHMI systems, the experience base in the nuclear power application domain is limited. Additionally, development of advanced reactor concepts, such as Generation IV designs and small modular reactors, introduces different plant conditions (e.g., higher temperatures, different coolants, etc.) and unique plant configurations (e.g., multiunit plants with shared systems, balance of plant architectures with reconfigurable co-generation options) that increase the need for enhanced ICHMI capabilities to fully achieve industry goals related to economic competitiveness, safety and reliability, sustainability, and proliferation resistance and physical protection. As a result, significant challenges remain to be addressed to enable the nuclear power industry to complete the transition to safe and comprehensive use of m

Holcomb, David Eugene [ORNL; Wood, Richard Thomas [ORNL

2013-01-01T23:59:59.000Z

225

Thermodynamic and transport properties of thoriaurania fuel of Advanced Heavy Water Reactor  

Science Journals Connector (OSTI)

High temperature thermochemistry of thoriaurania fuel for Advanced Heavy Water Reactor was investigated. Oxygen potential development within the matrix and distribution behaviors of the fission products (fps) in different phases were worked out with the help of thermodynamic and transport properties of the fps as well as fission generated oxygen and the detailed balance of the elements. Some of the necessary data for different properties were generated in this laboratory while others were taken from literatures. Noting the behavior of poor transports of gases and volatile species in the thoria rich fuel (thoria3mol% urania), the evaluation shows that the fuel will generally bear higher oxygen potential right from early stage of burnup, and Mo will play vital role to buffer the potential through the formation of its oxygen rich chemical states. The problems related to the poor transport and larger retention of fission gases (Xe) and volatiles (I, Te, Cs) are discussed.

M. Basu (Ali); R. Mishra; S.R. Bharadwaj; D. Das

2010-01-01T23:59:59.000Z

226

Relative performance properties of the ORNL Advanced Neutron Source Reactor with reduced enrichment fuels  

SciTech Connect (OSTI)

Three cores for the Advanced Neutron Source reactor, differing in size, enrichment, and uranium density in the fuel meat, have been analyzed. Performance properties of the reduced enrichment cores are compared with those of the HEU reference configuration. Core lifetime estimates suggest that none of these configurations will operate for the design goal of 17 days at 330 MW. With modes increases in fuel density and/or enrichment, however, the operating lifetimes of the HEU and MEU designs can be extended to the desired length. Achieving this lifetime with LEU fuel in any of the three studies cores, however, will require the successful development of denser fuels and/or structural materials with thermal neutron absorption cross sections substantially less than that of Al-6061. Relative to the HEU reference case, the peak thermal neutron flux in cores with reduced enrichment will be diminished by about 25--30%.

Bretscher, M.M.; Deen, J.R.; Hanan, N.A.; Matos, J.E.; Mo, S.C.; Pond, R.B.; Travelli, A.; Woodruff, W.L.

1994-12-31T23:59:59.000Z

227

SEPARATION OF HYDROGEN AND CARBON DIOXIDE USING A NOVEL MEMBRANE REACTOR IN ADVANCED FOSSIL ENERGY CONVERSION PROCESS  

SciTech Connect (OSTI)

Inorganic membrane reactors offer the possibility of combining reaction and separation in a single operation at high temperatures to overcome the equilibrium limitations experienced in conventional reactor configurations. Such attractive features can be advantageously utilized in a number of potential commercial opportunities, which include dehydrogenation, hydrogenation, oxidative dehydrogenation, oxidation and catalytic decomposition reactions. However, to be cost effective, significant technological advances and improvements will be required to solve several key issues which include: (a) permselective thin solid film, (b) thermal, chemical and mechanical stability of the film at high temperatures, and (c) reactor engineering and module development in relation to the development of effective seals at high temperature and high pressure. In this project, we are working on the development and application of palladium and palladium-silver alloy thin-film composite membranes in membrane reactor-separator configuration for simultaneous production and separation of hydrogen and carbon dioxide at high temperature. From our research on Pd-composite membrane, we have demonstrated that the new membrane has significantly higher hydrogen flux with very high perm-selectivity than any of the membranes commercially available. The steam reforming of methane by equilibrium shift in Pd-composite membrane reactor is being studied to demonstrate the potential application this new development. We designed and built a membrane reactor to study the reforming reaction. A two-dimensional pseudo-homogeneous reactor model was developed to study the performance of the membrane reactor parametrically. The important results are presented in this report.

Shamsuddin Illias

2002-06-10T23:59:59.000Z

228

Secondary Heat Exchanger Design and Comparison for Advanced High Temperature Reactor  

SciTech Connect (OSTI)

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

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

2012-06-01T23:59:59.000Z

229

Irradiation research capabilities at HFIR (High Flux Isotope Reactor) and ANS (Advanced Neutron Source)  

SciTech Connect (OSTI)

A variety of materials irradiation facilities exist in the High Flux Isotope Reactor (HFIR) and are planned for the Advanced Neutron Source (ANS) reactor. In 1986 the HFIR Irradiation Facilities Improvement (HIFI) project began modifications to the HFIR which now permit the operation of two instrumented capsules in the target region and eight capsules of 46-mm OD in the RB region. Thus, it is now possible to perform instrumented irradiation experiments in the highest continuous flux of thermal neutrons available in the western world. The new RB facilities are now large enough to permit neutron spectral tailoring of experiments and the modified method of access to these facilities permit rotation of experiments thereby reducing fluence gradients in specimens. A summary of characteristics of irradiation facilities in HFIR is presented. The ANS is being designed to provide the highest thermal neutron flux for beam facilities in the world. Additional design goals include providing materials irradiation and transplutonium isotope production facilities as good, or better than, HFIR. The reference conceptual core design consists of two annular fuel elements positioned one above the other instead of concentrically as in the HFIR. A variety of materials irradiation facilities with unprecedented fluxes are being incorporated into the design of the ANS. These will include fast neutron irradiation facilities in the central hole of the upper fuel element, epithermal facilities surrounding the lower fuel element, and thermal facilities in the reflector tank. A summary of characteristics of irradiation facilities presently planned for the ANS is presented. 2 tabs.

Thoms, K.R.

1990-01-01T23:59:59.000Z

230

Advanced Test Reactor (ATR) Facility 10CFR830 Safety Basis Related to Facility Experiments  

SciTech Connect (OSTI)

The Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR), a DOE Category A reactor, was designed to provide an irradiation test environment for conducting a variety of experiments. The ATR Safety Analysis Report, determined by DOE to meet the requirements of 10 CFR 830, Subpart B, provides versatility in types of experiments that may be conducted. This paper addresses two general types of experiments in the ATR facility and how safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore this type of experiment is addressed with more detail in the safety basis. This allows individual safety analyses for these experiments to be more routine and repetitive. The second type of experiment is less defined and is permitted under more general controls. Therefore, individual safety analyses for the second type of experiment tend to be more unique from experiment to experiment. Experiments are also discussed relative to "major modifications" and DOE-STD-1027-92. Application of the USQ process to ATR experiments is also discussed.

Tomberlin, Terry Alan

2002-06-01T23:59:59.000Z

231

Advanced Test Reactor (ATR) Facility 10CFR830 Safety Basis Related to Facility Experiments  

SciTech Connect (OSTI)

The Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR), a DOE Category A reactor, was designed to provide an irradiation test environment for conducting a variety of experiments. The ATR Safety Analysis Report, determined by DOE to meet the requirements of 10 CFR 830, Subpart B, provides versatility in types of experiments that may be conducted. This paper addresses two general types of experiments in the ATR facility and how safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore this type of experiment is addressed with more detail in the safety basis. This allows individual safety analyses for these experiments to be more routine and repetitive. The second type of experiment is less defined and is permitted under more general controls. Therefore, individual safety analyses for the second type of experiment tend to be more unique from experiment to experiment. Experiments are also discussed relative to ''major modifications'' and DOE-STD-1027-92. Application of the USQ process to ATR experiments is also discussed.

Tomberlin, T.A.

2002-06-19T23:59:59.000Z

232

Feasibility Study of Secondary Heat Exchanger Concepts for the Advanced High Temperature Reactor  

SciTech Connect (OSTI)

The work reported herein represents a significant step in the preliminary design of heat exchanger options (material options, thermal design, selection and evaluation methodology with existing challenges). The primary purpose of this study is to aid in the development and selection of the required heat exchanger for power production using either a subcritical or supercritical Rankine cycle.

Piyush Sabharwall

2011-09-01T23:59:59.000Z

233

Testing of advanced liquefaction concepts in HTI Run ALC-1: Coal cleaning and recycle solvent treatment  

SciTech Connect (OSTI)

In 1991, the Department of Energy initiated the Advanced Liquefaction Concepts Program to promote the development of new and emerging technology that has potential to reduce the cost of producing liquid fuels by direct coal liquefaction. Laboratory research performed by researchers at CAER, CONSOL, Sandia, and LDP Associates in Phase I is being developed further and tested at the bench scale at HTI. HTI Run ALC-1, conducted in the spring of 1996, was the first of four planned tests. In Run ALC-1, feed coal ash reduction (coal cleaning) by oil agglomeration, and recycle solvent quality improvement through dewaxing and hydrotreatment of the recycle distillate were evaluated. HTI`s bench liquefaction Run ALC-1 consisted of 25 days of operation. Major accomplishments were: 1) oil agglomeration reduced the ash content of Black Thunder Mine coal by 40%, from 5.5% to 3.3%; 2) excellent coal conversion of 98% was obtained with oil agglomerated coal, about 3% higher than the raw Black Thunder Mine coal, increasing the potential product yield by 2-3% on an MAF coal basis; 3) agglomerates were liquefied with no handling problems; 4) fresh catalyst make-up rate was decreased by 30%, with no apparent detrimental operating characteristics, both when agglomerates were fed and when raw coal was fed (with solvent dewaxing and hydrotreating); 5) recycle solvent treatment by dewaxing and hydrotreating was demonstrated, but steady-state operation was not achieved; and 6) there was some success in achieving extinction recycle of the heaviest liquid products. Performance data have not been finalized; they will be available for full evaluation in the new future.

Robbins, G.A.; Winschel, R.A.; Burke, F.P. [CONSOL, Inc., Library, PA (United States). Research and Development Dept.] [CONSOL, Inc., Library, PA (United States). Research and Development Dept.; Derbyshire, F.L.; Givens, E.N. [Kentucky Univ., Lexington, KY (United States). Center for Applied Energy Research] [Kentucky Univ., Lexington, KY (United States). Center for Applied Energy Research; Hu, J.; Lee, T.L.K. [Hydrocarbon Research, Inc., Lawrenceville, NJ (United States)] [Hydrocarbon Research, Inc., Lawrenceville, NJ (United States); Miller, J.E.; Stephens, H.P. [Sandia National Labs., Albuquerque, NM (United States)] [Sandia National Labs., Albuquerque, NM (United States); Peluso, M. [LDP Associates, Hamilton Square, NJ (United States)] [LDP Associates, Hamilton Square, NJ (United States)

1996-09-01T23:59:59.000Z

234

Effects of Levels of Automation for Advanced Small Modular Reactors: Impacts on Performance, Workload, and Situation Awareness  

SciTech Connect (OSTI)

The Human-Automation Collaboration (HAC) research effort is a part of the Department of Energy (DOE) sponsored Advanced Small Modular Reactor (AdvSMR) program conducted at Idaho National Laboratory (INL). The DOE AdvSMR program focuses on plant design and management, reduction of capital costs as well as plant operations and maintenance costs (O&M), and factory production costs benefits.

Johanna Oxstrand; Katya Le Blanc

2014-07-01T23:59:59.000Z

235

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

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

Water... it can be built on time and budget. Reactors currently under construction in Finland and France... are indeed well behind schedule. But there are several reactors that...

236

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

SciTech Connect (OSTI)

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

Not Available

1980-06-25T23:59:59.000Z

237

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

SciTech Connect (OSTI)

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

Not Available

1980-03-07T23:59:59.000Z

238

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

Science Journals Connector (OSTI)

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

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

2013-01-01T23:59:59.000Z

239

A global approach of the representativity concept: Application on a high-conversion light water reactor MOX lattice case  

SciTech Connect (OSTI)

The development of new types of reactor and the increase in the safety specifications and requirements induce an enhancement in both nuclear data knowledge and a better understanding of the neutronic properties of the new systems. This enhancement is made possible using ad hoc critical mock-up experiments. The main difficulty is to design these experiments in order to obtain the most valuable information. Its quantification is usually made by using representativity and transposition concepts. These theories enable to extract some information about a quantity of interest (an integral parameter) on a configuration, but generally a posteriori. This paper presents a more global approach of this theory, with the idea of optimizing the representativity of a new experiment, and its transposition a priori, based on a multiparametric approach. Using a quadratic sum, we show the possibility to define a global representativity which permits to take into account several quantities of interest at the same time. The maximization of this factor gives information about all quantities of interest. An optimization method of this value in relation to technological parameters (over-clad diameter, atom concentration) is illustrated on a high-conversion light water reactor MOX lattice case. This example tackles the problematic of plutonium experiment for the plutonium aging and a solution through the optimization of both the over-clad and the plutonium content. (authors)

Santos, N. D.; Blaise, P.; Santamarina, A. [CEA, DEN/DER/SPRC Cadarache, F-13108 Saint Paul-lez-Durance (France)

2013-07-01T23:59:59.000Z

240

E-Print Network 3.0 - advanced combustion concepts Sample Search...  

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nature for decision rules extracted with a data... plant," in Proc. Inst. Elect. Eng. Seminar Advanced Sensors In- strumentation ... Source: Kusiak, Andrew - Department of...

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

E-Print Network 3.0 - advanced propulsion concepts Sample Search...  

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on the field emission principle... advanced technology in power conversion. The main advantages of a propulsion system based on the field emis... AS THRUSTERS FOR ELECTRIC SPACE...

242

Analysis of fuel options for the breakeven core configuration of the Advanced Recycling Reactor  

SciTech Connect (OSTI)

A trade-off study is performed to determine the impacts of various fuel forms on the core design and core physics characteristics of the sodium-cooled Toshiba- Westinghouse Advanced Recycling Reactor (ARR). The fuel forms include oxide, nitride, and metallic forms of U and Th. The ARR core configuration is redesigned with driver and blanket regions in order to achieve breakeven fissile breeding performance with the various fuel types. State-of-the-art core physics tools are used for the analyses. In addition, a quasi-static reactivity balance approach is used for a preliminary comparison of the inherent safety performances of the various fuel options. Thorium-fueled cores exhibit lower breeding ratios and require larger blankets compared to the U-fueled cores, which is detrimental to core compactness and increases reprocessing and manufacturing requirements. The Th cores also exhibit higher reactivity swings through each cycle, which penalizes reactivity control and increases the number of control rods required. On the other hand, using Th leads to drastic reductions in void and coolant expansion coefficients of reactivity, with the potential for enhancing inherent core safety. Among the U-fueled ARR cores, metallic and nitride fuels result in higher breeding ratios due to their higher heavy metal densities. On the other hand, oxide fuels provide a softer spectrum, which increases the Doppler effect and reduces the positive sodium void worth. A lower fuel temperature is obtained with the metallic and nitride fuels due to their higher thermal conductivities and compatibility with sodium bonds. This is especially beneficial from an inherent safety point of view since it facilitates the reactor cool-down during loss of power removal transients. The advantages in terms of inherent safety of nitride and metallic fuels are maintained when using Th fuel. However, there is a lower relative increase in heavy metal density and in breeding ratio going from oxide to metallic or nitride Th fuels relative to the U counterpart fuels. (authors)

Stauff, N.E.; Klim, T.K.; Taiwo, T.A. [Argonne National Laboratory, Argonne, IL (United States); Fiorina, C. [Politecnico di Milano, Milan (Italy); Franceschini, F. [Westinghouse Electric Company LLC., Cranberry Township, Pennsylvania (United States)

2013-07-01T23:59:59.000Z

243

Evaluation Methodology for Advance Heat Exchanger Concepts Using Analytical Hierarchy Process  

SciTech Connect (OSTI)

The primary purpose of this study is to aid in the development and selection of the secondary/process heat exchanger (SHX) for power production and process heat application for a Next Generation Nuclear Reactors (NGNR). The potential options for use as an SHX are explored such as shell and tube, printed circuit heat exchanger. A shell and tube (helical coiled) heat exchanger is a recommended for a demonstration reactor because of its reliability while the reactor design is being further developed. The basic setup for the selection of the SHX has been established with evaluation goals, alternatives, and criteria. This study describes how these criteria and the alternatives are evaluated using the analytical hierarchy process (AHP).

