Sample records for reactor high temperature

  1. High Temperature Gas Reactors The Next Generation ?

    E-Print Network [OSTI]

    -Proof Advanced Reactor and Gas Turbine #12;Flow through Power Conversion Vessel 8 #12;9 TRISO Fuel Particle1 High Temperature Gas Reactors The Next Generation ? Professor Andrew C Kadak Massachusetts of Brayton vs. Rankine Cycle · High Temperature Helium Gas (900 C) · Direct or Indirect Cycle · Originally

  2. High Temperature Gas Reactors Briefing to

    E-Print Network [OSTI]

    Meltdown-Proof Advanced Reactor and Gas Turbine #12;TRISO Fuel Particle -- "Microsphere" · 0.9mm diameter · Utilizes gas turbine technology · Lower Power Density · Less Complicated Design (No ECCS) #12;AdvantagesHigh Temperature Gas Reactors Briefing to by Andrew C. Kadak, Ph.D. Professor of the Practice

  3. Modular Pebble Bed Reactor High Temperature Gas Reactor

    E-Print Network [OSTI]

    Built Site Assembled On--line Refueling Modules added to meet demand. No Reprocessing High Experiments Non-Proliferation Safeguards Waste Disposal Reactor Research/ Demonstration Facility

  4. Optimum Reactor Outlet Temperatures for High Temperature Gas-Cooled Reactors Integrated with Industrial Processes

    SciTech Connect (OSTI)

    Lee O. Nelson

    2011-04-01T23:59:59.000Z

    This report summarizes the results of a temperature sensitivity study conducted to identify the optimum reactor operating temperatures for producing the heat and hydrogen required for industrial processes associated with the proposed new high temperature gas-cooled reactor. This study assumed that primary steam outputs of the reactor were delivered at 17 MPa and 540C and the helium coolant was delivered at 7 MPa at 625925C. The secondary outputs of were electricity and hydrogen. For the power generation analysis, it was assumed that the power cycle efficiency was 66% of the maximum theoretical efficiency of the Carnot thermodynamic cycle. Hydrogen was generated via the hightemperature steam electrolysis or the steam methane reforming process. The study indicates that optimum or a range of reactor outlet temperatures could be identified to further refine the process evaluations that were developed for high temperature gas-cooled reactor-integrated production of synthetic transportation fuels, ammonia, and ammonia derivatives, oil from unconventional sources, and substitute natural gas from coal.

  5. High Temperature Gas Reactors Andrew C. Kadak, Ph.D.

    E-Print Network [OSTI]

    ­ fewer problems in accident · Utilizes gas turbine technology · Lower Power Density ­ no meltdownHigh Temperature Gas Reactors Andrew C. Kadak, Ph.D. Professor of the Practice Massachusetts Institute of Technology #12;#12;#12;#12;Presentation Overview · Introduction to Gas Reactors · Pebble Bed

  6. Advanced High Temperature Reactor Neutronic Core Design

    SciTech Connect (OSTI)

    Ilas, Dan [ORNL] [ORNL; Holcomb, David Eugene [ORNL] [ORNL; Varma, Venugopal Koikal [ORNL] [ORNL

    2012-01-01T23:59:59.000Z

    The AHTR is a 3400 MW(t) FHR class reactor design concept intended to serve as a central generating station type power plant. While significant technology development and demonstration remains, the basic design concept appears sound and tolerant of much of the remaining performance uncertainty. No fundamental impediments have been identified that would prevent widespread deployment of the concept. This paper focuses on the preliminary neutronic design studies performed at ORNL during the fiscal year 2011. After a brief presentation of the AHTR design concept, the paper summarizes several neutronic studies performed at ORNL during 2011. An optimization study for the AHTR core is first presented. The temperature and void coefficients of reactivity are then analyzed for a few configurations of interest. A discussion of the limiting factors due to the fast neutron fluence follows. The neutronic studies conclude with a discussion of the control and shutdown options. The studies presented confirm that sound neutronic alternatives exist for the design of the AHTR to maintain full passive safety features and reasonable operation conditions.

  7. FISSION REACTORS KEYWORDS: high-temperature

    E-Print Network [OSTI]

    Yildiz, Bilge

    that is directly cou- pled to an advanced gas-cooled reactor (AGR) is pro- posed in this paper. The system features conversion system, and the progress in the electrolysis cell materials field can help the econom- ical by a supercritical CO2 ~SCO2! power conversion system that is directly coupled to an advanced gas-cooled reactor

  8. Safety Issues for High Temperature Gas Reactors

    E-Print Network [OSTI]

    Risk Informed Safety Profile #12;LEVELS OF DEFENCE IN DEPTH (From INSAG-10) Control, limiting (reactivity insertion) Loss of Load Rod Ejection (more significant in block reactors) Failure of reactor effects and chemical attack on graphite Blow down loads and timing of accident event sequences

  9. Advanced High-Temperature, High-Pressure Transport Reactor Gasification

    SciTech Connect (OSTI)

    Michael Swanson; Daniel Laudal

    2008-03-31T23:59:59.000Z

    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.

  10. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    SciTech Connect (OSTI)

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; Davie Petti

    2009-10-01T23:59:59.000Z

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: High temperature gas reactor fuels behavior High temperature materials qualification Design methods development and validation Hydrogen production technologies Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  11. Reactor User Interface Technology Development Roadmaps for a High Temperature Gas-Cooled Reactor Outlet Temperature of 750 degrees C

    SciTech Connect (OSTI)

    Ian Mckirdy

    2010-12-01T23:59:59.000Z

    This report evaluates the technology readiness of the interface components that are required to transfer high-temperature heat from a High Temperature Gas-Cooled Reactor (HTGR) to selected industrial applications. This report assumes that the HTGR operates at a reactor outlet temperature of 750C and provides electricity and/or process heat at 700C to conventional process applications, including the production of hydrogen.

  12. Baseline Concept Description of a Small Modular High Temperature Reactor

    SciTech Connect (OSTI)

    Hans Gougar

    2014-05-01T23:59:59.000Z

    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.

  13. Medium-size high-temperature gas-cooled reactor

    SciTech Connect (OSTI)

    Peinado, C.O.; Koutz, S.L.

    1980-08-01T23:59:59.000Z

    This report summarizes high-temperature gas-cooled reactor (HTGR) experience for the 40-MW(e) Peach Bottom Nuclear Generating Station of Philadelphia Electric Company and the 330-MW(e) Fort St. Vrain Nuclear Generating Station of the Public Service Company of Colorado. Both reactors are graphite moderated and helium cooled, operating at approx. 760/sup 0/C (1400/sup 0/F) and using the uranium/thorium fuel cycle. The plants have demonstrated the inherent safety characteristics, the low activation of components, and the high efficiency associated with the HTGR concept. This experience has been translated into the conceptual design of a medium-sized 1170-MW(t) HTGR for generation of 450 MW of electric power. The concept incorporates inherent HTGR safety characteristics (a multiply redundant prestressed concrete reactor vessel (PCRV), a graphite core, and an inert single-phase coolant) and engineered safety features (core auxiliary cooling, relief valve, and steam generator dump systems).

  14. ANALYSIS OF A HIGH TEMPERATURE GAS-COOLED REACTOR POWERED HIGH TEMPERATURE ELECTROLYSIS HYDROGEN PLANT

    SciTech Connect (OSTI)

    M. G. McKellar; E. A. Harvego; A. M. Gandrik

    2010-11-01T23:59:59.000Z

    An updated reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322C and 750C, respectively. The reactor heat is used to produce heat and electric power to the HTE plant. A Rankine steam cycle with a power conversion efficiency of 44.4% was used to provide the electric power. The electrolysis unit used to produce hydrogen includes 1.1 million cells with a per-cell active area of 225 cm2. The reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 42.8% at a hydrogen production rate of 1.85 kg/s (66 million SCFD) and an oxygen production rate of 14.6 kg/s (33 million SCFD). An economic analysis of this plant was performed with realistic financial and cost estimating The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.03/kg of hydrogen was calculated assuming an internal rate of return of 10% and a debt to equity ratio of 80%/20% for a reactor cost of $2000/kWt and $2.41/kg of hydrogen for a reactor cost of $1400/kWt.

  15. High temperature gas-cooled reactor: gas turbine application study

    SciTech Connect (OSTI)

    Not Available

    1980-12-01T23:59:59.000Z

    The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project.

  16. Tritium Formation and Mitigation in High Temperature Reactors

    SciTech Connect (OSTI)

    Piyush Sabharwall; Carl Stoots

    2012-08-01T23:59:59.000Z

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. In order to prevent the tritium contamination of proposed reactor buildings and surrounding sites, this paper examines the root causes and potential solutions for the production of this radionuclide, including materials selection and inert gas sparging. A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450750C. Results of the diffusion model are presented for one steadystate value of tritium production in the reactor.

  17. Advanced High-Temperature, High-Pressure Transport Reactor Gasification

    SciTech Connect (OSTI)

    Michael L. Swanson

    2005-08-30T23:59:59.000Z

    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)

  18. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    SciTech Connect (OSTI)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Katoh, Yutai [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Wilson, Dane F [ORNL

    2008-08-01T23:59:59.000Z

    Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe and reliable operation. The focus of this document will be the overall range of DOE's structural materials research activities being conducted to support VHTR development. By far, the largest portion of material's R&D supporting VHTR development is that being performed directly as part of the Next-Generation Nuclear Plant (NGNP) Project. Supplementary VHTR materials R&D being performed in the DOE program, including university and international research programs and that being performed under direct contracts with the American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, will also be described. Specific areas of high-priority materials research that will be needed to deploy the NGNP and provide a basis for subsequent VHTRs are described, including the following: (1) Graphite: (a) Extensive unirradiated materials characterization and assessment of irradiation effects on properties must be performed to qualify new grades of graphite for nuclear service, including thermo-physical and mechanical properties and their changes, statistical variations from billot-to-billot and lot-to-lot, creep, and especially, irradiation creep. (b) Predictive models, as well as codification of the requirements and design methods for graphite core supports, must be developed to provide a basis for licensing. (2) Ceramics: Both fibrous and load-bearing ceramics must be qualified for environmental and radiation service as insulating materials. (3) Ceramic Composites: Carbon-carbon and SiC-SiC composites must be qualified for specialized usage in selected high-temperature components, such as core stabilizers, control rods, and insulating covers and ducting. This will require development of component-specific designs and fabrication processes, materials characterization, assessment of environmental and irradiation effects, and establishment of codes and standards for materials testing and design requirements. (4) Pressure Vessel Steels: (a) Qualification of short-term, high-temperature properties of light water rea

  19. Deterministic Modeling of the High Temperature Test Reactor

    SciTech Connect (OSTI)

    Ortensi, J.; Cogliati, J. J.; Pope, M. A.; Ferrer, R. M.; Ougouag, A. M.

    2010-06-01T23:59:59.000Z

    Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine INLs current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 column thin annular core, and the fully loaded core critical condition with 30 columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is used in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal z full core solver used in this study and is based on the Greens Function solution of the transverse integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, provide benchmarking capability for the DRAGON and the nodal diffusion solver codes. The results from this study show a consistent bias of 23% for the core multiplication factor. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement stems from the fact that during the experiments the control rods were adjusted to maintain criticality, whereas in the model, the rod positions were fixed. In addition, this work includes a brief study of a cross section generation approach that seeks to decouple the domain in order to account for neighbor effects. This spectral interpenetration is a dominant effect in annular HTR physics. This analysis methodology should be further explored in order to reduce the error that is systematically propagated in the traditional generation of cross sections.

  20. Preventing fuel failure for a beyond design basis accident in a fluoride salt cooled high temperature reactor

    E-Print Network [OSTI]

    Minck, Matthew J. (Matthew Joseph)

    2013-01-01T23:59:59.000Z

    The fluoride salt-cooled high-temperature reactor (FHR) combines high-temperature coated-particle fuel with a high-temperature salt coolant for a reactor with unique market and safety characteristics. This combination can ...

  1. Scaling Studies for High Temperature Test Facility and Modular High Temperature Gas-Cooled Reactor

    SciTech Connect (OSTI)

    Richard R. Schult; Paul D. Bayless; Richard W. Johnson; James R. Wolf; Brian Woods

    2012-02-01T23:59:59.000Z

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5-year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant (NGNP) project. Because the NRC's interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC).

  2. Tritium Formation and Mitigation in High-Temperature Reactor Systems

    SciTech Connect (OSTI)

    Piyush Sabharwall; Carl Stoots; Hans A. Schmutz

    2013-03-01T23:59:59.000Z

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450750 degrees C. Results of the diffusion model are presented for a steady production of tritium

  3. Tritium Formation and Mitigation in High-Temperature Reactors

    SciTech Connect (OSTI)

    Piyush Sabharwall; Carl Stoots

    2012-10-01T23:59:59.000Z

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450750 degrees C. Results of the diffusion model are presented for a steady production of tritium

  4. Tritium production analysis and management strategies for a Fluoride-salt-cooled high-temperature test reactor (FHTR)

    E-Print Network [OSTI]

    Rodriguez, Judy N

    2013-01-01T23:59:59.000Z

    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 ...