Piyush Sabharwall; Eung Soo Kim

2012-07-01T23:59:59.000Z

244

AdvAnced  

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

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

245

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

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, and J.F. Stubbins4) Title: Final Report on In-reactor Creep-fatigue Deformation... , Finland 3) Reactor Technology Design Department, SCKCEN, 200 Boeretang, B-2400 Mol,...

246

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

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

, and J.F. Stubbins4) Title: Final Report on In-reactor Creep-fatigue Deformation... , Finland 3) Reactor Technology Design Department, SCKCEN, 200 Boeretang, B-2400 Mol,...

247

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

SciTech Connect (OSTI)

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

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

2013-10-01T23:59:59.000Z

248

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

E-Print Network [OSTI]

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

Moore, Eugene James Thomas

2006-08-16T23:59:59.000Z

249

E-Print Network 3.0 - advanced industrial concepts Sample Search...  

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

SECA fuel cells available for FutureGen 2020 MW-Scale SECA fuel cells for Advanced Coal Power Plants... 2010 400kW Modules -Residential, Commercial, Industrial CHP...

250

ITP Metal Casting: Advanced Melting Technologies: Energy Saving Concepts and Opportunities for the Metal Casting Industry  

Broader source: Energy.gov [DOE]

The study examines current and emerging melting technologies and discusses their technical barriers to scale-up issues and research needed to advance these technologies, improving melting efficiency, lowering metal transfer heat loss, and reducing scrap.

251

DOE Office of Indian Energy Renewable Energy Project Development: Advanced Financing Concepts  

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

Concepts Concepts Why It Makes Sense to Bring on a Third-Party Partner Course Outline What we will cover...  About the DOE Office of Indian Energy Education Initiative  Concepts for Financing Renewable Energy Projects on Tribal Lands - Levelized Cost of Energy (LCOE) - Business Structures - Tax-Equity Partnerships - Introduction  Additional Information and Resources 2 Introduction The U.S. Department of Energy (DOE) Office of Indian Energy Policy and Programs is responsible for assisting Tribes with energy planning and development, infrastructure, energy costs, and electrification of Indian lands and homes. As part of this commitment and on behalf of DOE, the Office of Indian Energy is leading education and capacity building efforts in Indian Country.

252

Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors  

Science Journals Connector (OSTI)

Abstract A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in the fuel pellet, leading to a slight increase in plutonium production. A parametric study determined the geometric conditions required to match cycle length requirements for each alternate cladding material in a PWR. A method for estimating the end of cycle reactivity was implemented to compare each model to that of standard Zircaloy-4 cladding. By using a thinner cladding of 350?m and keeping a constant outer diameter, austenitic stainless steels require an increase of no more than 0.5wt% enriched 235U to match fuel cycle requirements, while the required increase for FeCrAl was about 0.1%. When modeling SiC (with slightly lower thermal absorption properties than that of Zircaloy), a standard cladding thickness could be implemented with marginally less enriched uranium (?0.1%). Moderator temperature and void coefficients were calculated throughout the depletion cycle. Nearly identical reactivity responses were found when coolant temperature and void properties were perturbed for each cladding material. By splitting the pellet into 10 equal areal sections, relative fission power as a function of radius was found to be similar for each cladding material. FeCrAl and 310SS cladding have a slightly higher fission power near the pellets periphery due to the harder neutron spectrum in the system, causing more 239Pu breeding. An economic assessment calculated the change in fuel pellet production costs for use of each cladding. Implementing FeCrAl alloys would increase fuel pellet production costs about 15% because of increased 235U enrichment and the additional UO2 pellet volume enabled by using thinner cladding.

Nathan Michael George; Kurt Terrani; Jeff Powers; Andrew Worrall; Ivan Maldonado

2015-01-01T23:59:59.000Z

253

Scoping analysis of the Advanced Test Reactor using SN2ND  

SciTech Connect (OSTI)

A detailed set of calculations was carried out for the Advanced Test Reactor (ATR) using the SN2ND solver of the UNIC code which is part of the SHARP multi-physics code being developed under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in DOE-NE. The primary motivation of this work is to assess whether high fidelity deterministic transport codes can tackle coupled dynamics simulations of the ATR. The successful use of such codes in a coupled dynamics simulation can impact what experiments are performed and what power levels are permitted during those experiments at the ATR. The advantages of the SN2ND solver over comparable neutronics tools are its superior parallel performance and demonstrated accuracy on large scale homogeneous and heterogeneous reactor geometries. However, it should be noted that virtually no effort from this project was spent constructing a proper cross section generation methodology for the ATR usable in the SN2ND solver. While attempts were made to use cross section data derived from SCALE, the minimal number of compositional cross section sets were generated to be consistent with the reference Monte Carlo input specification. The accuracy of any deterministic transport solver is impacted by such an approach and clearly it causes substantial errors in this work. The reasoning behind this decision is justified given the overall funding dedicated to the task (two months) and the real focus of the work: can modern deterministic tools actually treat complex facilities like the ATR with heterogeneous geometry modeling. SN2ND has been demonstrated to solve problems with upwards of one trillion degrees of freedom which translates to tens of millions of finite elements, hundreds of angles, and hundreds of energy groups, resulting in a very high-fidelity model of the system unachievable by most deterministic transport codes today. A space-angle convergence study was conducted to determine the meshing and angular cubature requirements for the ATR, and also to demonstrate the feasibility of performing this analysis with a deterministic transport code capable of modeling heterogeneous geometries. The work performed indicates that a minimum of 260,000 linear finite elements combined with a L3T11 cubature (96 angles on the sphere) is required for both eigenvalue and flux convergence of the ATR. A critical finding was that the fuel meat and water channels must each be meshed with at least 3 'radial zones' for accurate flux convergence. A small number of 3D calculations were also performed to show axial mesh and eigenvalue convergence for a full core problem. Finally, a brief analysis was performed with different cross sections sets generated from DRAGON and SCALE, and the findings show that more effort will be required to improve the multigroup cross section generation process. The total number of degrees of freedom for a converged 27 group, 2D ATR problem is {approx}340 million. This number increases to {approx}25 billion for a 3D ATR problem. This scoping study shows that both 2D and 3D calculations are well within the capabilities of the current SN2ND solver, given the availability of a large-scale computing center such as BlueGene/P. However, dynamics calculations are not realistic without the implementation of improvements in the solver.

Wolters, E.; Smith, M. (NE NEAMS PROGRAM); ( SC)

2012-07-26T23:59:59.000Z

254

Advanced Neutron Source (ANS) Project progress report  

SciTech Connect (OSTI)

This report discusses the following topics on the advanced neutron source: quality assurance (QA) program; reactor core development; fuel element specification; corrosion loop tests and analyses; thermal-hydraulic loop tests; reactor control concepts; critical and subcritical experiments; material data, structural tests, and analysis; cold source development; beam tube, guide, and instrument development; hot source development; neutron transport and shielding; I C research and development; facility concepts; design; and safety.

McBee, M.R.; Chance, C.M. (eds.) (Oak Ridge National Lab., TN (USA)); Selby, D.L.; Harrington, R.M.; Peretz, F.J. (Oak Ridge National Lab., TN (USA))

1990-04-01T23:59:59.000Z

255

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

SciTech Connect (OSTI)

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

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

1994-04-01T23:59:59.000Z

256

Project Profile: Commercial Development of an Advanced Linear-Fresnel-Based CSP Concept  

Broader source: Energy.gov [DOE]

SkyFuel, under the CSP R&D FOA, is developing a commercial linear-Fresnel-based advanced CSP system called Linear Power Tower (LPT). The company aims to make significant improvements in the cost and viability of utility-scale dispatchable solar power.

257

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

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

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

258

E-Print Network 3.0 - advanced reactors part Sample Search Results  

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

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

259

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

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

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

260

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

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

Hotel bargains... to iron out their differences over the site of the International Thermonuclear Experimental Reactor Source: Fusiongnition Research Experiment (FIRE) Collection:...

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

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

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

fusion plant deal 1 hour, 28 minutes ago Representatives of more... ) International Thermonuclear Experimental Reactor (ITER), which will be built at ... Source: Fusiongnition...

262

A SCOPING STUDY OF ADVANCED THORIUM FUEL CYCLES FOR CANDU REACTORS.  

E-Print Network [OSTI]

?? A study was conducted to scope the relative merits of various thorium fuel cycles in CANDU reactors. It was determined that, due to the (more)

Friedlander, Yonni

2011-01-01T23:59:59.000Z

263

Advanced atomization concept for CWF burning in small combustors, Phase 2. Final technical report  

SciTech Connect (OSTI)

The program describes a concept referred to as opposed-jet atomization, which is particularly applicable to coal-water fuel (CWF). In the present atomizer design, two opposed jets of CWF are directed at each other and externally encounter a perpendicular blast of air at the collision point to create a spray of much finer droplets. The present Phase 2 program involved further evaluation of the opposed-jet atomizer performance and related tasks.

McHale, E.T.; Heaton, H.L.

1991-12-01T23:59:59.000Z

264

Advanced atomization concept for CWF burning in small combustors, Phase 2  

SciTech Connect (OSTI)

The program describes a concept referred to as opposed-jet atomization, which is particularly applicable to coal-water fuel (CWF). In the present atomizer design, two opposed jets of CWF are directed at each other and externally encounter a perpendicular blast of air at the collision point to create a spray of much finer droplets. The present Phase 2 program involved further evaluation of the opposed-jet atomizer performance and related tasks.

McHale, E.T.; Heaton, H.L.

1991-12-01T23:59:59.000Z

265

Fission product transport and behavior during two postulated loss-of-flow transients in the Advanced Test Reactor  

SciTech Connect (OSTI)

The fission product behavior during two postulated loss-of-flow accidents (leading to high- and low-pressure core degradations) in the Advanced Test Reactor (ATR) has been analyzed. These transients are designated ATR transients LCP 15 (high pressure) and LPP9 (low pressure). Normally, transients of this nature would be easily mitigated using existing safety systems and procedures. In these analyses, failure of these safety systems was assumed so that core degradation and fission product release could be studied. A probabilistic risk analysis was performed that indicated that the probability of occurrence for these two transients is on the order of 10[sup [minus]5] and 10[sup [minus]7] per reactor year for LCP15 and LPP9, respectively. The fission product behavior analysis included calculations of the gaseous and highly volatile fission product (xenon, krypton, cesium, iodine, and tellurium) inventories in the fuel before accident initiation, release of the fission products from the fuel into the reactor vessel during core melt, the probable chemical forms, and transport of the fission products from the core through the reactor vessel and existing piping to the confinement. In addition to a base-case analysis of fission product behavior, a series of analyses was performed to determine the sensitivity of fission product release to several parameters including steam flow rate, (structural) aluminum oxidation, and initial aerosol size. The base-case analyses indicate that the volatile fission products (excluding the noble gases) will be transported as condensed species on zinc aerosols.

Adams, J.P.; Carboneau, M.L.; Hagrman, D.L. (Idaho National Engineering Lab. EG and G Idaho, Idaho Falls, ID (United States))

1993-07-01T23:59:59.000Z

266

INITIATORS AND TRIGGERING CONDITIONS FOR ADAPTIVE AUTOMATION IN ADVANCED SMALL MODULAR REACTORS  

SciTech Connect (OSTI)

It is anticipated that Advanced Small Modular Reactors (AdvSMRs) will employ high degrees of automation. High levels of automation can enhance system performance, but often at the cost of reduced human performance. Automation can lead to human out-of the loop issues, unbalanced workload, complacency, and other problems if it is not designed properly. Researchers have proposed adaptive automation (defined as dynamic or flexible allocation of functions) as a way to get the benefits of higher levels of automation without the human performance costs. Adaptive automation has the potential to balance operator workload and enhance operator situation awareness by allocating functions to the operators in a way that is sensitive to overall workload and capabilities at the time of operation. However, there still a number of questions regarding how to effectively design adaptive automation to achieve that potential. One of those questions is related to how to initiate (or trigger) a shift in automation in order to provide maximal sensitivity to operator needs without introducing undesirable consequences (such as unpredictable mode changes). Several triggering mechanisms for shifts in adaptive automation have been proposed including: operator initiated, critical events, performance-based, physiological measurement, model-based, and hybrid methods. As part of a larger project to develop design guidance for human-automation collaboration in AdvSMRs, researchers at Idaho National Laboratory have investigated the effectiveness and applicability of each of these triggering mechanisms in the context of AdvSMR. Researchers reviewed the empirical literature on adaptive automation and assessed each triggering mechanism based on the human-system performance consequences of employing that mechanism. Researchers also assessed the practicality and feasibility of using the mechanism in the context of an AdvSMR control room. Results indicate that there are tradeoffs associated with each mechanism, but that some are more applicable to the AdvSMR domain. The two mechanisms that consistently improve performance in laboratory studies are operator initiated adaptive automation based on hierarchical task delegation and the Electroencephalogram(EEG) based measure of engagement. Current EEG methods are intrusive and require intensive analysis; therefore it is not recommended for an AdvSMR control rooms at this time. Researchers also discuss limitations in the existing empirical literature and make recommendations for further research.

Katya L Le Blanc; Johanna h Oxstrand

2014-04-01T23:59:59.000Z

267

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

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

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

268

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

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

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

269

Stability analysis of the boiling water reactor : methods and advanced designs  

E-Print Network [OSTI]

Density Wave Oscillations (DWOs) are known to be possible when a coolant undergoes considerable density reduction while passing through a heated channel. In the development of boiling water reactors (BWRs), there has been ...

Hu, Rui, Ph. D. Massachusetts Institute of Technology

2010-01-01T23:59:59.000Z

270

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

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

20 03012006 09:51 AMLoading "People's Daily Online --Chinese experimental thermonuclear reactor on discharge test in July" Page 1 of 1http:english.people.com.cn200603...

271

Radiation Resistance of XLPE Nano-dielectrics for Advanced Reactor Applications  

SciTech Connect (OSTI)

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

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

2014-01-01T23:59:59.000Z

272

Research reactors - an overview  

SciTech Connect (OSTI)

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

West, C.D.

1997-03-01T23:59:59.000Z

273

Preliminary Results of an On-Line, Multi-Spectrometer Fission Product Monitoring System to Support Advanced Gas Reactor Fuel Testing and Qualification in the Advanced Test Reactor at the Idaho National Laboratory  

SciTech Connect (OSTI)

The Advanced Gas Reactor -1 (AGR-1) experiment is the first experiment in a series of eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments scheduled for placement in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and will continue irradiation for about 2.5 years. During this time six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The goals of the irradiation experiment is to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. This paper presents the preliminary test details of the fuel performance, as measured by the control and acquisition software.

Dawn M. Scates; John K. Hartwell; John B. Walter; Mark W. Drigert

2007-10-01T23:59:59.000Z

274

Reactor Materials | Department of Energy  

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

Benefits Crosscutting Technology Development Reactor Materials Advanced Sensors and Instrumentation Proliferation and Terrorism Risk Assessment Advanced Methods for Manufacturing...

275

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

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

& Advanced Separations Technology ITM Syngas... ) Fossil-Based Hydrogen Production Praxair Praxair SNL TIAX Integrated Ceramic Membrane ... Source: DOE Office of Energy...

276

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

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

of Physics, Stanford University Collection: Physics 69 PYROLYSIS, THERMAL GASIFICATION, AND LIQUEFACTION OF SOLID WASTES AND RESIDUES Summary: with advanced thermal...