  5. THE INTEGRATION OF PROCESS HEAT APPLICATIONS TO HIGH TEMPERATURE GAS REACTORS

    SciTech Connect (OSTI)

    Michael G. McKellar

    2011-11-01T23:59:59.000Z

    A high temperature gas reactor, HTGR, can produce industrial process steam, high-temperature heat-transfer gases, and/or electricity. In conventional industrial processes, these products are generated by the combustion of fossil fuels such as coal and natural gas, resulting in significant emissions of greenhouse gases such as carbon dioxide. Heat or electricity produced in an HTGR could be used to supply process heat or electricity to conventional processes without generating any greenhouse gases. Process heat from a reactor needs to be transported by a gas to the industrial process. Two such gases were considered in this study: helium and steam. For this analysis, it was assumed that steam was delivered at 17 MPa and 540 C and helium was delivered at 7 MPa and at a variety of temperatures. The temperature of the gas returning from the industrial process and going to the HTGR must be within certain temperature ranges to maintain the correct reactor inlet temperature for a particular reactor outlet temperature. The returning gas may be below the reactor inlet temperature, ROT, but not above. The optimal return temperature produces the maximum process heat gas flow rate. For steam, the delivered pressure sets an optimal reactor outlet temperature based on the condensation temperature of the steam. ROTs greater than 769.7 C produce no additional advantage for the production of steam.

  6. Utility/user requirements for the Modular High Temperature Gas-Cooled Reactor Plant

    SciTech Connect (OSTI)

    Swart, F.E.

    1987-06-01T23:59:59.000Z

    The purpose of this document is to set forth the top level Utilty/User requirements for a Modular High Temperature Gas-Cooled Reactor electric generating plant that incorporates 4 reactors and 2 turbine-generators to produce a nominal electrical output of 550 MW net.

  7. A VUV Photoionization Study of the Combustion-Relevant Reaction of the Phenyl Radical (C6H5) with Propylene (C3H6) in a High Temperature Chemical Reactor

    E-Print Network [OSTI]

    Zhang, Fangtong

    2012-01-01T23:59:59.000Z

    a High Temperature Chemical Reactor Fangtong Zhang, Ralf I.a high temperature chemical reactor under combustion-likea high temperature chemical reactor under combustion-like

  8. Considerations of Alloy N for Fluoride Salt-Cooled High-Temperature Reactor Applications

    SciTech Connect (OSTI)

    Ren, Weiju [ORNL; Muralidharan, Govindarajan [ORNL; Wilson, Dane F [ORNL; Holcomb, David Eugene [ORNL

    2011-01-01T23:59:59.000Z

    Fluoride Salt-Cooled High-Temperature Reactors (FHRs) are a promising new class of thermal-spectrum nuclear reactors. The reactor structural materials must possess high-temperature strength and chemical compatibility with the liquid fluoride salt as well as with a power cycle fluid such as supercritical water while remaining resistant to residual air within the containment. Alloy N was developed for use with liquid fluoride salts and it possesses adequate strength and chemical compatibility up to about 700 C. A distinctive property of FHRs is that their maximum allowable coolant temperature is restricted by their structural alloy maximum service temperature. As the reactor thermal efficiency directly increases with the maximum coolant temperature, higher temperature resistant alloys are strongly desired. This paper reviews the current status of Alloy N and its relevance to FHRs including its design principles, development history, high temperature strength, environmental resistance, metallurgical stability, component manufacturability, ASME codification status, and reactor service requirements. The review will identify issues and provide guidance for improving the alloy properties or implementing engineering solutions.

  9. High-temperature gas-cooled reactors: preliminary safety and environmental information document. Volume IV

    SciTech Connect (OSTI)

    Not Available

    1980-01-01T23:59:59.000Z

    Information is presented concerning medium-enriched uranium/thorium once-through fuel cycle; medium-enrichment uranium-233/thorium recycle fuel; high-enrichment uranium-235/thorium recycle (spiked) fuel cycle; high-enrichment uranium-233/thorium recycle (spiked) fuel cycle; and gas-turbine high-temperature gas-cooled reactor.

  10. Development of ceramic membrane reactors for high temperature gas cleanup. Final report

    SciTech Connect (OSTI)

    Roberts, D.L.; Abraham, I.C.; Blum, Y.; Gottschlich, D.E.; Hirschon, A.; Way, J.D.; Collins, J.

    1993-06-01T23:59:59.000Z

    The objective of this project was to develop high temperature, high pressure catalytic ceramic membrane reactors and to demonstrate the feasibility of using these membrane reactors to control gaseous contaminants (hydrogen sulfide and ammonia) in integrated gasification combined cycle (IGCC) systems. Our strategy was to first develop catalysts and membranes suitable for the IGCC application and then combine these two components as a complete membrane reactor system. We also developed a computer model of the membrane reactor and used it, along with experimental data, to perform an economic analysis of the IGCC application. Our results have demonstrated the concept of using a membrane reactor to remove trace contaminants from an IGCC process. Experiments showed that NH{sub 3} decomposition efficiencies of 95% can be achieved. Our economic evaluation predicts ammonia decomposition costs of less than 1% of the total cost of electricity; improved membranes would give even higher conversions and lower costs.

  11. Optimizing Neutron Thermal Scattering Effects in very High Temperature Reactors

    SciTech Connect (OSTI)

    Hawari, Ayman; Ougouag, Abderrafi

    2014-07-08T23:59:59.000Z

    This project aims to develop a holistic understanding of the phenomenon of neutron thermalization in the VHTR. Neutron thermaliation is dependent on the type and structure of the moderating material. The fact that the moderator (and reflector) in the VHTR is a solid material will introduce new and interesting considerations that do not apply in other (e.g. light water) reactors. The moderator structure is expected to undergo radiation induced changes as the irradiation (or burnup) history progresses. In this case, the induced changes in structure will have a direct impact on many properties including the neutronic behavior. This can be easily anticipated if one recognizes the dependence of neutron thermalization on the scattering law of the moderator. For the pebble bed reactor, it is anticipated that the moderating behavior can be tailored, e.g. using moderators that consist of composite materials, which could allow improved optimization of the moderator-to-fuel ratio.

  12. Plutonium and minor actinide utilisation in a pebble-bed high temperature reactor

    SciTech Connect (OSTI)

    Petrov, B. Y.; Kuijper, J. C.; Oppe, J.; De Haas, J. B. M. [Nuclear Research and Consultancy Group, Westerduinweg 3, 1755 ZG Petten (Netherlands)

    2012-07-01T23:59:59.000Z

    This paper contains results of the analysis of the pebble-bed high temperature gas-cooled PUMA reactor loaded with plutonium and minor actinide (Pu/MA) fuel. Starting from knowledge and experience gained in the Euratom FP5 projects HTR-N and HTR-N1, this study aims at demonstrating the potential of high temperature reactors to utilize or transmute Pu/MA fuel. The work has been performed within the Euratom FP6 project PUMA. A number of different fuel types and fuel configurations have been analyzed and compared with respect to incineration performance and safety-related reactor parameters. The results show the excellent plutonium and minor actinide burning capabilities of the high temperature reactor. The largest degree of incineration is attained in the case of an HTR fuelled by pure plutonium fuel as it remains critical at very deep burnup of the discharged pebbles. Addition of minor actinides to the fuel leads to decrease of the achievable discharge burnup and therefore smaller fraction of actinides incinerated during reactor operation. The inert-matrix fuel design improves the transmutation performance of the reactor, while the 'wallpaper' fuel does not have advantage over the standard fuel design in this respect. After 100 years of decay following the fuel discharge, the total amount of actinides remains almost unchanged for all of the fuel types considered. Among the plutonium isotopes, only the amount of Pu-241 is reduced significantly due to its relatively short half-life. (authors)

  13. Secondary heat exchanger design and comparison for advanced high temperature reactor

    SciTech Connect (OSTI)

    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

    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)

  14. Application of Gamma code coupled with turbomachinery models for high temperature gas-cooled reactors

    SciTech Connect (OSTI)

    Chang Oh

    2008-02-01T23:59:59.000Z

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTRs higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-ofcoolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of a toxic gas, CO, and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. GAMMA code is being developed to implement turbomachinery models in the power conversion unit (PCU) and ultimately models associated with the hydrogen plant. Some preliminary results will be described in this paper.

  15. High Temperature Gas Reactors: Assessment of Applicable Codes and Standards

    SciTech Connect (OSTI)

    McDowell, Bruce K.; Nickolaus, James R.; Mitchell, Mark R.; Swearingen, Gary L.; Pugh, Ray

    2011-10-31T23:59:59.000Z

    Current interest expressed by industry in HTGR plants, particularly modular plants with power up to about 600 MW(e) per unit, has prompted NRC to task PNNL with assessing the currently available literature related to codes and standards applicable to HTGR plants, the operating history of past and present HTGR plants, and with evaluating the proposed designs of RPV and associated piping for future plants. Considering these topics in the order they are arranged in the text, first the operational histories of five shut-down and two currently operating HTGR plants are reviewed, leading the authors to conclude that while small, simple prototype HTGR plants operated reliably, some of the larger plants, particularly Fort St. Vrain, had poor availability. Safety and radiological performance of these plants has been considerably better than LWR plants. Petroleum processing plants provide some applicable experience with materials similar to those proposed for HTGR piping and vessels. At least one currently operating plant - HTR-10 - has performed and documented a leak before break analysis that appears to be applicable to proposed future US HTGR designs. Current codes and standards cover some HTGR materials, but not all materials are covered to the high temperatures envisioned for HTGR use. Codes and standards, particularly ASME Codes, are under development for proposed future US HTGR designs. A 'roadmap' document has been prepared for ASME Code development; a new subsection to section III of the ASME Code, ASME BPVC III-5, is scheduled to be published in October 2011. The question of terminology for the cross-duct structure between the RPV and power conversion vessel is discussed, considering the differences in regulatory requirements that apply depending on whether this structure is designated as a 'vessel' or as a 'pipe'. We conclude that designing this component as a 'pipe' is the more appropriate choice, but that the ASME BPVC allows the owner of the facility to select the preferred designation, and that either designation can be acceptable.

  16. Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

    SciTech Connect (OSTI)

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01T23:59:59.000Z

    Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

  17. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    SciTech Connect (OSTI)

    B. Boer; A. M. Ougouag

    2010-09-01T23:59:59.000Z

    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).

  18. Irradiation performance of AGR-1 high temperature reactor fuel

    SciTech Connect (OSTI)

    Paul A. Demkowicz; John D. Hunn; Robert N. Morris; Charles A. Baldwin; Philip L. Winston; Jason M. Harp; Scott A. Ploger; Tyler Gerczak; Isabella J. van Rooyen; Fred C. Montgomery; Chinthaka M. Silva

    2014-10-01T23:59:59.000Z

    The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO-coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.5% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuelincluding the extent of fission product release and the evolution of kernel and coating microstructureswas evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocrabon and compact matrix. The capsule-average fractional release from the compacts was 110 4 to 510 4 for 154Eu and 810 7 to 310 5 for 90Sr. The average 134Cs release from compacts was <310 6 when all particles maintained intact SiC. An estimated four particles out of 2.98105 experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs release in two capsules to approximately 10 5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. Palladium, silver, and uranium were found in the SiC layer of irradiated particles, and characterization of these elements within the SiC microstructure is the subject of ongoing focused study.

  19. Preliminary requirements for a Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR)

    SciTech Connect (OSTI)

    Massie, M.; Forsberg, C.; Forget, B. [Dept. of Nuclear Science and Engineering, Massachusetts Inst. of Technology, Cambridge, MA 02139 (United States); Hu, L. W. [Nuclear Reactor Laboratory, Massachusetts Inst. of Technology, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States)

    2012-07-01T23:59:59.000Z

    A Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR) design is being developed at MIT to provide the first demonstration and test of a salt-cooled reactor using high-temperature fuel. The first step is to define the requirements. The top level requirements are (1) provide the confidence that a larger demonstration reactor is warranted and (2) develop the necessary data for a larger-scale reactor. Because requirements will drive the design of the FHTR, a significant effort is being undertaken to define requirements and understand the tradeoffs that will be required for a practical design. The preliminary requirements include specifications for design parameters and necessary tests of major reactor systems. Testing requirements include demonstration of components, systems, and procedures for refueling, instrumentation, salt temperature control to avoid coolant freezing, salt chemistry and volume control, tritium monitoring and control, and in-service inspection. Safety tests include thermal hydraulics, neutronics - including intrinsic core shutdown mechanisms such as Doppler feedback - and decay heat removal systems. Materials and coolant testing includes fuels (including mechanical wear and fatigue) and system corrosion behavior. Preliminary analysis indicates a thermal power output below 30 MW, an initial core using pebble-bed or prismatic-block fuel, peak outlet temperatures of at least 700 deg. C, and use of FLi{sup 7}Be ({sup 7}LiF-BeF{sub 2}) coolant. The option to change-out the reactor core, fuel type, and major components is being investigated. While the FHTR will be used for materials testing, its primary mission is as a reactor system performance test to enable the design and licensing of a FHR demonstration power reactor. (authors)

  20. Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)

    SciTech Connect (OSTI)

    Ingersoll, D.T.