277

Advanced combustor design concepts to control NO{sub x} and air toxics. Quarterly report  

SciTech Connect (OSTI)

The University of Utah, Massachusetts Institute of Technology (MIT), Reaction Engineering International (REI) and ABB/Combustion Engineering have joined together in this research proposal to develop fundamental understanding regarding the impact of fuel and combustion changes on ignition stability and flame characteristics because these critically affect: NO{sub x} emissions, carbon burnout, and emissions of air toxics; existing laboratory and bench scale facilities are being used to generate critical missing data which will be used to improve the NO{sub x} and carbon burnout submodels in comprehensive combustion simulation tools currently being used by industrial boiler manufacturers. To ensure effective and timely transfer of This technology, a major manufacturer (ABB) and a combustion model supplier (REI) have been included as part of the team from the early conception of the proposal. ABB/Combustion Engineering is providing needed fundamental data on the extent of volatile evolution from commercial coals as well as background information on current design needs in industrial practice. MIT is responsible for the development of an improved char nitrogen oxidation model which will ultimately be incorporated into an enhanced NO{sup x} submodel. Reaction Engineering International is providing the lead engineering staff for the experimental studies and an overall industrial focus for the work based on their use of the combustion simulation tools for a wide variety of industries. The University of Utah is conducting bench scale experimentation to (1) investigate alternative methods for enhancing flame stability to reduce NO{sub x} emissions and (2) characterize air toxic emissions under ultralow NO{sub x} conditions. Accomplishments for this quarter are presented to the solid sampling system and char nitrogen modeling.

Pershing, D.W.; Lighty, J.; Veranth, J. [Utah Univ., Salt Lake City, UT (United States). Coll. of Engineering; Sarofim, A.; Goel, S. [Massachusetts Inst. of Tech., Cambridge, MA (United States)

1995-04-28T23:59:59.000Z

278

Advanced light water reactor plants System 80+{trademark} design certification program. Annual progress report, October 1, 1994--September 30, 1995  

SciTech Connect (OSTI)

The purpose of this report is to provide the status of the progress that was made towards Design Certification of System 80+{trademark} during the US government`s 1995 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW (1350 MWe) Pressurized Water Reactor (PWR). The design covers an essentially complete plant. It is based on EPRI ALWR Utility Requirements Document (URD) improvements to the Standardized System 80 Nuclear Steam Supply System (NSSS) in operation at Palo Verde Units 1, 2, and 3. The NSSS is a traditional two-loop arrangement with two steam generators, two hot legs and four cold legs, each with a reactor coolant pump. The System 80+ standard design houses the NSSS in a spherical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual barrier to radioactivity release. Other major features include an all-digital, human-factors-engineered control room, an alternate electrical AC power source, an In-Containment Refueling Water Storage Tank (IRWST), and plant arrangements providing complete separation of redundant trains in safety systems.

NONE

1998-09-01T23:59:59.000Z

279

Plan for fully decontaminating and decommissioning of the Westinghouse Advanced Reactors Division Fuel Laboratories at Cheswick, Revision 3  

SciTech Connect (OSTI)

The project scope of work included the complete decontamination and decommissioning (D and D) of the Westinghouse ARD Fuel Laboratories at the Cheswick Site in the shortest possible time. This has been accomplished in the following four phases: (1) preparation of documents and necessary paperwork; packaging and shipping of all special nuclear materials in an acceptable form to a reprocessing agency; (2) decontamination of all facilities, glove boxes and equipment; loading of generated waste into bins, barrels and strong wooden boxes; (3) shipping of all bins, barrels and boxes containing waste to the designated burial site; removal of all utility services from the laboratories; (4) final survey of remaining facilities and certification for nonrestricted use; preparation of final report. This volume contains the following 3 attachments: (1) Plan for Fully Decontamination and Decommissioning of the Westinghouse Advanced Reactors Division Fuel Laboratories at Cheswick; (2) Environmental Assessment for Decontamination and Decommissioning the Westinghouse Advanced Reactors Division Plutonium Fuel Laboratories, Cheswick, PA; and (3) WARD-386, Quality Assurance Program Description for Decontamination and Decommissioning Activities.

Not Available

1982-01-01T23:59:59.000Z

280

Generic small modular reactor plant design.  

SciTech Connect (OSTI)

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

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

2012-12-01T23:59:59.000Z

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

An Analysis of Methanol and Hydrogen Production via High-Temperature Electrolysis Using the Sodium Cooled Advanced Fast Reactor  

SciTech Connect (OSTI)

Integration of an advanced, sodium-cooled fast spectrum reactor into nuclear hybrid energy system (NHES) architectures is the focus of the present study. A techno-economic evaluation of several conceptual system designs was performed for the integration of a sodium-cooled Advanced Fast Reactor (AFR) with the electric grid in conjunction with wind-generated electricity. Cases in which excess thermal and electrical energy would be reapportioned within an integrated energy system to a chemical plant are presented. The process applications evaluated include hydrogen production via high temperature steam electrolysis and methanol production via steam methane reforming to produce carbon monoxide and hydrogen which feed a methanol synthesis reactor. Three power cycles were considered for integration with the AFR, including subcritical and supercritical Rankine cycles and a modified supercritical carbon dioxide modified Brayton cycle. The thermal efficiencies of all of the modeled power conversions units were greater than 40%. A thermal efficiency of 42% was adopted in economic studies because two of the cycles either performed at that level or could potentially do so (subcritical Rankine and S-CO2 Brayton). Each of the evaluated hybrid architectures would be technically feasible but would demonstrate a different internal rate of return (IRR) as a function of multiple parameters; all evaluated configurations showed a positive IRR. As expected, integration of an AFR with a chemical plant increases the IRR when must-take wind-generated electricity is added to the energy system. Additional dynamic system analyses are recommended to draw detailed conclusions on the feasibility and economic benefits associated with AFR-hybrid energy system operation.

Shannon M. Bragg-Sitton; Richard D. Boardman; Robert S. Cherry; Wesley R. Deason; Michael G. McKellar

2014-03-01T23:59:59.000Z

282

SEPARATION OF HYDROGEN AND CARBON DIOXIDE USING A NOVEL MEMBRANE REACTOR IN ADVANCED FOSSIL ENERGY CONVERSION PROCESS  

SciTech Connect (OSTI)

Inorganic membrane reactors offer the possibility of combining reaction and separation in a single operation at high temperatures to overcome the equilibrium limitations experienced in conventional reactor configurations. Such attractive features can be advantageously utilized in a number of potential commercial opportunities, which include dehydrogenation, hydrogenation, oxidative dehydrogenation, oxidation and catalytic decomposition reactions. However, to be cost effective, significant technological advances and improvements will be required to solve several key issues which include: (a) permselective thin solid film, (b) thermal, chemical and mechanical stability of the film at high temperatures, and (c) reactor engineering and module development in relation to the development of effective seals at high temperature and high pressure. In this project, we are working on the development and application of palladium and palladium-silver alloy thin-film composite membranes in membrane reactor-separator configuration for simultaneous production and separation of hydrogen and carbon dioxide at high temperature. From our research on Pd-composite membrane, we have demonstrated that the new membrane has significantly higher hydrogen flux with very high perm-selectivity than any of the membranes commercially available. The steam reforming of methane by equilibrium shift in Pd-composite membrane reactor is being studied to demonstrate the potential application of this new development. A two-dimensional, pseudo-homogeneous membrane-reactor model was developed to investigate the steam-methane reforming (SMR) reactions in a Pd-based membrane reactor. Radial diffusion was taken into consideration to account for the concentration gradient in the radial direction due to hydrogen permeation through the membrane. With appropriate reaction rate expressions, a set of partial differential equations was derived using the continuity equation for the reaction system. The equations were solved by finite difference method. The solution of the model equations is complicated by the coupled reactions. At the inlet, if there is no hydrogen, rate expressions become singular. To overcome this problem, the first element of the reactor was treated as a continuous stirred tank reactor (CSTR). Several alternative numerical schemes were implemented in the solution algorithm to get a converged, stable solution. The model was also capable of handling steam-methane reforming reactions under non-membrane condition and equilibrium reaction conversions. Some of the numerical results were presented in the previous report. To test the membrane reactor model, we fabricated Pd-stainless steel membranes in tubular configuration using electroless plating method coupled with osmotic pressure. Scanning Electron Microscopy (SEM) and Energy Dispersive Xray (EDX) were used to characterize the fabricated Pd-film composite membranes. Gas-permeation tests were performed to measure the permeability of hydrogen, nitrogen and helium using pure gas. Some of these results are discussed in this progress report.

Shamsuddin Ilias

2004-02-17T23:59:59.000Z

283

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

SciTech Connect (OSTI)

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

Blaine Grover

2012-10-01T23:59:59.000Z

284

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

Science Journals Connector (OSTI)

......during their short/long term-storage is investigated in...radio-toxicity of radioactive waste is widely regarded...exchangers of the spent fuel storage bay. The decay power...VVER type reactors at long-term storage. Radiat. Prot. Dosim......

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

2010-01-01T23:59:59.000Z

285

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

E-Print Network [OSTI]

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

Jang, Si Young

2005-11-01T23:59:59.000Z

286

10 CFR 830 Major Modification Determination for Advanced Test Reactor RDAS and LPCIS Replacement  

SciTech Connect (OSTI)

The replacement of the ATR Control Complex's obsolete computer based Reactor Data Acquisition System (RDAS) and its safety-related Lobe Power Calculation and Indication System (LPCIS) software application is vitally important to ensure the ATR remains available to support this national mission. The RDAS supports safe operation of the reactor by providing 'real-time' plant status information (indications and alarms) for use by the reactor operators via the Console Display System (CDS). The RDAS is a computer support system that acquires analog and digital information from various reactor and reactor support systems. The RDAS information is used to display quadrant and lobe powers via a display interface more user friendly than that provided by the recorders and the Control Room upright panels. RDAS provides input to the Nuclear Engineering ATR Surveillance Data System (ASUDAS) for fuel burn-up analysis and the production of cycle data for experiment sponsors and the generation of the Core Safety Assurance Package (CSAP). RDAS also archives and provides for retrieval of historical plant data which may be used for event reconstruction, data analysis, training and safety analysis. The RDAS, LPCIS and ASUDAS need to be replaced with state-of-the-art technology in order to eliminate problems of aged computer systems, and difficulty in obtaining software upgrades, spare parts, and technical support. The major modification criteria evaluation of the project design did not lead to the conclusion that the project is a major modification. The negative major modification determination is driven by the fact that the project requires a one-for-one equivalent replacement of existing systems that protects and maintains functional and operational requirements as credited in the safety basis.

David E. Korns

2012-05-01T23:59:59.000Z

287

Advanced-ignition-concept exploration on OMEGA This article has been downloaded from IOPscience. Please scroll down to see the full text article.  

E-Print Network [OSTI]

at the Omega Laser Facility. Integrated fast-ignition experiments with room-temperature re-entrant cone targets from the short-pulse laser into the core by fast electrons. In shock- ignition experiments, sphericalAdvanced-ignition-concept exploration on OMEGA This article has been downloaded from IOPscience

288

Advanced Core Design And Fuel Management For Pebble-Bed Reactors  

SciTech Connect (OSTI)

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

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

2004-10-01T23:59:59.000Z

289

Advanced Test Reactor Safety Basis Upgrade Lessons Learned Relative to Design Basis Verification and Safety Basis Management  

SciTech Connect (OSTI)

The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The reactor also provides other irradiation services such as radioisotope production. The ATR and its support facilities are located at the Test Reactor Area of the Idaho National Engineering and Environmental Laboratory (INEEL). An audit conducted by the Department of Energy's Office of Independent Oversight and Performance Assurance (DOE OA) raised concerns that design conditions at the ATR were not adequately analyzed in the safety analysis and that legacy design basis management practices had the potential to further impact safe operation of the facility.1 The concerns identified by the audit team, and issues raised during additional reviews performed by ATR safety analysts, were evaluated through the unreviewed safety question process resulting in shutdown of the ATR for more than three months while these concerns were resolved. Past management of the ATR safety basis, relative to facility design basis management and change control, led to concerns that discrepancies in the safety basis may have developed. Although not required by DOE orders or regulations, not performing design basis verification in conjunction with development of the 10 CFR 830 Subpart B upgraded safety basis allowed these potential weaknesses to be carried forward. Configuration management and a clear definition of the existing facility design basis have a direct relation to developing and maintaining a high quality safety basis which properly identifies and mitigates all hazards and postulated accident conditions. These relations and the impact of past safety basis management practices have been reviewed in order to identify lessons learned from the safety basis upgrade process and appropriate actions to resolve possible concerns with respect to the current ATR safety basis. The need for a design basis reconstitution program for the ATR has been identified along with the use of sound configuration management principles in order to support safe and efficient facility operation.

G. L. Sharp; R. T. McCracken

2004-05-01T23:59:59.000Z

290

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

SciTech Connect (OSTI)

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

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

2011-12-29T23:59:59.000Z

291

Advanced Test Reactor Complex Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables  

SciTech Connect (OSTI)

U.S. Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Advanced Test Reactor Complex facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. U.S. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool to develop the radioactive waste management basis.

Lisa Harvego; Brion Bennett

2011-11-01T23:59:59.000Z

292

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

SciTech Connect (OSTI)

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

Not Available

1993-12-31T23:59:59.000Z

293

Advanced Accelerator Concepts Workshop  

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

acceleration at the BNL-ATF Thomas Marshall GeVm WAKE FIELDS GENERATED BY A TRAIN OF pC, FEMTOSECOND BUNCHES IN A PLANAR DIELECTRIC MICROSTRUCTURE Changbiao Wang GeVm...

294

Advanced Accelerator Concepts Workshop  

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

EM Structure-Based Accelerators Working Group Group-Leader: Wayne Kimura, STI Optronics (wkimura@stioptronics.com) Group-Co-leader: Steve Lidia, LBNL (SMLidia@lbl.gov)...

295

FROM CONCEPT TO REALITY, IN-SITU DECOMMISSIONING OF THE P AND R REACTORS AT THE SAVANNAH RIVER SITE  

SciTech Connect (OSTI)

SRS recently completed an approximately three year effort to decommission two SRS reactors: P-Reactor (Building 105-P) and R-Reactor (Building 105-R). Completed in December 2011, the concurrent decommissionings marked the completion of two relatively complex and difficult facility disposition projects at the SRS. Buildings 105-P and 105-R began operating as production reactors in the early 1950s with the mission of producing weapons material (e.g., tritium and plutonium-239). The 'P' Reactor and was shutdown in 1991 while the 'R' Reactor and was shutdown in 1964. In the intervening period between shutdown and deactivation & decommissioning (D&D), Buildings 105-P and 105-R saw limited use (e.g., storage of excess heavy water and depleted uranium oxide). For Building 105-P, deactivation was initiated in April 2007 and was essentially complete by June 2010. For Building 105-R, deactivation was initiated in August 2008 and was essentially complete by September 2010. For both buildings, the primary objective of deactivation was to remove/mitigate hazards associated with the remaining hazardous materials, and thus prepare the buildings for in-situ decommissioning. Deactivation removed the following hazardous materials to the extent practical: combustibles/flammables, residual heavy water, acids, friable asbestos (as needed to protect workers performing deactivation and decommissioning), miscellaneous chemicals, lead/brass components, Freon(reg sign), oils, mercury/PCB containing components, mold and some radiologically-contaminated equipment. In addition to the removal of hazardous materials, deactivation included the removal of hazardous energy, exterior metallic components (representing an immediate fall hazard), and historical artifacts along with the evaporation of water from the two Disassembly Basins. Finally, so as to facilitate occupancy during the subsequent in-situ decommissioning, deactivation implemented repairs to the buildings and provided temporary power.

Musall, J.; Blankenship, J.; Griffin, W.

2012-01-09T23:59:59.000Z

296

Modular Pebble-Bed Reactor Project: Laboratory-Directed Research and Development Program FY 2002 Annual Report  

SciTech Connect (OSTI)

This report documents the results of our research in FY-02 on pebble-bed reactor technology under our Laboratory Directed Research and Development (LDRD) project entitled the Modular Pebble-Bed Reactor. The MPBR is an advanced reactor concept that can meet the energy and environmental needs of future generations under DOEs Generation IV initiative. Our work is focused in three areas: neutronics, core design and fuel cycle; reactor safety and thermal hydraulics; and fuel performance.

Petti, David Andrew; Dolan, Thomas James; Miller, Gregory Kent; Moore, Richard Leroy; Terry, William Knox; Ougouag, Abderrafi Mohammed-El-Ami; Oh, Chang H; Gougar, Hans D

2002-11-01T23:59:59.000Z

297

Advanced UNpPu fuel to achieve long-life core in heavy water reactor  

Science Journals Connector (OSTI)

The objective of this paper is to look at the possibility of approaching the long-life core comparable with reactor life-time. The main issues are centered on UNpPu fuel in a tight lattice design with heavy water as a coolant. It is found that in a hard neutron spectrum thus obtained, a large fraction of 238Pu produced by neutron capture in 237Np not only protects plutonium against uncontrolled proliferation, but substantially contributes in keeping criticality due to improved fissile properties (its capture-to-fission ratio drops below unit). Equilibrium fuel composition demonstrates excellent conversion properties that yield the burn-up value as high as 200 GWd/t at extremely small reactivity swings.