    2004-07-29T23:59:59.000Z

    A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the equivalent temperature of heat delivered to either the power conversion system or a hydrogen production plant. Using a comparative cost analysis, the construction costs per unit output are projected to be 50-55% of the costs for modular gas-cooled or sodium-cooled reactor systems. This is primarily a consequence of substantially larger power output and higher conversion efficiency for the AHTR. The AHTR has a number of unique technical challenges in meeting the NGNP requirements; however, it appears to offer advantages over high-temperature helium-cooled reactors and provides an alternative development path to achieve the NGNP requirements. Primary challenges include optimizing the core design for improved response to transients, designing an internal blanket to thermally protect the reactor vessel, and engineering solutions to high-temperature refueling and maintenance.

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

    SciTech Connect (OSTI)

    Sterbentz, James W

    2007-05-01T23:59:59.000Z

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

  2. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    SciTech Connect (OSTI)

    Shropshire, D.E.; Herring, J.S.

    2004-10-03T23:59:59.000Z

    The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim storage, packaging, transportation, waste forms, waste treatment, decontamination and decommissioning issues; and low-level waste (LLW) and high-level waste (HLW) disposal.

  3. Integration of High-Temperature Gas-Cooled Reactors into Industrial Process Applications

    SciTech Connect (OSTI)

    Lee Nelson

    2011-09-01T23:59:59.000Z

    This report is a summary of analyses performed by the NGNP project to determine whether it is technically and economically feasible to integrate high temperature gas cooled reactor (HTGR) technology into industrial processes. To avoid an overly optimistic environmental and economic baseline for comparing nuclear integrated and conventional processes, a conservative approach was used for the assumptions and calculations.

  4. Assessment of the high temperature fission chamber technology for the French fast reactor program

    SciTech Connect (OSTI)

    Jammes, C.; Filliatre, P.; Geslot, B.; Domenech, T.; Normand, S. [Commissariat a l'Energie Atomique, CEA (France)

    2011-07-01T23:59:59.000Z

    High temperature fission chambers are key instruments for the control and protection of the sodium-cooled fast reactor. First, the developments of those neutron detectors, which are carried out either in France or abroad are reviewed. Second, the French realizations are assessed with the use of the technology readiness levels in order to identify tracks of improvement. (authors)

  5. High Temperature Gas-cooled Reactor Projected Markets and Scoping Economics

    SciTech Connect (OSTI)

    Larry Demick

    2010-08-01T23:59:59.000Z

    The NGNP Project has the objective of developing the high temperature gas-cooled reactor (HTGR) technology to supply high temperature process heat to industrial processes as a substitute for burning of fossil fuels, such as natural gas. Applications of the HTGR technology that have been evaluated by the NGNP Project for supply of process heat include supply of electricity, steam and high-temperature gas to a wide range of industrial processes, and production of hydrogen and oxygen for use in petrochemical, refining, coal to liquid fuels, chemical, and fertilizer plants.

  6. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    SciTech Connect (OSTI)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01T23:59:59.000Z

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: Identifies pre-conceptual design requirements Develops test loop equipment schematics and layout Identifies space allocations for each of the facility functions, as required Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems Identifies pre-conceptual utility and support system needs Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  7. Preliminary Study of Bypass Flow in Prismatic Core of Very High Temperature Reactor Using Small-Scale Model

    E-Print Network [OSTI]

    Kanjanakijkasem, Worasit 1975-

    2012-11-29T23:59:59.000Z

    Very high temperature reactor (VHTR) is one of the candidates for Generation IV reactor. It can be continuously operated with average core outlet temperature between 900C and 950C, so the core temperature is one of the key features in the design...

  8. Method for fabricating wrought components for high-temperature gas-cooled reactors and product

    DOE Patents [OSTI]

    Thompson, Larry D. (San Diego, CA); Johnson, Jr., William R. (San Diego, CA)

    1985-01-01T23:59:59.000Z

    A method and alloys for fabricating wrought components of a high-temperature gas-cooled reactor are disclosed. These wrought, nickel-based alloys, which exhibit strength and excellent resistance to carburization at elevated temperatures, include aluminum and titanium in amounts and ratios to promote the growth of carburization resistant films while preserving the wrought character of the alloys. These alloys also include substantial amounts of molybdenum and/or tungsten as solid-solution strengtheners. Chromium may be included in concentrations less than 10% to assist in fabrication. Minor amounts of carbon and one or more carbide-forming metals also contribute to high-temperature strength.

  9. EVALUATION OF ZERO-POWER, ELEVATED-TEMPERATURE MEASUREMENTS AT JAPANS HIGH TEMPERATURE ENGINEERING TEST REACTOR

    SciTech Connect (OSTI)

    John D. Bess; Nozomu Fujimoto; James W. Sterbentz; Luka Snoj; Atsushi Zukeran

    2011-03-01T23:59:59.000Z

    The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about one-half of that of future HTGRs, and the high excess reactivity of the HTTR, necessary for compensation of temperature, xenon, and burnup effects during power operations, is similar to that of future HTGRs. During the start-up core physics tests of the HTTR, various annular cores were formed to provide experimental data for verification of design codes for future HTGRs. The experimental benchmark performed and currently evaluated in this report pertains to the data available for two zero-power, warm-critical measurements with the fully-loaded HTTR core. Six isothermal temperature coefficients for the fully-loaded core from approximately 340 to 740 K have also been evaluated. These experiments were performed as part of the power-up tests (References 1 and 2). Evaluation of the start-up core physics tests specific to the fully-loaded core (HTTR-GCR-RESR-001) and annular start-up core loadings (HTTR-GCR-RESR-002) have been previously evaluated.