K. Nikitin; M. Saito; V. Artisyuk; A. Chmelev; V. Apse

1999-01-01T23:59:59.000Z

298

A Framework to Expand and Advance Probabilistic Risk Assessment to Support Small Modular Reactors  

SciTech Connect (OSTI)

During the early development of nuclear power plants, researchers and engineers focused on many aspects of plant operation, two of which were getting the newly-found technology to work and minimizing the likelihood of perceived accidents through redundancy and diversity. As time, and our experience, has progressed, the realization of plant operational risk/reliability has entered into the design, operation, and regulation of these plants. But, to date, we have only dabbled at the surface of risk and reliability technologies. For the next generation of small modular reactors (SMRs), it is imperative that these technologies evolve into an accepted, encompassing, validated, and integral part of the plant in order to reduce costs and to demonstrate safe operation. Further, while it is presumed that safety margins are substantial for proposed SMR designs, the depiction and demonstration of these margins needs to be better understood in order to optimize the licensing process.

Curtis Smith; David Schwieder; Robert Nourgaliev; Cherie Phelan; Diego Mandelli; Kellie Kvarfordt; Robert Youngblood

2012-09-01T23:59:59.000Z

299

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

SciTech Connect (OSTI)

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

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

2014-06-10T23:59:59.000Z

300

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

DOE Patents [OSTI]

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

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

2013-11-05T23:59:59.000Z

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


301

Advanced ultrasonic imaging of piping in Dresden and Vermont Yankee reactors  

SciTech Connect (OSTI)

Pacific Northwest Laboratory (PNL) has been developing the synthetic aperture focusing technique for ultrasonic testing (SAFT-UT) for the US Nuclear Regulatory Commission (NRC). Objectives are to make a real-time field system for performing all required inservice inpsections on nuclear power plants. The NRC funded the laboratory phase of the work at the University of Michigan. The PNL role is to transfer the technology to the field. This paper reports on the experience gained during 1984 involving the inspection of two boiling water reactor (BWR) plants in the United States. These field trials were very successful and demonstrated the advantages of using a flexible, high resolution imaging system to aid in detecting and interpreting ultrasonic reflectors.

Doctor, S.R.; Hall, T.E.; Crawford, S.L.

1985-03-01T23:59:59.000Z

302

The advanced thermionics initiative. program update  

SciTech Connect (OSTI)

The United States Air Force has had a long standing interest in thermionic space power dating back to the early 1960s when a heat pipe cooled thermionic converter was demonstrated through work at the predecessor to Wright Laboratory (WL). With the exception of the short hiatus in the mid-70s, Air Force thermionics work at Wright Laboratory has continued to the present time with thermionic technology programs including the burst power thermionic phase change concepts, heat pipe cooled planar diodes, and advanced in-core concept developments such as composite materials, insulators and oxygenation. The Advanced Thermionics Initiative (ATI) program was organized to integrate thermionic technology advances into a converter suitable for in-core reactor applications in the 10 to 40 kWe power range. As an advanced thermionics technology program, the charter and philosophy of the ATI program is to provide the needed advanced converter concepts in support of national thermionic space power programs.

Lamp, T.R.; Donovan, B.D. (Aerospace Power Division, Wright Laboratory, Wright-Patterson AFB, Ohio 45433 (United States))

1993-01-20T23:59:59.000Z

303

An Advanced Computational Scheme for the Optimization of 2D Radial Reflectors in Pressurized Water Reactors  

E-Print Network [OSTI]

This paper presents a computational scheme for the determination of equivalent 2D multi-group heterogeneous reflectors in a Pressurized Water Reactor (PWR). The proposed strategy is to define a full-core calculation consistent with a reference lattice code calculation such as the Method Of Characteristics (MOC) as implemented in APOLLO2 lattice code. The computational scheme presented here relies on the data assimilation module known as "Assimilation de donn\\'{e}es et Aide \\`{a} l'Optimisation (ADAO)" of the SALOME platform developed at \\'{E}lectricit\\'{e} De France (EDF), coupled with the full-core code COCAGNE and with the lattice code APOLLO2. A first validation of the computational scheme is made using the OPTEX reflector model developed at \\'{E}cole Polytechnique de Montr\\'{e}al (EPM). As a result, we obtain 2D multi-group, spatially heterogeneous 2D reflectors, using both diffusion or $\\text{SP}_{\\text{N}}$ operators. We observe important improvements of the power discrepancies distribution over the cor...

Clerc, Thomas; Leroyer, Hadrien; Argaud, Jean-Philippe; Bouriquet, Bertrand; Ponot, Aglique

2014-01-01T23:59:59.000Z

304

Some Specific CASL Requirements for Advanced Multiphase Flow Simulation of Light Water Reactors  

SciTech Connect (OSTI)

Because of the diversity of physical phenomena occuring in boiling, flashing, and bubble collapse, and of the length and time scales of LWR systems, it is imperative that the models have the following features: Both vapor and liquid phases (and noncondensible phases, if present) must be treated as compressible. Models must be mathematically and numerically well-posed. The models methodology must be multi-scale. A fundamental derivation of the multiphase governing equation system, that should be used as a basis for advanced multiphase modeling in LWR coolant systems, is given in the Appendix using the ensemble averaging method. The remainder of this work focuses specifically on the compressible, well-posed, and multi-scale requirements of advanced simulation methods for these LWR coolant systems, because without these are the most fundamental aspects, without which widespread advancement cannot be claimed. Because of the expense of developing multiple special-purpose codes and the inherent inability to couple information from the multiple, separate length- and time-scales, efforts within CASL should be focused toward development of a multi-scale approaches to solve those multiphase flow problems relevant to LWR design and safety analysis. Efforts should be aimed at developing well-designed unified physical/mathematical and high-resolution numerical models for compressible, all-speed multiphase flows spanning: (1) Well-posed general mixture level (true multiphase) models for fast transient situations and safety analysis, (2) DNS (Direct Numerical Simulation)-like models to resolve interface level phenmena like flashing and boiling flows, and critical heat flux determination (necessarily including conjugate heat transfer), and (3) Multi-scale methods to resolve both (1) and (2) automatically, depending upon specified mesh resolution, and to couple different flow models (single-phase, multiphase with several velocities and pressures, multiphase with single velocity and pressure, etc.) A unified, multi-scale approach is advocated to extend the necessary foundations and build the capability to simultaneously solve the fluid dynamic interface problems (interface resolution) as well as multiphase mixtures (homogenization).

R. A. Berry

2010-11-01T23:59:59.000Z

305

Site Considerations for Repowering With Advanced Circulating Pressurized Fluidized Bed Combustion (APFBC) from the L.V. Sutton Station Concept Assessment  

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

Tonnemacher et al., Site Considerations for Repowering With APFBC from the L.V. Sutton Station Concept Assessment Tonnemacher et al., Site Considerations for Repowering With APFBC from the L.V. Sutton Station Concept Assessment paper 970562 Page 1 of 36 Site Considerations for Repowering with Advanced Circulating Pressurized Fluidized Bed Combustion (APFBC) from the L.V. Sutton Station Concept Assessment Gary C. Tonnemacher, P.E., and David C. Killen, P.E. Carolina Power & Light Company Raleigh, North Carolina Richard E. Weinstein, P.E., Harvey N. Goldstein, P.E., and Jay S. White Parsons Power Group Inc. Reading, Pennsylvania Robert W. Travers, P.E. U.S. Department of Energy Office of Fossil Energy / Germantown, Maryland electronic mail addresses/phone no. electronic mail addresses/phone no. Tonnemacher{ Gary.Tonnemacher@CPLC.COM 919 / 546-6091 Goldstein { Harvey_N_Goldstein@Parsons.COM

306

The Windscale Advanced Gas Cooled Reactor (WAGR) Decommissioning Project A Close Out Report for WAGR Decommissioning Campaigns 1 to 10 - 12474  

SciTech Connect (OSTI)

The reactor core of the Windscale Advanced Gas-Cooled Reactor (WAGR) has been dismantled as part of an ongoing decommissioning project. The WAGR operated until 1981 as a development reactor for the British Commercial Advanced Gas cooled Reactor (CAGR) power programme. Decommissioning began in 1982 with the removal of fuel from the reactor core which was completed in 1983. Subsequently, a significant amount of engineering work was carried out, including removal of equipment external to the reactor and initial manual dismantling operations at the top of the reactor, in preparation for the removal of the reactor core itself. Modification of the facility structure and construction of the waste packaging plant served to provide a waste route for the reactor components. The reactor core was dismantled on a 'top-down' basis in a series of 'campaigns' related to discrete reactor components. This report describes the facility, the modifications undertaken to facilitate its decommissioning and the strategies employed to recognise the successful decommissioning of the reactor. Early decommissioning tasks at the top of the reactor were undertaken manually but the main of the decommissioning tasks were carried remotely, with deployment systems comprising of little more than crane like devices, intelligently interfaced into the existing structure. The tooling deployed from the 3 tonne capacity (3te) hoist consisted either purely mechanical devices or those being electrically controlled from a 'push-button' panel positioned at the operator control stations, there was no degree of autonomy in the 3te hoist or any of the tools deployed from it. Whilst the ATC was able to provide some tele-robotic capabilities these were very limited and required a good degree of driver input which due to the operating philosophy at WAGR was not utilised. The WAGR box proved a successful waste package, adaptable through the use of waste box furniture specific to the waste-forms generated throughout the various decommissioning campaigns. The use of low force compaction for insulation and soft wastes provided a simple, robust and cost effective solution as did the direct encapsulation of LLW steel components in the later stages of reactor decommissioning. Progress through early campaigns was good, often bettering the baseline schedule, especially when undertaking the repetitive tasks seen during Neutron Shield and Graphite Core decommissioning, once the operators had become experienced with the equipment, though delays became more pronounced, mainly as a result of increased failures due to the age and maintainability of the RDM and associated equipment. Extensive delays came about as a result of the unsupported insulation falling away from the pressure vessel during removal and the inability of the ventilation system to manage the sub micron particulate generated during IPOPI cutting operations, though the in house development of revised and new methodologies ultimately led to the successful completion of PV and I removal. In a programme spanning over 12 years, the decommissioning of the reactor pressure vessel and core led to the production 110 ILW and 75 LLW WAGR boxes, with 20 LLW ISO freight containers of primary reactor wastes, resulting in an overall packaged volume of approximately 2500 cubic metres containing the estimated 460 cubic metres of the reactor structure. (authors)

Halliwell, Chris [Sellafield Ltd, Sellafield (United Kingdom)

2012-07-01T23:59:59.000Z

307

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

SciTech Connect (OSTI)

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

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

1994-04-01T23:59:59.000Z

308

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

Science Journals Connector (OSTI)

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

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

309

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

Science Journals Connector (OSTI)

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

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

310

Simulator platform for fast reactor operation and safety technology demonstration  

SciTech Connect (OSTI)

A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J. (Nuclear Engineering Division)

2012-07-30T23:59:59.000Z

311

Concepts and Tests for the Remote-Controlled Dismantling of the Biological Shield and Form work of the KNK Reactor - 13425  

SciTech Connect (OSTI)

The compact sodium-cooled nuclear reactor facility Karlsruhe (KNK), a prototype Fast Breeder, is currently in an advanced stage of dismantling. Complete dismantling is based on 10 partial licensing steps. In the frame of the 9. decommissioning permit, which is currently ongoing, the dismantling of the biological shield is foreseen. The biological shield consists of heavy reinforced concrete with built-in steel fitments, such as form-work of the reactor tank, pipe sleeves, ventilation channels, and measuring devices. Due to the activation of the inner part of the biological shield, dismantling has to be done remote-controlled. During a comprehensive basic design phase a practical dismantling strategy was developed. Necessary equipment and tools were defined. Preliminary tests revealed that hot wire plasma cutting is the most favorable cutting technology due to the geometrical boundary conditions, the varying distance between cutter and material, and the heavy concrete behind the steel form-work. The cutting devices will be operated remotely via a carrier system with an industrial manipulator. The carrier system has expandable claws to adjust to the varying diameter of the reactor shaft during dismantling progress. For design approval of this prototype development, interaction between manipulator and hot wire plasma cutting was tested in a real configuration. For the demolition of the concrete structure, an excavator with appropriate tools, such as a hydraulic hammer, was selected. Other mechanical cutting devices, such as a grinder or rope saw, were eliminated because of concrete containing steel spheres added to increase the shielding factor of the heavy concrete. Dismantling of the biological shield will be done in a ring-wise manner due to static reasons. During the demolition process, the excavator is positioned on its tripod in three concrete recesses made prior to the dismantling of the separate concrete rings. The excavator and the manipulator carrier system will be operated alternately. Main boundary condition for all the newly designed equipment is the decommissioning housing of limited space within the reactor building containment. To allow for a continuous removal of the concrete rubble, an additional opening on the lowest level of the reactor shaft will be made. All equipment and the interaction of the tools have to be tested before use in the controlled area. Therefore a full-scale model of the biological shield will be provided in a mock-up. The tests will be performed in early 2014. The dismantling of the biological shield is scheduled for 2015. (authors)

Neff, Sylvia; Graf, Anja; Petrick, Holger; Rothschmitt, Stefan [WAK Rueckbau- und Entsorgungs- GmbH, P.O.Box 12 63, 76339 Eggenstein- Leopoldshafen (Germany)] [WAK Rueckbau- und Entsorgungs- GmbH, P.O.Box 12 63, 76339 Eggenstein- Leopoldshafen (Germany); Klute, Stefan [Siempelkamp Nukleartechnik GmbH, Am Taubenfeld 25/1, 69123 Heidelberg (Germany)] [Siempelkamp Nukleartechnik GmbH, Am Taubenfeld 25/1, 69123 Heidelberg (Germany); Stanke, Dieter [Siempelkamp NIS Ingenieurgesellschaft mbH, Industriestrasse 13, 63755 Alzenau (Germany)] [Siempelkamp NIS Ingenieurgesellschaft mbH, Industriestrasse 13, 63755 Alzenau (Germany)

2013-07-01T23:59:59.000Z

312

Dysprosium as a resonance absorber and its effect on the coolant void reactivity in Advanced Heavy Water Reactor (AHWR)  

Science Journals Connector (OSTI)

Dysprosium has been used as a slow neutron absorber in the fuel assembly of Advanced Heavy Water Reactor (AHWR) to achieve a negative coolant void reactivity. Dysprosium as occurring in nature has as many as seven isotopes namely, 156Dy, 158Dy, 160Dy, 161Dy, 162Dy, 163Dy, and 164Dy. Of these, the isotope 164Dy has the largest absorption cross section for thermal neutrons. In the past, nuclear data libraries used in our studies have considered only 164Dy isotope and this was sufficient for performing foil activation studies of Dy. The other isotopes of Dy have significant resonances and could affect the design. The treatment of the dysprosium isotopes with resonance tabulations is required for a more accurate estimation of the lattice characteristics like the coolant void reactivity. The use of resonance tabulations for the dysprosium isotopes and its effect on the coolant void reactivity of AHWR fuel cluster has been studied in this paper. Also, the treatment of the stand-alone structural rod with dysprosium as burnable absorber having resonance tabulations has been done for the first time.

Umasankari Kannan; S. Ganesan

2010-01-01T23:59:59.000Z

313

2011 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond  

SciTech Connect (OSTI)

This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond from November 1, 2010 through October 31, 2011. The report contains the following information: Facility and system description Permit required effluent monitoring data and loading rates Groundwater monitoring data Status of compliance activities Noncompliance and other issues Discussion of the facility's environmental impacts During the 2011 permit year, approximately 166 million gallons of wastewater were discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

Mike Lewis

2012-02-01T23:59:59.000Z

314

The Conception of Thermonuclear Reactor on the Principle of Gravitational Confinement of Dense High-temperature Plasma  

E-Print Network [OSTI]

The work of Fisenko S. I., & Fisenko I. S. (2009). The old and new concepts of physics, 6 (4), 495, shows the key fact of the existence of gravitational radiation as a radiation of the same level as electromagnetic. The obtained results strictly correspond to the framework of relativistic theory of gravitation and quantum mechanics. The given work contributes into further elaboration of the findings considering their application to dense high-temperature plasma of multiple-charge ions. This is due to quantitative character of electron gravitational emission spectrum such that amplification of gravitational emission may take place only in multiple-charge ion high-temperature plasma.