  10. An Experimental Test Facility to Support Development of the Fluoride Salt Cooled High Temperature Reactor

    SciTech Connect (OSTI)

    Yoder Jr, Graydon L [ORNL] [ORNL; Aaron, Adam M [ORNL] [ORNL; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)] [University of Tennessee, Knoxville (UTK); Fugate, David L [ORNL] [ORNL; Holcomb, David Eugene [ORNL] [ORNL; Kisner, Roger A [ORNL] [ORNL; Peretz, Fred J [ORNL] [ORNL; Robb, Kevin R [ORNL] [ORNL; Wilgen, John B [ORNL] [ORNL; Wilson, Dane F [ORNL] [ORNL

    2014-01-01T23:59:59.000Z

    The need for high-temperature (greater than 600 C) energy exchange and delivery systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The Fluoride Salt-Cooled High-Temperature Reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process. The experimental facility described here has been constructed to support the development of the Fluoride Salt Cooled High Temperature Reactor concept. The facility is capable of operating at up to 700 C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system; trace heating system; and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride salt heat transfer inside a heated pebble bed similar to that used for the core of the pebble bed advanced high-temperature reactor. This document describes the details of the loop design, auxiliary systems used to support the facility, the inductive heating system, and facility capabilities.

  11. A New Scanning Tunneling Microscope Reactor Used for High Pressure and High Temperature Catalysis Studies

    E-Print Network [OSTI]

    Tao, Feng

    2009-01-01T23:59:59.000Z

    bowl glued at the end of the scanning tube. The arrowbowl at the end of the scanning tube. FIG. 6. (a) STM imageA New Scanning Tunneling Microscope Reactor Used for High

  12. Design Configurations and Coupling High Temperature Gas-Cooled Reactor and Hydrogen Plant

    SciTech Connect (OSTI)

    Chang H. Oh; Eung Soo Kim; Steven Sherman

    2008-04-01T23:59:59.000Z

    The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood.

  13. Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

    SciTech Connect (OSTI)

    Barsell, A.W.

    1980-05-01T23:59:59.000Z

    This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core.

  14. HYBRID SULFUR CYCLE FLOWSHEETS FOR HYDROGEN PRODUCTION USING HIGH-TEMPERATURE GAS-COOLED REACTORS

    SciTech Connect (OSTI)

    Gorensek, M.

    2011-07-06T23:59:59.000Z

    Two hybrid sulfur (HyS) cycle process flowsheets intended for use with high-temperature gas-cooled reactors (HTGRs) are presented. The flowsheets were developed for the Next Generation Nuclear Plant (NGNP) program, and couple a proton exchange membrane (PEM) electrolyzer for the SO2-depolarized electrolysis step with a silicon carbide bayonet reactor for the high-temperature decomposition step. One presumes an HTGR reactor outlet temperature (ROT) of 950 C, the other 750 C. Performance was improved (over earlier flowsheets) by assuming that use of a more acid-tolerant PEM, like acid-doped poly[2,2'-(m-phenylene)-5,5'-bibenzimidazole] (PBI), instead of Nafion{reg_sign}, would allow higher anolyte acid concentrations. Lower ROT was accommodated by adding a direct contact exchange/quench column upstream from the bayonet reactor and dropping the decomposition pressure. Aspen Plus was used to develop material and energy balances. A net thermal efficiency of 44.0% to 47.6%, higher heating value basis is projected for the 950 C case, dropping to 39.9% for the 750 C case.

  15. Fluoride Salt-Cooled High-Temperature Reactor Technology Development and Demonstration Roadmap

    SciTech Connect (OSTI)

    Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Mays, Gary T [ORNL; Pointer, William David [ORNL; Robb, Kevin R [ORNL; Yoder Jr, Graydon L [ORNL

    2013-11-01T23:59:59.000Z

    Fluoride salt-cooled High-temperature Reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics, and fully passive safety. This roadmap describes the principal remaining FHR technology challenges and the development path needed to address the challenges. This roadmap also provides an integrated overview of the current status of the broad set of technologies necessary to design, evaluate, license, construct, operate, and maintain FHRs. First-generation FHRs will not require any technology breakthroughs, but do require significant concept development, system integration, and technology maturation. FHRs are currently entering early phase engineering development. As such, this roadmap is not as technically detailed or specific as would be the case for a more mature reactor class. The higher cost of fuel and coolant, the lack of an approved licensing framework, the lack of qualified, salt-compatible structural materials, and the potential for tritium release into the environment are the most obvious issues that remain to be resolved.

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

    SciTech Connect (OSTI)

    Trond Bjornard; John Hockert

    2011-08-01T23:59:59.000Z

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

  17. Thermal Hydraulic Analyses for Coupling High Temperature Gas-Cooled Reactor to Hydrogen Plant

    SciTech Connect (OSTI)

    C.H. Oh; R. Barner; C. B. Davis; S. Sherman; P. Pickard

    2006-08-01T23:59:59.000Z

    The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood. A number of possible configurations for a system that transfers heat between the nuclear reactor and the hydrogen and/or electrical generation plants were identified. These configurations included both direct and indirect cycles for the production of electricity. Both helium and liquid salts were considered as the working fluid in the intermediate heat transport loop. Methods were developed to perform thermal-hydraulic and cycle-efficiency evaluations of the different configurations and coolants. The thermal-hydraulic evaluations estimated the sizes of various components in the intermediate heat transport loop for the different configurations. The relative sizes of components provide a relative indication of the capital cost associated with the various configurations. Estimates of the overall cycle efficiency of the various configurations were also determined. The evaluations determined which configurations and coolants are the most promising from thermalhydraulic and efficiency points of view.

  18. HTGR (High Temperature Gas-Cooled Reactor) ingress analysis using MINET

    SciTech Connect (OSTI)

    Van Tuyle, G.J.; Yang, J.W.; Kroeger, P.G.; Mallen, A.N.; Aronson, A.L.

    1989-04-01T23:59:59.000Z

    Modeling of water/steam ingress into the primary (helium) cooling circuit of a High Temperature Gas-Cooled Reactor (HTGR) is described. This modeling was implemented in the MINET Code, which is a program for analyzing transients in intricate fluid flow and heat transfer networks. Results from the simulation of a water ingress event postulated for the Modular HTGR are discussed. 27 refs., 6 figs., 6 tabs.

  19. INTEGRATION OF HIGH TEMPERATURE GAS REACTORS WITH IN SITU OIL SHALE RETORTING

    SciTech Connect (OSTI)

    Eric P. Robertson; Michael G. McKellar; Lee O. Nelson

    2011-05-01T23:59:59.000Z

    This paper evaluates the integration of a high-temperature gas-cooled reactor (HTGR) to an in situ oil shale retort operation producing 7950 m3/D (50,000 bbl/day). The large amount of heat required to pyrolyze the oil shale and produce oil would typically be provided by combustion of fossil fuels, but can also be delivered by an HTGR. Two cases were considered: a base case which includes no nuclear integration, and an HTGR-integrated case.