Stanislav Fisenko; Igor Fisenko

2010-06-27T23:59:59.000Z

315

A FEASIBILITY AND OPTIMIZATION STUDY TO DETERMINE COOLING TIME AND BURNUP OF ADVANCED TEST REACTOR FUELS USING A NONDESTRUCTIVE TECHNIQUE  

SciTech Connect (OSTI)

The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method for ATR applications the technique was tested using one-isotope, multi-isotope and fuel simulated sources. Burnup calibrations were perfomed using convoluted and deconvoluted data. The calibrations results showed burnup prediction by this method improves using deconvolution. The final stage of the deconvolution method development was to perform an irradiation experiment in order to create a surrogate fuel source to test the deconvolution method using experimental data. A conceptual design of the fuel scan system is path forward using the rugged LaBr3 detector in an above the water configuration and deconvolution algorithms.

Jorge Navarro

2013-12-01T23:59:59.000Z

316

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

SciTech Connect (OSTI)

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

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

2013-07-01T23:59:59.000Z

317

Application of USNRC NUREG/CR-6661 and draft DG-1108 to evolutionary and advanced reactor designs  

SciTech Connect (OSTI)

For the seismic design of evolutionary and advanced nuclear reactor power plants, there are definite financial advantages in the application of USNRC NUREG/CR-6661 and draft Regulatory Guide DG-1108. NUREG/CR-6661, 'Benchmark Program for the Evaluation of Methods to Analyze Non-Classically Damped Coupled Systems', was by Brookhaven National Laboratory (BNL) for the USNRC, and Draft Regulatory Guide DG-1108 is the proposed revision to the current Regulatory Guide (RG) 1.92, Revision 1, 'Combining Modal Responses and Spatial Components in Seismic Response Analysis'. The draft Regulatory Guide DG-1108 is available at http://members.cox.net/apolloconsulting, which also provides a link to the USNRC ADAMS site to search for NUREG/CR-6661 in text file or image file. The draft Regulatory Guide DG-1108 removes unnecessary conservatism in the modal combinations for closely spaced modes in seismic response spectrum analysis. Its application will be very helpful in coupled seismic analysis for structures and heavy equipment to reduce seismic responses and in piping system seismic design. In the NUREG/CR-6661 benchmark program, which investigated coupled seismic analysis of structures and equipment or piping systems with different damping values, three of the four participants applied the complex mode solution method to handle different damping values for structures, equipment, and piping systems. The fourth participant applied the classical normal mode method with equivalent weighted damping values to handle differences in structural, equipment, and piping system damping values. Coupled analysis will reduce the equipment responses when equipment, or piping system and structure are in or close to resonance. However, this reduction in responses occurs only if the realistic DG-1108 modal response combination method is applied, because closely spaced modes will be produced when structure and equipment or piping systems are in or close to resonance. Otherwise, the conservatism in the current Regulatory Guide 1.92, Revision 1, will overshadow the advantage of coupled analysis. All four participants applied the realistic modal combination method of DG-1108. Consequently, more realistic and reduced responses were obtained. (authors)

Chang 'Apollo', Chen [Apollo Consulting, Inc., Surprise, AZ 85374-4605 (United States)

2006-07-01T23:59:59.000Z

318

Sixteenth water reactor safety information meeting: Proceedings: Volume 5, NUREG-1150, accident managment, recent advances in severe accident research, TMI-2, BWR Mark l shell failure  

SciTech Connect (OSTI)

This five-volume report contains 141 papers out of the 175 that were presented at the Sixteenth Water Reactor Safety Information Meeting held at the National Institute of Standards and Technology, Gaithersburg, Maryland, during the week of October 24--27, 1988. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included twenty different papers presented by researchers from Germany, Italy, Japan, Sweden, Switzerland, Taiwan and the United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This document, Volume 5, discusses NUREG-1150, Accident Management, Recent Advances in Severe Accident Research, BWR Mark I Shell Failure, and the Three Mile Island-2 Reactor.

Weiss, A.J. (comp.)

1989-03-01T23:59:59.000Z

319

Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information  

SciTech Connect (OSTI)

Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

M. Chen; CM Regan; D. Noe

2006-01-09T23:59:59.000Z

320

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

Science Journals Connector (OSTI)

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

H. E. Knee

1991-01-01T23:59:59.000Z

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

Fast Spectrum Molten Salt Reactor Options  

SciTech Connect (OSTI)

During 2010, fast-spectrum molten-salt reactors (FS-MSRs) were selected as a transformational reactor concept for light-water reactor (LWR)-derived heavy actinide disposition by the Department of Energy-Nuclear Energy Advanced Reactor Concepts (ARC) program and were the subject of a preliminary scoping investigation. Much of the reactor description information presented in this report derives from the preliminary studies performed for the ARC project. This report, however, has a somewhat broader scope-providing a conceptual overview of the characteristics and design options for FS-MSRs. It does not present in-depth evaluation of any FS-MSR particular characteristic, but instead provides an overview of all of the major reactor system technologies and characteristics, including the technology developments since the end of major molten salt reactor (MSR) development efforts in the 1970s. This report first presents a historical overview of the FS-MSR technology and describes the innovative characteristics of an FS-MSR. Next, it provides an overview of possible reactor configurations. The following design features/options and performance considerations are described including: (1) reactor salt options-both chloride and fluoride salts; (2) the impact of changing the carrier salt and actinide concentration on conversion ratio; (3) the conversion ratio; (4) an overview of the fuel salt chemical processing; (5) potential power cycles and hydrogen production options; and (6) overview of the performance characteristics of FS-MSRs, including general comparative metrics with LWRs. The conceptual-level evaluation includes resource sustainability, proliferation resistance, economics, and safety. The report concludes with a description of the work necessary to begin more detailed evaluation of FS-MSRs as a realistic reactor and fuel cycle option.

Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Patton, Bruce W [ORNL; Howard, Rob L [ORNL; Harrison, Thomas J [ORNL

2011-07-01T23:59:59.000Z

322

Development of a Simulation Model and Safety Evaluation for a Depressurization Accident Without Reactor Scram in an Advanced HTGR  

SciTech Connect (OSTI)

It is important to use analyses to prove outstanding inherent reactor safety during a severe accident in order to convince the public and licensing authority of the safety advantage of the high-temperature gas-cooled reactor (HTGR). In this study, the simulation of a depressurization accident without reactor scram (DAWS) was performed for a future HTGR with 450-MW thermal output, introducing the annular core of pin-in-block-type fuel, which was originally designed in Japan. The DAWS has the possibility of becoming one of the severe accidents postulated in the HTGR. To perform an accurate simulation, a new analytical model for reactor dynamics and indirect decay heat removal from the surface of the reactor pressure vessel (RPV) during the DAWS was developed. The features of the new simulation model are as follows:1. A single-channel model is coupled with a two-dimensional reactor thermal model in the new simulation model. The reactor kinetics with a single-channel model during the DAWS is simulated taking into account heat removal from the reactor calculated in the R-Z reactor thermal model, including the RPV and indirect vessel cooling system. No conventional calculation codes with a single channel have a heat removal model from an RPV or were able to simulate precisely the transient during DAWS.2. A xenon buildup and decay model for the reactivity calculation is made in addition to one point-kinetics approximation to simulate a recriticality and a power oscillation following the initiation of the DAWS.3. A transient simulation can be performed for two kinds of core models of pin-in-block- and multihole-type fuels.The accurate evaluation of xenon density and core temperature is of prime importance in the simulation of the DAWS. From the simulation result with a proper safety margin, it was confirmed that the safety performance of passive decay heat removal with cooling indirectly from the surface of the RPV is outstanding for the DAWS, and a severe-accident-free HTGR can be designed. The newly developed code is applicable to the detailed safety evaluation necessary to future HTGR design.

Nakagawa, Shigeaki; Saikusa, Akio; Kunitomi, Kazuhiko [Japan Atomic Energy Research Institute (Japan)

2001-02-15T23:59:59.000Z

323

TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis  

SciTech Connect (OSTI)

The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients.

Liles, D.R.; Mahaffy, J.H.

1986-07-01T23:59:59.000Z

324

Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports  

SciTech Connect (OSTI)

This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A & 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met.

Not Available

1994-01-15T23:59:59.000Z

325

Energy Department Announces New Investments in Advanced Nuclear...  

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

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

326

Small Modular Reactors Presentation to Secretary of Energy Advisory Board -  

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

Small Modular Reactors Presentation to Secretary of Energy Advisory Small Modular Reactors Presentation to Secretary of Energy Advisory Board - Deputy Assistant Secretary John Kelly Small Modular Reactors Presentation to Secretary of Energy Advisory Board - Deputy Assistant Secretary John Kelly DOE Small Modular Reactor Program (SMR) Research, Development & Deployment (RD&D) to enable the deployment of a fleet of SMRs in the United States SMR Program is a new program for FY 2011 Structured to address the need to enable the deployment of mature, near-term SMR designs based on known LWR technology Conduct needed R&D activities to advance the understanding and demonstration of innovative reactor technologies and concepts John_Kelly-SEAB_SMRBriefing_July20_2011_final.pdf More Documents & Publications Meeting Materials: June 12, 2012

327

LIGHT WATER REACTOR SUSTAINABILITY PROGRAM ADVANCED INSTRUMENTATION, INFORMATION, AND CONTROL SYSTEMS TECHNOLOGIES TECHNICAL PROGRAM PLAN FOR 2013  

SciTech Connect (OSTI)

Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

Bruce Hallbert; Ken Thomas

2014-07-01T23:59:59.000Z

328

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

SciTech Connect (OSTI)

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

Not Available

1980-11-14T23:59:59.000Z

329

Advanced Fuel Cycle Program  

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

Working with INL Community Outreach Visitor Information Calendar of Events ATR National Scientific User Facility Center for Advanced Energy Studies Light Water Reactor...

330

Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the Advanced Neutron Source reactor at Oak Ridge National Laboratory  

SciTech Connect (OSTI)

This paper describes modeling and analysis to evaluate the extent of core damage during flow blockage events in the Advanced Neutron Source (ANS) reactor planned to be built at ORNL. Damage propagation is postulated to occur from thermal conduction between dmaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur beause of fuel plate swelling during fission product vapor release or plate buckling. Complex phenomena of damage propagation were modeled using a one-dimensional heat transfer model. A parametric study was done for several uncertain variables. The study included investigating effects of plate contact area, convective heat transfer coefficient, thermal conductivity on fuel swelling, and initial temperature of the plate being contacted by the damaged plate. Also, the side support plates were modeled to account for their effects of damage propagation. Results provide useful insights into how variouss uncertain parameters affect damage propagation.

Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V.

1995-12-31T23:59:59.000Z

331

Risk Management for Sodium Fast Reactors.  

SciTech Connect (OSTI)

Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.

Denman, Matthew R; Groth, Katrina; Cardoni, Jeffrey N; Wheeler, Timothy A.

2015-01-01T23:59:59.000Z

332

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

SciTech Connect (OSTI)

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

Not Available

1980-04-18T23:59:59.000Z

333

DUAL-MODE PROPULSION SYSTEM ENABLING CUBESAT EXPLORATION OF THE SOLAR SYSTEM NASA Innovative Advanced Concepts (NIAC) Phase I Final Report  

SciTech Connect (OSTI)

It is apparent the cost of planetary exploration is rising as mission budgets declining. Currently small scientific beds geared to performing limited tasks are being developed and launched into low earth orbit (LEO) in the form of small-scale satellite units, i.e., CubeSats. These micro- and nano-satellites are gaining popularity among the university and science communities due to their relatively low cost and design flexibility. To date these small units have been limited to performing tasks in LEO utilizing solar-based power. If a reasonable propulsion system could be developed, these CubeSat platforms could perform exploration of various extra-terrestrial bodies within the solar system engaging a broader range of researchers. Additionally, being mindful of mass, smaller cheaper launch vehicles (~1,000 kgs to LEO) can be targeted. This, in effect, allows for beneficial explora-tion to be conducted within limited budgets. Researchers at the Center for Space Nuclear Re-search (CSNR) are proposing a low mass, radioisotope-based, dual-mode propulsion system capable of extending the exploration realm of these CubeSats out of LEO. The proposed radioisotope-based system would leverage the high specific energies [J/kg] associated with radioisotope materials and enhance their inherent low specific powers [W/g]. This is accomplished by accumulating thermal energy from nuclear decay within a central core over time. This allows for significant amounts of power to be transferred to a flowing gas over short periods of time. In the proposed configuration the stored energy can be utilized in two ways: (1) with direct propellant injection to the core, the energy can be converted into thrust through the use of a converging-diverging nozzle and (2) by flowing a working fluid through the core and subsequent Brayton engine, energy within the core can be converted to electrical energy. The first scenario achieves moderate ranges of thrust, but at a higher Isp than traditional chemical-based systems. The second scenario allows for the production of electrical power, which is then available for electric-based propulsion. Additionally, once at location the production of electrical power can be dedicated to the payloads communication system for data transfer. Ultimately, the proposed dual-mode propulsion platform capitalizes on the benefits of two types of propulsion methods the thrust of thermal propulsion ideal for quick orbital maneuvers and the specific impulse of electric propulsion ideal for efficient inter-planetary travel. Previous versions of this RTR-based concept have been studied for various applications [NETS 1-3]. The current version of this concept is being matured through a NASA Innovative Advanced Concepts (NIAC) Phase I grant, awarded for FY 2014. In this study the RTR concept is being developed to deliver a 6U CubeSat payload to the orbit of the Saturnian moon - Enceladus. Additionally, this study will develop an entire mission architecture for Enceladus targeting a total allowable launch mass of 1,000 kg.

Nathan Jerred; Troy Howe; Adarsh Rajguru; Dr. Steven Howe

2014-06-01T23:59:59.000Z

334

Advanced direct coal liquefaction concepts  

SciTech Connect (OSTI)

During the first quarter of FY 1993, the Project proceeded close to the Project Plan. The analysis of the feed material has been completed as far as possible. Some unplanned distillation was needed to correct the boiling range of the Black Thunder solvent used during the autoclave tests. Additional distillation will be required if the same solvent is to be used for the bench unit tests. A decision on this is still outstanding. The solvent to be used with Illinois No. 6 coal has not yet been defined. As a result, the procurement of the feed and the feed analysis is somewhat behind schedule. Agglomeration tests with Black Thunder coal indicates that small agglomerates can be formed. However, the ash removal is quite low (about 10%), which is not surprising in view of the low ash content of the coal. The first series of autoclave tests with Black Thunder coal was completed as planned. Also, additional runs are in progress as repeats of previous runs or at different operating conditions based on the data obtained so far. The results are promising indicating that almost complete solubilization (close to 90%) of Black Thunder coal can be achieved in a CO/H[sub 2]O environment at our anticipated process conditions. The design of the bench unit has been completed. In contrast to the originally planned modifications, the bench unit is now designed based on a computerized control and data acquisition system. All major items of equipment have been received, and prefabrication of assemblies and control panels is proceeding on schedule. Despite a slight delay in the erection of the structural steel, it is anticipated that the bench unit will be operational at the beginning of April 1993.

Berger, D.J.; Parker, R.J.; Simpson, P.L. (Canadian Energy Development, Inc., Edmonton, AB (Canada))

1992-01-01T23:59:59.000Z

335

Methods for quantifying uncertainty in fast reactor analyses.  