  20. Overall plant design specification Modular High Temperature Gas-cooled Reactor. Revision 9

    SciTech Connect (OSTI)

    NONE

    1990-05-01T23:59:59.000Z

    Revision 9 of the ``Overall Plant Design Specification Modular High Temperature Gas-Cooled Reactor,`` DOE-HTGR-86004 (OPDS) has been completed and is hereby distributed for use by the HTGR Program team members. This document, Revision 9 of the ``Overall Plant Design Specification`` (OPDS) reflects those changes in the MHTGR design requirements and configuration resulting form approved Design Change Proposals DCP BNI-003 and DCP BNI-004, involving the Nuclear Island Cooling and Spent Fuel Cooling Systems respectively.

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

    SciTech Connect (OSTI)

    Not Available

    1986-10-01T23:59:59.000Z

    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).

  2. NGNP: High Temperature Gas-Cooled Reactor Key Definitions, Plant Capabilities, and Assumptions

    SciTech Connect (OSTI)

    Wayne Moe

    2013-05-01T23:59:59.000Z

    This document provides key definitions, plant capabilities, and inputs and assumptions related to the Next Generation Nuclear Plant to be used in ongoing efforts related to the licensing and deployment of a high temperature gas-cooled reactor. These definitions, capabilities, and assumptions were extracted from a number of NGNP Project sources such as licensing related white papers, previously issued requirement documents, and preapplication interactions with the Nuclear Regulatory Commission (NRC).

  3. High-Temperature Nuclear Reactors for In-Situ Recovery of Oil from Oil Shale

    SciTech Connect (OSTI)

    Forsberg, Charles W. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6165 (United States)

    2006-07-01T23:59:59.000Z

    The world is exhausting its supply of crude oil for the production of liquid fuels (gasoline, jet fuel, and diesel). However, the United States has sufficient oil shale deposits to meet our current oil demands for {approx}100 years. Shell Oil Corporation is developing a new potentially cost-effective in-situ process for oil recovery that involves drilling wells into oil shale, using electric heaters to raise the bulk temperature of the oil shale deposit to {approx}370 deg C to initiate chemical reactions that produce light crude oil, and then pumping the oil to the surface. The primary production cost is the cost of high-temperature electrical heating. Because of the low thermal conductivity of oil shale, high-temperature heat is required at the heater wells to obtain the required medium temperatures in the bulk oil shale within an economically practical two to three years. It is proposed to use high-temperature nuclear reactors to provide high-temperature heat to replace the electricity and avoid the factor-of-2 loss in converting high-temperature heat to electricity that is then used to heat oil shale. Nuclear heat is potentially viable because many oil shale deposits are thick (200 to 700 m) and can yield up to 2.5 million barrels of oil per acre, or about 125 million dollars/acre of oil at $50/barrel. The concentrated characteristics of oil-shale deposits make it practical to transfer high-temperature heat over limited distances from a reactor to the oil shale deposits. (author)

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

    SciTech Connect (OSTI)

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

    1989-07-01T23:59:59.000Z

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

  5. MHTGR (modular high-temperature gas-cooled reactor) control: A non-safety related system

    SciTech Connect (OSTI)

    Rodriguez, C.; Swart, F.

    1988-06-01T23:59:59.000Z

    The modular high-temperature gas-cooled reactor (MHTGR) design meets stringent top-level safety regulatory criteria and user requirements that call for high plant availability and no disruption of the public's day to day activities during normal and off-normal operation of the plant. These requirements lead to a plant design that relies mainly on physical properties and passive design features to ensure plant safety regardless of operator actions, plus simplicity and automation to ensure high plant availability and lower cost of operations. The plant does not require safety-related operator actions, and it does not require the control room to be safety related.

  6. The passive safety characteristics of modular high temperature gas-cooled reactor fuel elements

    SciTech Connect (OSTI)

    Goodin, D.T.; Kania, M.J.; Nabielek, H.; Schenk, W.; Verfondern, K.

    1988-01-01T23:59:59.000Z

    High-Temperature Gas-Cooled Reactors (HTGR) in both the US and West Germany use an all-ceramic, coated fuel particle to retain fission products. Data from irradiation, postirradiation examinations and postirradiation heating experiments are used to study the performance capabilities of the fuel particles. The experimental results from fission product release tests with HTGR fuel are discussed. These data are used for development of predictive fuel performance models for purposes of design, licensing, and risk analyses. During off normal events, where temperatures may reach up to 1600/degree/C, the data show that no significant radionuclide releases from the fuel will occur.

  7. Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

    DOE Patents [OSTI]

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

    2011-03-01T23:59:59.000Z

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

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

    DOE Patents [OSTI]

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

    2010-02-23T23:59:59.000Z

    Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

  9. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    DOE Patents [OSTI]

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03T23:59:59.000Z

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

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

    SciTech Connect (OSTI)

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

    2010-09-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2008-08-01T23:59:59.000Z

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

  12. Coupling RELAP5-3D and Fluent to analyze a Very High Temperature Reactor (VHTR) outlet plenum

    E-Print Network [OSTI]

    Anderson, Nolan Alan

    2006-10-30T23:59:59.000Z

    The Very High Temperature Reactor (VHTR) system behavior should be predicted during normal operating conditions and during transient conditions. To predict the VHTR system behavior there is an urgent need for development, ...

  13. Investigation and design of a secure, transportable fluoride-salt-cooled high-temperature reactor (TFHR) for isolated locations

    E-Print Network [OSTI]

    Macdonald, Ruaridh (Ruaridh R.)

    2014-01-01T23:59:59.000Z

    In this work we describe a preliminary design for a transportable fluoride salt cooled high temperature reactor (TFHR) intended for use as a variable output heat and electricity source for off-grid locations. The goals of ...

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

    E-Print Network [OSTI]

    Wang, Huhu 1985-

    2012-12-13T23:59:59.000Z

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

  15. REPRESENTATIVE SOURCE TERMS AND THE INFLUENCE OF REACTOR ATTRIBUTES ON FUNCTIONAL CONTAINMENT IN MODULAR HIGH-TEMPERATURE GAS-COOLED REACTORS

    SciTech Connect (OSTI)

    Petti, D. A.; Hobbins, R. R.; Lowry, Peter P.; Gougar, Hans

    2013-11-01T23:59:59.000Z

    Modular high-temperature gas-cooled reactors (MHTGRs) offer a high degree of passive safety. The low power density of the reactor and the high heat capacity of the graphite core result in slow transients that do not challenge the integrity of the robust TRISO fuel. Another benefit of this fuel form and the surrounding graphite is their superior ability to retain fission products under all anticipated normal and off-normal conditions, which limits reactor accident source terms to very low values. In this paper, we develop estimates of the source term for a generic MHTGR to illustrate the performance of the radionuclide barriers that comprise the MHTGR functional containment. We also examine the influence of initial fuel quality, fuel performance/failure, reactor outlet temperature, and retention outside of the reactor core on the resultant source term to the environment.