SciTech Connect (OSTI)

Liquid-metal-cooled fast reactors in the form of sodium-cooled fast reactors have been successfully built and tested in the U.S. and throughout the world. However, no fast reactor has operated in the U.S. for nearly fourteen years. More importantly, the U.S. has not constructed a fast reactor in nearly 30 years. In addition to reestablishing the necessary industrial infrastructure, the development, testing, and licensing of a new, advanced fast reactor concept will likely require a significant base technology program that will rely more heavily on modeling and simulation than has been done in the past. The ability to quantify uncertainty in modeling and simulations will be an important part of any experimental program and can provide added confidence that established design limits and safety margins are appropriate. In addition, there is an increasing demand from the nuclear industry for best-estimate analysis methods to provide confidence bounds along with their results. The ability to quantify uncertainty will be an important component of modeling that is used to support design, testing, and experimental programs. Three avenues of UQ investigation are proposed. Two relatively new approaches are described which can be directly coupled to simulation codes currently being developed under the Advanced Simulation and Modeling program within the Reactor Campaign. A third approach, based on robust Monte Carlo methods, can be used in conjunction with existing reactor analysis codes as a means of verification and validation of the more detailed approaches.

Fanning, T. H.; Fischer, P. F.

2008-04-07T23:59:59.000Z

336

Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor  

SciTech Connect (OSTI)

The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.

S.T. Revankar; W. Zhou; Gavin Henderson

2008-07-08T23:59:59.000Z

337

Prometheus Hot Leg Piping Concept  

SciTech Connect (OSTI)

The Naval Reactors Prime Contractor Team (NRPCT) recommended the development of a gas cooled reactor directly coupled to a Brayton energy conversion system as the Space Nuclear Power Plant (SNPP) for NASA's Project Prometheus. The section of piping between the reactor outlet and turbine inlet, designated as the hot leg piping, required unique design features to allow the use of a nickel superalloy rather than a refractory metal as the pressure boundary. The NRPCT evaluated a variety of hot leg piping concepts for performance relative to SNPP system parameters, manufacturability, material considerations, and comparison to past high temperature gas reactor (HTGR) practice. Manufacturability challenges and the impact of pressure drop and turbine entrance temperature reduction on cycle efficiency were discriminators between the piping concepts. This paper summarizes the NRPCT hot leg piping evaluation, presents the concept recommended, and summarizes developmental issues for the recommended concept.

Gribik, Anastasia M. [Bechtel Bettis, Inc., Bettis Atomic Power Laboratory, West Mifflin, PA 15122 (United States); DiLorenzo, Peter A. [KAPL, Inc., Knolls Atomic Power Laboratory, Schenectady, NY 12301 (United States)

2007-01-30T23:59:59.000Z

338

Towards a desalination initiative using cogeneration with an advanced reactor type and uranium recovered from Moroccan phosphoric acid production  

Science Journals Connector (OSTI)

Morocco is known to be among the first few countries to produce phosphate and phosphoric acid. Moroccan phosphate contains substantial amounts of uranium. This uranium can be recovered from the phosphate ore as a by-product during the production of phosphoric acid. Uranium extraction processes linked with phosphoric acid fabrication have been used industrially in some countries. This is done mainly by solvent extraction. Although, the present price of uranium is low in the international market, such uranium recovery could be considered as a side product of phosphoric acid production. The price of uranium has a very small impact on the cost of nuclear energy obtained from it. This paper focuses on the extraction of uranium salt from phosphate rock. If uranium is recovered in Morocco in the proposed manner, it could serve as feed for a number of nuclear power plants. The natural uranium product would have to be either enriched or blended as mixed-oxide fuel to manufacture adequate nuclear fuel. Part of this fuel would feed a desalination initiative using a high temperature reactor of the new generation, chosen for its intrinsic safety, sturdiness, ease of maintenance, thermodynamic characteristics and long fuel life between reloads, that is, good economy. ?n international cooperation based on commercial contract schemes would concern: the general project and uranium extraction; uranium enrichment and fuel fabrication services; the nuclear power plant; and the desalination plant. This paper presents the overall feasibility of the general project with some quantitative preliminary figures and cost estimates.

Michel Lung; Abdelaali Kossir; Driss Msatef

2005-01-01T23:59:59.000Z

339

Use of phenomena identification and ranking (PIRT) process in research related to design certification of the AP600 advanced passive light water reactor (LWR)  

SciTech Connect (OSTI)

The AP600 LWR is a new advanced passive design that has been submitted to the USNRC for design certification. Within the certification process the USNRC will perform selected system thermal hydraulic response audit studies to help confirm parts of the vendor`s safety analysis submittal. Because of certain innovative design features of the safety systems, new experimental data and related advances in the system thermal hydraulic analysis computer code are being developed by the USNRC. The PIRT process is being used to focus the experimental and analytical work to obtain a sufficient and cost effective research effort. The objective of this paper is to describe the application and most significant results of the PIRT process, including several innovative features needed in the application to accommodate the short design certification schedule. The short design certification schedule has required that many aspects of the USNRC experimental and analytical research be performed in parallel, rather than in series as was normal for currently operating LWRS. This has required development and use of management techniques that focus and integrate the various diverse parts of the research. The original PIRTs were based on inexact knowledge of an evolving reactor design, and concentrated on the new passive features of the design. Subsequently, the PIRTs have evolved in two more stages as the design became more firm and experimental and analytical data became available. A fourth and final stage is planned and in progress to complete the PIRT development. The PIRTs existing at the end of each development stage have been used to guide the experimental program, scaling analyses and code development supporting the audit studies.

Wilson, G.E.; Fletcher, C.D. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Eltawila, F. [Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research

1996-07-01T23:59:59.000Z

340

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

Science Journals Connector (OSTI)

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

L.A. Macedo; W.M. Torres

2011-01-01T23:59:59.000Z

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341

Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the Advanced Neutron Source Reactor at Oak Ridge National Laboratory  

SciTech Connect (OSTI)

This paper describes modeling and analysis to evaluate the extent of core damage during flow blockage events in the Advanced Neutron Source (ANS) reactor planned to be built at the Oak Ridge National Laboratory (ORNL). Damage propagation is postulated to occur from thermal conduction between damaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur because of fuel plate swelling during fission product vapor release or plate buckling. Complex phenomena of damage propagation were modeled using a one-dimensional heat transfer model. A scoping study was conducted to learn what parameters are important for core damage propagation, and to obtain initial estimates of core melt mass for addressing recriticality and steam explosion events. The study included investigating the effects of the plate contact area, the convective heat transfer coefficient, thermal conductivity upon fuel swelling, and the initial temperature of the plate being contacted by the damaged plate. Also, the side support plates were modeled to account for their effects on damage propagation. The results provide useful insights into how various uncertain parameters affect damage propagation.

Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V.

1995-09-01T23:59:59.000Z

342

2010 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Sites Advanced Test Reactor Complex Cold Waste Pond  

SciTech Connect (OSTI)

This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (#LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Sites Advanced Test Reactor Complex Cold Waste Pond from November 1, 2009 through October 31, 2010. The report contains the following information: Facility and system description Permit required effluent monitoring data and loading rates Groundwater monitoring data Status of compliance activities Discussion of the facilitys environmental impacts During the 2010 permit year, approximately 164 million gallons of wastewater were discharged to the Cold Waste Pond. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

mike lewis

2011-02-01T23:59:59.000Z

343

2013 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Sites Advanced Test Reactor Complex Cold Waste Pond  

SciTech Connect (OSTI)

This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (#LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Sites Advanced Test Reactor Complex Cold Waste Pond from November 1, 2012October 31, 2013. The report contains the following information: Facility and system description Permit required effluent monitoring data and loading rates Groundwater monitoring data Status of compliance activities Noncompliance issues Discussion of the facilitys environmental impacts. During the 2013 permit year, approximately 238 million gallons of wastewater was discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters are below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

Mike Lewis

2014-02-01T23:59:59.000Z

344

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

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

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

345

Chapter 1 - Reactor configurations and design parameters for thermochemical conversion of biomass into fuels, energy, and chemicals  

Science Journals Connector (OSTI)

Abstract This chapter describes reactors for thermochemical conversion of lignocellulosic biomass into fuels, energy, and chemicals. The chapter covers basic definitions and concepts involved in biofuels and thermochemical conversion of biomass, and it also includes more advanced topics such as the main reactor configurations currently in use for thermochemical technologies, important parameters for reactor design, discussion of how parameters affect reactor performance, and several examples and case studies. The focus is on fast pyrolysis and gasification systems. The topics discussed include energy and carbon efficiencies, convenience of operation and scale-up, and several other parameters related to reactor design. After reading this chapter, the reader will understand the main characteristics of reactors for thermochemical conversion of biomass, their strengths, and their weaknesses for specific applications.

Fernando L.P. Resende

2014-01-01T23:59:59.000Z

346

ITP Metal Casting: Advanced Melting Technologies: Energy Saving...  

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

Advanced Melting Technologies: Energy Saving Concepts and Opportunities for the Metal Casting Industry ITP Metal Casting: Advanced Melting Technologies: Energy Saving Concepts and...

347

Nuclear Reactor (atomic reactor)  

Science Journals Connector (OSTI)

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

2008-01-01T23:59:59.000Z

348

Three-dimensional imaging and precision metrology for liquid-salt-cooled reactors  

SciTech Connect (OSTI)

The liquid-salt-cooled very high temperature reactor, also called the Advanced High-Temperature Reactor (AHTR), is a new large high-temperature reactor concept that combines in a novel way four established technologies: (1) coated-particle graphite-matrix nuclear fuels, (2) Brayton power cycles, (3) passive safety systems and plant designs previously developed for liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants. The AHTR will require refueling, in-service inspection, and maintenance (RIM) with supporting instrumentation systems. The fluoride salts that are being evaluated as potential reactor coolants have melting points between 350 and 500 deg. C, values that imply minimum RIM temperatures between 400 and 550 deg. C. These salts are transparent over a wider range of the light spectrum than is water. The high temperatures, the optical characteristics of the coolant, and advances in metrology may enable the use of lasers to create three-dimensional images of the reactor interior to assist refueling, monitor vibrations in components, map fluid flow, and enable inspections of internal reactor components. A description of the reactor and an initial evaluation of the use of optical techniques for AHTR instrumentation are provided. (authors)

Forsberg, C. W. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6165 (United States); Varma, V. K.; Burgess, T. W. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6304 (United States)

2006-07-01T23:59:59.000Z

349

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

SciTech Connect (OSTI)

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

Buden, D.

1993-08-11T23:59:59.000Z

350

Development of a Technical Basis and Guidance for Advanced SMR Function Allocation  

SciTech Connect (OSTI)

This report presents the results from three key activities for FY13 that influence the definition of new concepts of operations for advanced Small Modular Reactors (AdvSMR: a) the development of a framework for the analysis of the functional environmental, and structural attributes, b) the effect that new technologies and operational concepts would have on the way functions are allocated to humans or machines or combinations of the two, and c) the relationship between new concepts of operations, new function allocations, and human performance requirements.

Jacques Hugo; David Gertman; Jeffrey Joe; Ronal Farris; April Whaley; Heather Medema

2013-09-01T23:59:59.000Z

351

The Path to Fusion Energy for Concepts Currently at the Concept Exploration Level  

E-Print Network [OSTI]

confinement concepts · Spheromak · FRC Concepts operating in a very different parameter space · Magnetized") · Electrostatic confinement · Others #12;EBH 1/13/03 Development plan for toroidal confinement ­ spheromak Status of spheromaks This wide range of reactor visions offers opportunities for a better reactor, but the development

352

Nuclear Data Measurements for 21st Century Reactor Physics Applications  

SciTech Connect (OSTI)

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

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

2003-03-01T23:59:59.000Z

353

Tribal Renewable Energy Advanced Course: Project Development...  

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

Concepts Tribal Renewable Energy Advanced Course: Project Development Concepts Watch the DOE Office of Indian Energy renewable energy course entitled "Tribal Renewable Energy...

354

Combustion Model for Engine Concept Development | Department...  

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

Combustion Model for Engine Concept Development Presentation shows how 1-cylinder testing, 3D combustion CFD and 1D gas exchange with an advanced combustion model are used...

355

Refueling Liquid-Salt-Cooled Very High-Temperature Reactors  

SciTech Connect (OSTI)

The liquid-salt-cooled very high-temperature reactor (LS-VHTR), also called the Advanced High-Temperature Reactor (AHTR), is a new reactor concept that combines in a novel way four established technologies: (1) coated-particle graphite-matrix nuclear fuels, (2) Brayton power cycles, (3) passive safety systems and plant designs previously developed for liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants. Depending upon goals, the peak coolant operating temperatures are between 700 and 1000 deg. C, with reactor outputs between 2400 and 4000 MW(t). Several fluoride salt coolants that are being evaluated have melting points between 350 and 500 deg. C, values that imply minimum refueling temperatures between 400 and 550 deg. C. At operating conditions, the liquid salts are transparent and have physical properties similar to those of water. A series of refueling studies have been initiated to (1) confirm the viability of refueling, (2) define methods for safe rapid refueling, and (3) aid the selection of the preferred AHTR design. Three reactor cores with different fuel element designs (prismatic, pebble bed, and pin-type fuel assembly) are being evaluated. Each is a liquid-salt-cooled variant of a graphite-moderated high-temperature reactor. The refueling studies examined applicable refueling experience from high-temperature reactors (similar fuel element designs) and sodium-cooled fast reactors (similar plant design with liquid coolant, high temperatures, and low pressures). The findings indicate that refueling is viable, and several approaches have been identified. The study results are described in this paper. (authors)

Forsberg, Charles W. [Oak Ridge National Laboratory, P.O. Box 2008 Oak Ridge, TN 37831 (United States); Peterson, Per F. [Nuclear Engineering Department, University of California at Berkeley, 6124a Etcheverry Hall, Berkeley, CA 94720 (United States); Cahalan, James E. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States); Enneking, Jeffrey A. [Areva NP (United States); Phil MacDonald [Consultant, Cedar Hill, TX (United States)

2006-07-01T23:59:59.000Z

356

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

Science Journals Connector (OSTI)

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

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

2010-01-01T23:59:59.000Z

357

Session: CSP Advanced Systems -- Advanced Overview (Presentation)  

SciTech Connect (OSTI)

The project description is: (1) it supports crosscutting activities, e.g. advanced optical materials, that aren't tied to a single CSP technology and (2) it supports the 'incubation' of new concepts in preliminary stages of investigation.

Mehos, M.

2008-04-01T23:59:59.000Z

358

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

SciTech Connect (OSTI)

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

359

Presented by CASL: The Consortium for Advanced  

E-Print Network [OSTI]

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

360

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

E-Print Network [OSTI]

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

Faust, Christophor Randall

1995-01-01T23:59:59.000Z

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

Global Nuclear Energy Partnership Fact Sheet - Develop Advanced Burner  

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

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

362

Advanced atomization concept for CWF burning in small combustors. Phase 2, Quarterly technical progress report No. 3, 1 April 1991--30 June 1991  

SciTech Connect (OSTI)

The present project involves the second phase of research on a new concept in coal-water fuel (CWF) atomization that is applicable to burning in small combustors. It is intended to address the most important problem associated with CWF combustion; i.e., production of small spray droplets in an efficient manner by an atomization device. Phase 1 of this work was successfully completed with the development of an opposed-jet atomizer that met the goals of the first contract. Performance as a function of operating conditions was measured, and the technical feasibility of the device established in the Atlantic Research Atomization Test Facility employing a Malvern Particle Size Analyzer. Testing then proceeded to a combustion stage in a test furnace at a firing rate of 0.5 to 1.5 MMBtu/H.

Heaton, H.; McHale, E.

1991-12-31T23:59:59.000Z

363

Design Concepts for the Cherenkov Telescope Array CTA: An Advanced Facility for Ground-Based High-Energy Gamma-Ray Astronomy  

SciTech Connect (OSTI)

Ground-based gamma-ray astronomy has had a major breakthrough with the impressive results obtained using systems of imaging atmospheric Cherenkov telescopes. Ground-based gamma-ray astronomy has a huge potential in astrophysics, particle physics and cosmology. CTA is an international initiative to build the next generation instrument, with a factor of 5-10 improvement in sensitivity in the 100 GeV-10 TeV range and the extension to energies well below 100 GeV and above 100 TeV. CTA will consist of two arrays (one in the north, one in the south) for full sky coverage and will be operated as open observatory. The design of CTA is based on currently available technology. This document reports on the status and presents the major design concepts of CTA.

Actis, M

2012-04-17T23:59:59.000Z

364

Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors  

SciTech Connect (OSTI)

A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs.

Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

1989-10-01T23:59:59.000Z

365

An Innovative Three-Dimensional Heterogeneous Coarse-Mesh Transport Method for Advanced and Generation IV Reactor Core Analysis and Design  

SciTech Connect (OSTI)

This project has resulted in a highly efficient method that has been shown to provide accurate solutions to a variety of 2D and 3D reactor problems. The goal of this project was to develop (1) an accurate and efficient three-dimensional whole-core neutronics method with the following features: based sollely on transport theory, does not require the use of cross-section homogenization, contains a highly accurate and self-consistent global flux reconstruction procedure, and is applicable to large, heterogeneous reactor models, and to (2) create new numerical benchmark problems for code cross-comparison.

Farzad Rahnema

2009-11-12T23:59:59.000Z

366

MHTGR: New production reactor summary of experience base  

SciTech Connect (OSTI)

Worldwide interest in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) stems from the capability of the system to retain the advanced fuel and thermal performance while providing unparalleled levels of safety. The small power level of the MHTGR and its passive systems give it a margin of safety not attained by other concepts being developed for power generation. This report covers the experience base for the key nuclear system, components, and processes related to the MHTGR-NPR. 9 refs., 39 figs., 9 tabs.

Not Available

1988-03-01T23:59:59.000Z

367

E-Print Network 3.0 - application test reactor Sample Search...  

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

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

368

E-Print Network 3.0 - aquifer sediment reactors Sample Search...  

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

sediment reactors Search Powered by Explorit Topic List Advanced Search Sample search results for: aquifer sediment reactors Page: << < 1 2 3 4 5 > >> 1 Theme 1. Exposure:...

369

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

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

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

370

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

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

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

371

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

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

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

372

Advanced ceramic materials for next-generation nuclear applications  

Science Journals Connector (OSTI)

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

John Marra

2011-01-01T23:59:59.000Z

373

Nuclear Energy Enabling Technologies (NEET) Reactor Materials  

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

Enabling Technologies (NEET) Reactor Materials Enabling Technologies (NEET) Reactor Materials Award Recipient Estimated Award Amount* Award Location Supporting Organizations Project Description University of Nebraska $979,978 Lincoln, NE Massachusetts Institute of Technology (Cambridge, MA), Texas A&M (College Station, TX) Project will explore the development of advanced metal/ceramic composites. These improvements could lead to more efficient production of electricity in advanced reactors. Oak Ridge National Laboratory $849,000 Oak Ridge, TN University of Wisconsin-Madison (Madison, WI) Project will develop novel high-temperature high-strength steels with the help of computational modeling, which could lead to increased efficiency in advanced reactors. Pacific Northwest National Laboratory

374

Reactor and Nuclear Systems Division (RNSD)  

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

RNSD Home RNSD Home Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation Transport Reactor Physics Thermal Hydraulics & Irradiation Engineering Used Fuel Systems Staff Details (CV/Bios) Publications Org Chart Contact Us ORNL Staff Only Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation Transport Reactor Physics Thermal Hydraulics & Irradiation Engineering Used Fuel Systems Reactor and Nuclear Systems Division News Highlights U.S. Rep. Fleischmann touts ORNL as national energy treasure Martin Peng wins Fusion Power Associates Leadership Award

375

High Efficiency Solar Fuels Reactor Concept  

Broader source: Energy.gov [DOE]

This presentation was delivered at the SunShot Concentrating Solar Power (CSP) Program Review 2013, held April 2325, 2013 near Phoenix, Arizona.

376

Production of Advanced Biofuels via Liquefaction Hydrothermal...  

Office of Scientific and Technical Information (OSTI)

Laboratory Production of Advanced Biofuels via Liquefaction Golden, Colorado April 5, 2013 REPORT 30352.0001 HYDROTHERMAL LIQUEFACTION REACTOR DESIGN REPORT TABLE OF CONTENTS...

377

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

E-Print Network [OSTI]

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

Hejzlar, P.

378

Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor  

SciTech Connect (OSTI)

The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a standard, UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge burnup level, while retaining its inherent safety characteristics. Using generic pebble bed reactor cores, this task will perform physics calculations to evaluate the capabilities of the pebble bed reactor to perform utilization and destruction of LWR used-fuel transuranics. The task will use established benchmarked models, and will introduce modeling advancements appropriate to the nature of the fuel considered (high TRU content and high burn-up).

B. Boer; A. M. Ougouag

2010-09-01T23:59:59.000Z

379

ThermalHydraulic Analysis and Optimization of HCPB Breeder Unit for Future Fusion Reactor  

Science Journals Connector (OSTI)

China has proposed several concepts of breeder blanket for future fusion reactor, one of which is the helium-cooled...

Songsong Qi; Guodong Wang; Changqi Chen; Hongjun Tang

2014-09-01T23:59:59.000Z

380

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

SciTech Connect (OSTI)

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

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

2010-11-15T23:59:59.000Z

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

Heavy Liquid Metal Reactor Development - Nuclear Engineering Division  

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

> Heavy Liquid Metal Reactor Development > Heavy Liquid Metal Reactor Development Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics and Fuel Cycle Analysis Nuclear Data Program Advanced Reactor Development Overview Advanced Fast Reactor (AFR) Heavy Liquid Metal Reactor Development Generation IV Nuclear Waste Form and Repository Performance Modeling Nuclear Energy Systems Design and Development Other Capabilities Work with Argonne Contact us For Employees Site Map Help Advanced Reactor Development and Technology Heavy Liquid Metal Reactor Development Bookmark and Share STAR-LM: Simplified, Modular, Small Reactor Featuring Flow-thru Fuel Cartridge STAR-LM: Simplified, Modular, Small Reactor Featuring Flow-thru Fuel Cartridge. Click on image to view larger image. Argonne has traditionally been the foremost institute in the US for

382

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

Science Journals Connector (OSTI)

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

Gnyaz Ablay

2013-01-01T23:59:59.000Z

383

Instrumentation and Control and Human Machine Interface Science and Technology Roadmap in Support of Advanced Reactors and Fuel Programs in the U.S.  

SciTech Connect (OSTI)

The purpose of this paper is to provide an overview of the current status of the Instrumentation, Control and Human Machine Interface (ICHMI) Science and Technology Roadmap (Reference xi) that was developed to address the major challenges in this technical area for the Gen IV and other U.S. Department of Energy (DOE) initiatives that support future deployments of nuclear energy systems. Reliable, capable ICHMI systems will be necessary for the advanced nuclear plants to be economically competitive. ICHMI enables measurement, control, protection, monitoring, and maintenance for processes and components. Through improvements in the technologies and demonstration of their use to facilitate licensing, ICHMI can contribute to the reduction of plant operations and maintenance costs while helping to ensure high plant availability. The impact of ICHMI can be achieved through effective use of the technologies to improve operational efficiency and optimize use of human resources. However, current licensing experience with digital I&C systems has provided lessons learned concerning the difficulties that can be encountered when introducing advanced technologies with expanded capabilities. Thus, in the development of advanced nuclear power designs, it will be important to address both the technical foundations of ICHMI systems and their licensing considerations. The ICHMI roadmap will identify the necessary research, development and demonstration activities that are essential to facilitate necessary technology advancement and resolve outstanding issues.

Miller, Don W.; Arndt, Steven A.; Dudenhoeffer, Donald D.; Hallbert, Bruce P.; Bond, Leonard J.; Holcomb, David E.; Wood, Richard T.; Naser, Joseph A.; O'Hara, John M.; Quinn, Edward L.

2008-06-01T23:59:59.000Z

384

Hot-Gas Filter Testing with a Transport Reactor Gasifier  

SciTech Connect (OSTI)

Today, coal supplies over 55% of the electricity consumed in the United States and will continue to do so well into the next century. One of the technologies being developed for advanced electric power generation is an integrated gasification combined cycle (IGCC) system that converts coal to a combustible gas, cleans the gas of pollutants, and combusts the gas in a gas turbine to generate electricity. The hot exhaust from the gas turbine is used to produce steam to generate more electricity from a steam turbine cycle. The utilization of advanced hot-gas particulate and sulfur control technologies together with the combined power generation cycles make IGCC one of the cleanest and most efficient ways available to generate electric power from coal. One of the strategic objectives for U.S. Department of Energy (DOE) IGCC research and development program is to develop and demonstrate advanced gasifiers and second-generation IGCC systems. Another objective is to develop advanced hot-gas cleanup and trace contaminant control technologies. One of the more recent gasification concepts to be investigated is that of the transport reactor gasifier, which functions as a circulating fluid-bed gasifier while operating in the pneumatic transport regime of solid particle flow. This gasifier concept provides excellent solid-gas contacting of relatively small particles to promote high gasification rates and also provides the highest coal throughput per unit cross-sectional area of any other gasifier, thereby reducing capital cost of the gasification island.

Swanson, M.L.; Hajicek, D.R.

2002-09-18T23:59:59.000Z

385

Actinide burning in the integral fast reactor  

SciTech Connect (OSTI)

During the past few years, Argonne National Laboratory has been developing the integral fast reactor (IFR), an advanced liquid-metal reactor concept. In the IFR, the inherent properties of liquid-metal cooling are combined with a new metallic fuel and a radically different refining process to allow breakthroughs in passive safety, fuel cycle economics, and waste management. A key feature of the IFR concept is its unique pyroprocessing. Pyroprocessing has the potential to radically improve long-term waste management strategies by exploiting the following attributes: 1. Minor actinides accompany plutonium product stream; therefore, actinide recycling occurs naturally. Actinides, the primary source of long-term radiological toxicity, are removed from the waste stream and returned to the reactor for in situ burning, generating useful energy. 2. High-level waste volume from pyroprocessing call be reduced substantially as compared with direct disposal of spent fuel. 3. Decay heat loading in the repository can be reduced by a large factor, especially for the long-term burden. 4. Low-level waste generation is minimal. 5. Troublesome fission products, such as [sup 99]Tc, [sup 129]I, and [sup 14]C, are contained and immobilized. Singly or in combination, the foregoing attributes provide important improvements in long-term waste management in terms of the ease in meeting technical performance requirements (perhaps even the feasibility of demonstrating that technical performance requirements can be met) and perhaps also in ultimate public acceptance. Actinide recycling, if successfully developed, could well help the current repository program by providing an opportunity to enhance capacity utilization and by deferring the need for future repositories. It also represents a viable technical backup option in the event unforeseen difficulties arise in the repository licensing process.

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

1993-01-01T23:59:59.000Z

386

Generic repository design concepts and thermal analysis (FY11).  

SciTech Connect (OSTI)

Reference concepts for geologic disposal of used nuclear fuel and high-level radioactive waste in the U.S. are developed, including geologic settings and engineered barriers. Repository thermal analysis is demonstrated for a range of waste types from projected future, advanced nuclear fuel cycles. The results show significant differences among geologic media considered (clay/shale, crystalline rock, salt), and also that waste package size and waste loading must be limited to meet targeted maximum temperature values. In this study, the UFD R&D Campaign has developed a set of reference geologic disposal concepts for a range of waste types that could potentially be generated in advanced nuclear FCs. A disposal concept consists of three components: waste inventory, geologic setting, and concept of operations. Mature repository concepts have been developed in other countries for disposal of spent LWR fuel and HLW from reprocessing UNF, and these serve as starting points for developing this set. Additional design details and EBS concepts will be considered as the reference disposal concepts evolve. The waste inventory considered in this study includes: (1) direct disposal of SNF from the LWR fleet, including Gen III+ advanced LWRs being developed through the Nuclear Power 2010 Program, operating in a once-through cycle; (2) waste generated from reprocessing of LWR UOX UNF to recover U and Pu, and subsequent direct disposal of used Pu-MOX fuel (also used in LWRs) in a modified-open cycle; and (3) waste generated by continuous recycling of metal fuel from fast reactors operating in a TRU burner configuration, with additional TRU material input supplied from reprocessing of LWR UOX fuel. The geologic setting provides the natural barriers, and establishes the boundary conditions for performance of engineered barriers. The composition and physical properties of the host medium dictate design and construction approaches, and determine hydrologic and thermal responses of the disposal system. Clay/shale, salt, and crystalline rock media are selected as the basis for reference mined geologic disposal concepts in this study, consistent with advanced international repository programs, and previous investigations in the U.S. The U.S. pursued deep geologic disposal programs in crystalline rock, shale, salt, and volcanic rock in the years leading up to the Nuclear Waste Policy Act, or NWPA (Rechard et al. 2011). The 1987 NWPA amendment act focused the U.S. program on unsaturated, volcanic rock at the Yucca Mountain site, culminating in the 2008 license application. Additional work on unsaturated, crystalline rock settings (e.g., volcanic tuff) is not required to support this generic study. Reference disposal concepts are selected for the media listed above and for deep borehole disposal, drawing from recent work in the U.S. and internationally. The main features of the repository concepts are discussed in Section 4.5 and summarized in Table ES-1. Temperature histories at the waste package surface and a specified distance into the host rock are calculated for combinations of waste types and reference disposal concepts, specifying waste package emplacement modes. Target maximum waste package surface temperatures are identified, enabling a sensitivity study to inform the tradeoff between the quantity of waste per disposal package, and decay storage duration, with respect to peak temperature at the waste package surface. For surface storage duration on the order of 100 years or less, waste package sizes for direct disposal of SNF are effectively limited to 4-PWR configurations (or equivalent size and output). Thermal results are summarized, along with recommendations for follow-on work including adding additional reference concepts, verification and uncertainty analysis for thermal calculations, developing descriptions of surface facilities and other system details, and cost estimation to support system-level evaluations.

Howard, Robert (Oak Ridge National Laboratory, Oak Ridge, TN); Dupont, Mark (Savannah River Nuclear Solutions, Aiken, SC); Blink, James A. (Lawrence Livermore National Laboratory, Livermore, CA); Fratoni, Massimiliano (Lawrence Livermore National Laboratory, Livermore, CA); Greenberg, Harris (Lawrence Livermore National Laboratory, Livermore, CA); Carter, Joe (Savannah River Nuclear Solutions, Aiken, SC); Hardin, Ernest L.; Sutton, Mark A. (Lawrence Livermore National Laboratory, Livermore, CA)

2011-08-01T23:59:59.000Z

387

SSTAR: The U.S. Lead-Cooled Fast Reactor (LFR)  

SciTech Connect (OSTI)

It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10-700 MWe) on small or immature grids in developing nations. Most recently, the Global Nuclear Energy Partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the Small Secure Transportable Autonomous Reactor (SSTAR) reactor has been under ongoing development under the U.S. Generation IV Nuclear Energy Systems Initiative. It a system designed to provide energy security to developing nations while incorporating features to achieve nonproliferation aims, anticipating GNEP objectives. This paper presents the motivation for development of internationally deployable nuclear energy systems as well as a summary of one such system, SSTAR, which is the U.S. Generation IV Lead-cooled Fast Reactor system.

Smith, C F; Halsey, W G; Brown, N W; Sienicki, J J; Moisseytsev, A; Wade, D C

2007-09-25T23:59:59.000Z

388

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

SciTech Connect (OSTI)

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

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

2009-03-01T23:59:59.000Z

389

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

SciTech Connect (OSTI)

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

Simos, N.

2011-05-01T23:59:59.000Z

390

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

SciTech Connect (OSTI)

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

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

1993-07-01T23:59:59.000Z

391

Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation  

SciTech Connect (OSTI)

The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri

2005-09-27T23:59:59.000Z

392

Advanced Concepts in Josephson Junction Reflection Amplifiers  

E-Print Network [OSTI]

Low-noise amplification atmicrowave frequencies has become increasingly important for the research related to superconducting qubits and nanoelectromechanical systems. The fundamental limit of added noise by a phase-preserving amplifier is the standard quantum limit, often expressed as noise temperature $T_{q} = \\hbar {\\omega}/2k_{B}$. Towards the goal of the quantum limit, we have developed an amplifier based on intrinsic negative resistance of a selectively damped Josephson junction. Here we present measurement results on previously proposed wide-band microwave amplification and discuss the challenges for improvements on the existing designs. We have also studied flux-pumped metamaterial-based parametric amplifiers, whose operating frequency can be widely tuned by external DC-flux, and demonstrate operation at $2\\omega$ pumping, in contrast to the typical metamaterial amplifiers pumped via signal lines at $\\omega$.

Pasi Lhteenmki; Visa Vesterinen; Juha Hassel; G. S. Paraoanu; Heikki Sepp; Pertti Hakonen

2014-05-20T23:59:59.000Z

393

Advanced Combustion Concepts - Enabling Systems and Solutions...  

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

ins talled and vehicle available for application, emis s ion and fuel economy optimization. 5 G as oline S ys tems | 5172013 | 2013 R obert B os ch LLC and affiliates ....

394

Advanced wet-dry cooling tower concept  

E-Print Network [OSTI]

The purpose of this years' work has been to test and analyze the new dry cooling tower surface previously developed. The model heat transfer test apparatus built last year has been instrumented for temperature, humidity ...

Snyder, Troxell Kimmel

395

Advanced Combustion Concepts - Enabling Systems and Solutions...  

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

III C AR B OB D II Quality and Safety R eliability R obus tness IS O26262 Brand building B rand Identity & -value Image, e.g. Innovation Cost F or entry...

396

Project Profile: Commercial Development of an Advanced Linear...  

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

Commercial Development of an Advanced Linear-Fresnel-Based CSP Concept Project Profile: Commercial Development of an Advanced Linear-Fresnel-Based CSP Concept SkyFuel logo SkyFuel,...

397

Computational Neutronics Methods and Transmutation Performance Analyses for Fast Reactors  

SciTech Connect (OSTI)

The once-through fuel cycle strategy in the United States for the past six decades has resulted in an accumulation of Light Water Reactor (LWR) Spent Nuclear Fuel (SNF). This SNF contains considerable amounts of transuranic (TRU) elements that limit the volumetric capacity of the current planned repository strategy. A possible way of maximizing the volumetric utilization of the repository is to separate the TRU from the LWR SNF through a process such as UREX+1a, and convert it into fuel for a fast-spectrum Advanced Burner Reactor (ABR). The key advantage in this scenario is the assumption that recycling of TRU in the ABR (through pyroprocessing or some other approach), along with a low capture-to-fission probability in the fast reactors high-energy neutron spectrum, can effectively decrease the decay heat and toxicity of the waste being sent to the repository. The decay heat and toxicity reduction can thus minimize the need for multiple repositories. This report summarizes the work performed by the fuel cycle analysis group at the Idaho National Laboratory (INL) to establish the specific technical capability for performing fast reactor fuel cycle analysis and its application to a high-priority ABR concept. The high-priority ABR conceptual design selected is a metallic-fueled, 1000 MWth SuperPRISM (S-PRISM)-based ABR with a conversion ratio of 0.5. Results from the analysis showed excellent agreement with reference values. The independent model was subsequently used to study the effects of excluding curium from the transuranic (TRU) external feed coming from the LWR SNF and recycling the curium produced by the fast reactor itself through pyroprocessing. Current studies to be published this year focus on analyzing the effects of different separation strategies as well as heterogeneous TRU target systems.

R. Ferrer; M. Asgari; S. Bays; B. Forget

2007-03-01T23:59:59.000Z

398

Tritium production analysis and management strategies for a Fluoride-salt-cooled high-temperature test reactor (FHTR)  

E-Print Network [OSTI]

The Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is a test reactor concept that aims to demonstrate the neutronics, thermal-hydraulics, materials, tritium management, and to address other reactor operational ...

Rodriguez, Judy N

2013-01-01T23:59:59.000Z

399

Transport reactor development status  

SciTech Connect (OSTI)

This project is part of METC`s Power Systems Development Facility (PSDF) located at Wilsonville, Alabama. The primary objective of the Advanced Gasifier module is to produce vitiated gases for intermediate-term testing of Particulate Control Devices (PCDs). The Transport reactor potentially allows particle size distribution, solids loading, and particulate characteristics in the off-gas stream to be varied in a number of ways. Particulates in the hot gases from the Transport reactor will be removed in the PCDs. Two PCDs will be initially installed in the module; one a ceramic candle filter, the other a granular bed filter. After testing of the initial PCDs they will be removed and replaced with PCDs supplied by other vendors. A secondary objective is to verify the performance of a Transport reactor for use in advanced Integrated Gasification Combined Cycle (IGCC), Integrated Gasification Fuel Cell (IG-FC), and Pressurized Combustion Combined Cycle (PCCC) power generation units. This paper discusses the development of the Transport reactor design from bench-scale testing through pilot-scale testing to design of the Process Development Unit (PDU-scale) facility at Wilsonville.

Rush, R.E.; Fankhanel, M.O.; Campbell, W.M.

1994-10-01T23:59:59.000Z

400

Research Program of a Super Fast Reactor  

SciTech Connect (OSTI)

Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is not breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie; Terai, Takayuki; Nagasaki, Shinya; Muroya, Yusa; Abe, Hiroaki [Nuclear Professional School / Department of Nuclear Engineering and Management, The University of Tokyo, Tokaimura, Naka-gun, Ibaraki, 319-1188 (Japan); Mori, Hideo [Department of Mechanical Engineering, Kyushu University (Japan); Akiba, Masato; Akimoto, Hajime; Okumura, Keisuke; Akasaka, Naoaki [Japan Atomic Energy Agency (Japan); GOTO, Shoji [Tokyo Electric Power Company (Japan)

2006-07-01T23:59:59.000Z

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


401

E-Print Network 3.0 - advanced fusion material Sample Search...  

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

Physics and Fusion 5 Fusion Energy Program Presentation to Summary: International Thermonuclear Experimental Reactor Plasma Technologies Fusion Technologies Advanced Materials......

402

E-Print Network 3.0 - advanced tokamak plasmas Sample Search...  

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

Physics and Fusion 15 Fusion Energy Program Presentation to Summary: International Thermonuclear Experimental Reactor Plasma Technologies Fusion Technologies Advanced Materials......

403

E-Print Network 3.0 - advanced deuterium fusion Sample Search...  

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

Physics and Fusion 2 Fusion Energy Program Presentation to Summary: International Thermonuclear Experimental Reactor Plasma Technologies Fusion Technologies Advanced Materials......

404

NUCLEAR REACTORS.  

E-Print Network [OSTI]

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

Belachew, Dessalegn

2010-01-01T23:59:59.000Z

405

Advanced Fuels Synthesis  

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

Advanced Fuels Synthesis Advanced Fuels Synthesis Coal and Coal/Biomass to Liquids Advanced Fuels Synthesis The Advanced Fuels Synthesis Key Technology is focused on catalyst and reactor optimization for producing liquid hydrocarbon fuels from coal/biomass mixtures, supports the development and demonstration of advanced separation technologies, and sponsors research on novel technologies to convert coal/biomass to liquid fuels. Active projects within the program portfolio include the following: Fischer-Tropsch fuels synthesis Small Scale Coal Biomass Liquids Production Using Highly Selective Fischer Tropsch Catalyst Small Scale Pilot Plant for the Gasification of Coal and Coal/Biomass Blends and Conversion of Derived Syngas to Liquid Fuels Via Fischer-Tropsch Synthesis Coal Fuels Alliance: Design and Construction of Early Lead Mini Fischer-Tropsch Refinery

406

International Effort to Design Nuclear Fusion Reactor Launched  

Science Journals Connector (OSTI)

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

DERMOT A. O'SULLLVAN

1988-05-23T23:59:59.000Z

407

Advanced Fuel Cycles Activities in IAEA  

SciTech Connect (OSTI)

Considerable scientific and technical progress in many areas of Partitioning and Transmutation (P and T) has been recognized as probable answers to ever-growing issues threatening sustainability, environmental protection and non-proliferation. These recent global developments such as Russian initiative on Global Nuclear Infrastructure-International Fuel Centre and the US initiative on Global Nuclear Energy Partnership (GNEP) have made advanced fuel cycles as one of the decisive influencing factor for the future growth of nuclear energy. International Atomic Energy Agency has initiated the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) with overall objective of bringing together technology holders and technology users to consider jointly the international and national actions required achieving desired innovations in nuclear reactors and fuel cycles. One of the interesting common features of these initiatives (INPRO, GNEP and GNI-IFC) is closed fast reactor fuel cycles and proliferation resistance. Any fuel cycle that integrate P and T into it is also known as 'Advanced Fuel Cycle' (AFC) that could achieve reduction of plutonium and Minor Actinide (MA) elements (namely Am, Np, Cm, etc.). In this regard, some Member States are also evaluating alternative concepts involving the use of thorium fuel cycle, inert-matrix fuel or coated particle fuel. Development of 'fast reactors with closed fuel cycles' would be the most essential step for implementation of P and T. The scale of realization of any AFC depends on the maturity of the development of all these elemental technologies such as recycling MA, Pu as well as reprocessed uranium. In accordance with the objectives of the Agency, the programme B entitled 'Nuclear Fuel cycle technologies and materials' initiated several activities aiming to strengthen the capabilities of interested Member States for policy making, strategic planning, technology development and implementation of safe, reliable, economically efficient, proliferation resistant, environmentally sound and secure nuclear fuel cycle programmes. The paper describes some on-going IAEA activities in the area of: MA-fuel and target, thorium fuel cycle, coated particle fuel, MA-property database, inert matrix fuels, liquid metal cooled fast reactor fuels and fuel cycles, management of reprocessed uranium and proliferation resistance in fuel cycle. (authors)

Nawada, H.P.; Ganguly, C. [Nuclear Fuel Cycle and Materials Section, Division of Nuclear Fuel Cycle and Waste Technology, Department of Nuclear Energy, International Atomic Energy Agency, Vienna (Austria)

2007-07-01T23:59:59.000Z

408

Mass tracking and material accounting in the Integral Fast Reactor (IFR)  

SciTech Connect (OSTI)

The Integral Fast Reactor (IFR) is a generic advanced liquid metal cooled reactor concept being developed at Argonne National Laboratory (ANL). There are a number of technical features of the IFR which contribute to its potential as a next-generation reactor. These are associated with large safety margins with regard to off-normal events involving the heat transport system, and the use of metallic fuel which makes possible the utilization of innovative fuel cycle processes. The latter feature permits fuel cycle closure the compact, low-cost reprocessing facilities, collocated with the reactor plant. These primary features are being demonstrated in the facilities at ANL-West, utilizing Experimental Breeder Reactor 2 and the associated Fuel Cycle Facility (FCF) as an IFR prototype. The demonstration of this IFR prototype includes the design and implementation of the Mass-Tracking System (MTG). In this system, data from the operations of the FCF, including weights and batch-process parameters, are collected and maintained by the MTG running on distributed workstations. The components of the MTG System include: (1) an Oracle database manager with a Fortran interface, (2) a set of MTG Tasks'' which collect, manipulate and report data, (3) a set of MTG Terminal Sessions'' which provide some interactive control of the Tasks, and (4) a set of servers which manage the Tasks and which provide the communications link between the MTG System and Operator Control Stations, which control process equipment and monitoring devices within the FCF.

Orechwa, Y.; Adams, C.H.; White, A.M.

1991-01-01T23:59:59.000Z

409

Mass tracking and material accounting in the integral fast reactor (IFR)  

SciTech Connect (OSTI)

This paper reports on the Integral Fast Reactor (IFR) which is a generic advanced liquid metal cooled reactor concept being developed at Argonne National Laboratory. There are a number of technical features of the IFR which contribute to its potential as a next-generation reactor. These are associated with large safety margins with regard to off-normal events involving the heat transport system, and the use of metallic fuel which makes possible the utilization of innovative fuel cycle processes. The latter feature permits fuel cycle closure with compact, low-cost reprocessing facilities, collocated with the reactor plant. These primary features are being demonstrated in the facilities at ANL-West, utilizing Experimental Breeder Reactor II and the associated Fuel Cycle Facility (FCF) as an IFR prototype. The demonstration of this IFR prototype includes the design and implementation of the Mass-tracking System (MTG). In this system, data from the operations of the FCF, including weights and batch-process parameters, are collected and maintained by the MTG running on distributed workstations.

Orechwa, Y.; Adams, C.H.; White, A.M. (Argonne National Lab., IL (United States). Reactor Analysis and Safety Div.)

1991-01-01T23:59:59.000Z

410

naval reactors  

National Nuclear Security Administration (NNSA)

After operating for 34 years and training over 14,000 sailors, the Department of Energy S1C Prototype Reactor Site in Windsor, Connecticut, was returned to "green field"...

411

Advanced Modeling & Simulation | Department of Energy  

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

Advanced Modeling & Simulation Advanced Modeling & Simulation Advanced Modeling & Simulation Advanced Modeling & Simulation ADVANCING THE STATE OF THE ART Innovation advances science. Historically, innovation resulted almost exclusively from fundamental theories combined with observation and experimentation over time. With advancements in engineering, computing power and visualization tools, scientists from all disciplines are gaining insights into physical systems in ways not possible with traditional approaches alone. Modeling and simulation has a long history with researchers and scientists exploring nuclear energy technologies. In fact, the existing fleet of currently operating reactors was licensed with computational tools that were produced or initiated in the 1970s. Researchers and scientists in

412

Concept:U.S. National Software Tools | Open Energy Information  

Open Energy Info (EERE)

Concept Concept Edit History Facebook icon Twitter icon » Concept:U.S. National Software Tools Jump to: navigation, search Description of concept "U.S. National Software Tools"RDF feed [[Category:Tools]] [[Developer.IsDOELab::true]] Pages of concept "U.S. National Software Tools" Showing 190 pages belonging to that concept. A A Policymaker's Guide to Feed-In Tariff Policy Design Advanced Process Engineering Co-Simulator (APECS) Africa - Technical Potential of Solar Energy to Address Energy Poverty and Avoid GHG Emissions Alternative Fuel and Advanced Technology Vehicles Pilot Program Emissions Benefit Tool Alternative Fuels and Advanced Vehicles Data Center - Codes and Standards Resources Alternative Fuels and Advanced Vehicles Data Center - Federal and State Incentives and Laws Database

413

Status of materials handbooks for particle accelerator and nuclear reactor applications  

SciTech Connect (OSTI)

In support of research and development for accelerator applications, a materials handbook was developed in August of 1998 funded by the Accelerator Production of Tritium Project. This handbook, presently called Advanced Fuel Cycle Initiative (AFCI) Materials Handbook, Materials Data for Particle Accelerator Applications, has just issued Revision 5 and contains detailed information showing the effects of irradiation on many properties for a wide variety of materials. Development of a web-accessible materials database for Generation IV Reactor Programs has been ongoing for about three years. This handbook provides a single authoritative source for qualified materials data applicable to all Generation IV reactor concepts. A beta version of this Gen IV Materials Handbook has been completed and is presently under evaluation.

Maloy, Stuart [Los Alamos National Laboratory (LANL); Rogers, Berylene [Los Alamos National Laboratory (LANL); Ren, Weiju [ORNL; Philip, Rittenhouse [Consultant

2008-01-01T23:59:59.000Z

414

Light Water Reactor Sustainability  

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

4 Light Water Reactor Sustainability ACCOMPLISHMENTS REPORT 2014 Accomplishments Report | Light Water Reactor Sustainability 2 T he mission of the Light Water Reactor...

415

Advanced Water-Gas Shift Membrane Reactor  

SciTech Connect (OSTI)

The overall objectives for this project were: (1) to identify a suitable PdCu tri-metallic alloy membrane with high stability and commercially relevant hydrogen permeation in the presence of trace amounts of carbon monoxide and sulfur; and (2) to identify and synthesize a water gas shift catalyst with a high operating life that is sulfur and chlorine tolerant at low concentrations of these impurities. This work successfully achieved the first project objective to identify a suitable PdCu tri-metallic alloy membrane composition, Pd{sub 0.47}Cu{sub 0.52}G5{sub 0.01}, that was selected based on atomistic and thermodynamic modeling alone. The second objective was partially successful in that catalysts were identified and evaluated that can withstand sulfur in high concentrations and at high pressures, but a long operating life was not achieved at the end of the project. From the limited durability testing it appears that the best catalyst, Pt-Re/Ce{sub 0.333}Zr{sub 0.333}E4{sub 0.333}O{sub 2}, is unable to maintain a long operating life at space velocities of 200,000 h{sup -1}. The reasons for the low durability do not appear to be related to the high concentrations of H{sub 2}S, but rather due to the high operating pressure and the influence the pressure has on the WGS reaction at this space velocity.

Sean Emerson; Thomas Vanderspurt; Susanne Opalka; Rakesh Radhakrishnan; Rhonda Willigan

2009-01-07T23:59:59.000Z

416