  16. Representative Source Terms and the Influence of Reactor Attributes on Functional Containment in Modular High-Temperature Gas-Cooled Reactors

    SciTech Connect (OSTI)

    D. A. Petti; Hans Gougar; Dick Hobbins; Pete Lowry

    2013-11-01T23:59:59.000Z

    Modular high-temperature gas-cooled reactors (MHTGRs) offer a high degree of passive safety. The low power density of the reactor and the high heat capacity of the graphite core result in slow transients that do not challenge the integrity of the robust TRISO fuel. Another benefit of this fuel form and the surrounding graphite is their superior ability to retain fission products under all anticipated normal and off-normal conditions, which limits reactor accident source terms to very low values. In this paper, we develop estimates of the source term for a generic MHTGR to illustrate the performance of the radionuclide barriers that comprise the MHTGR functional containment. We also examine the influence of initial fuel quality, fuel performance/failure, reactor outlet temperature, and retention outside of the reactor core on the resultant source term to the environment.

  17. Commercial-Scale Performance Predictions for High-Temperature Electrolysis Plants Coupled to Three Advanced Reactor Types

    SciTech Connect (OSTI)

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

    2007-09-01T23:59:59.000Z

    This report presents results of system analyses that have been developed to assess the hydrogen production performance of commercial-scale high-temperature electrolysis (HTE) plants driven by three different advanced reactor power-cycle combinations: a high-temperature helium cooled reactor coupled to a direct Brayton power cycle, a supercritical CO2-cooled reactor coupled to a direct recompression cycle, and a sodium-cooled fast reactor coupled to a Rankine cycle. The system analyses were performed using UniSim software. The work described in this report represents a refinement of previous analyses in that the process flow diagrams include realistic representations of the three advanced reactors directly coupled to the power cycles and integrated with the high-temperature electrolysis process loops. In addition, this report includes parametric studies in which the performance of each HTE concept is determined over a wide range of operating conditions. Results of the study indicate that overall thermal-to- hydrogen production efficiencies (based on the low heating value of the produced hydrogen) in the 45 - 50% range can be achieved at reasonable production rates with the high-temperature helium cooled reactor concept, 42 - 44% with the supercritical CO2-cooled reactor and about 33 - 34% with the sodium-cooled reactor.

  18. High-Temperature Carbon-Irradiation Issues for the Sombrero ICF Reactor

    SciTech Connect (OSTI)

    T. Munsat

    1999-10-01T23:59:59.000Z

    In order to assess the feasibility of carbon materials for the first-wall of the Sombrero KrF laser-driven ICF fusion reactor, published experimental results relating to mechanical and thermal properties of graphites and carbon-fiber-composites (CFC's) under neutron irradiation and high heat loads are reviewed. Results are compared to published design requirements for the Sombrero ICF reactor, with particular attention to three separate issues of concern: 1. Erosion rates of the first wall are highly sensitive to the thermal conductivity value, which is itself environment-sensitive (radiation and high temperature). Erosion rates at the first wall are calculated using a high-temperature post-irradiation conductivity value of 50 W/m*K, with complete erosion of the first wall layer predicted within 14 months Sombrero full-power operation (f.p.o.), illustrating the sensitivity of erosion rates to thermal conductivity assumptions. 2. Radiation-induced swelling in 2-D and 'pseudo 3-D' CFC's is consistently {approx}20% under high-temperature neutron damage of 5 dpa (4 months f.p.o.). This level of swelling would pose technical challenges to the engineering of the target chamber modules. 3. Total tritium retention is predicted to be {approx} 0.5 to 5 kg in the Sombrero chamber within 8 months f.p.o., which may call into question safety-status assumptions of the CFC-based chamber design. These results indicate the urgency of high-temperature neutron-irradiation tests of fully symmetric 3-D CFC's in order to support the plausibility of a carbon first-wall IFE chamber such as proposed for Sombrero.

  19. Computational Fluid Dynamics Analyses on Very High Temperature Reactor Air Ingress

    SciTech Connect (OSTI)

    Chang H Oh; Eung S. Kim; Richard Schultz; David Petti; Hyung S. Kang

    2009-07-01T23:59:59.000Z

    A preliminary computational fluid dynamics (CFD) analysis was performed to understand density-gradient-induced stratified flow in a Very High Temperature Reactor (VHTR) air-ingress accident. Various parameters were taken into consideration, including turbulence model, core temperature, initial air mole-fraction, and flow resistance in the core. The gas turbine modular helium reactor (GT-MHR) 600 MWt was selected as the reference reactor and it was simplified to be 2-D geometry in modeling. The core and the lower plenum were assumed to be porous bodies. Following the preliminary CFD results, the analysis of the air-ingress accident has been performed by two different codes: GAMMA code (system analysis code, Oh et al. 2006) and FLUENT CFD code (Fluent 2007). Eventually, the analysis results showed that the actual onset time of natural convection (~160 sec) would be significantly earlier than the previous predictions (~150 hours) calculated based on the molecular diffusion air-ingress mechanism. This leads to the conclusion that the consequences of this accident will be much more serious than previously expected.

  20. Secondary Heat Exchanger Design and Comparison for Advanced High Temperature Reactor

    SciTech Connect (OSTI)

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

    2012-06-01T23:59:59.000Z

    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.

  1. KEY DESIGN REQUIREMENTS FOR THE HIGH TEMPERATURE GAS-COOLED REACTOR NUCLEAR HEAT SUPPLY SYSTEM

    SciTech Connect (OSTI)

    L.E. Demick

    2010-09-01T23:59:59.000Z

    Key requirements that affect the design of the high temperature gas-cooled reactor nuclear heat supply system (HTGR-NHSS) as the NGNP Project progresses through the design, licensing, construction and testing of the first of a kind HTGR based plant are summarized. These requirements derive from pre-conceptual design development completed to-date by HTGR Suppliers, collaboration with potential end users of the HTGR technology to identify energy needs, evaluation of integration of the HTGR technology with industrial processes and recommendations of the NGNP Project Senior Advisory Group.

  2. High Temperature Gas-Cooled Reactor Projected Markets and Preliminary Economics

    SciTech Connect (OSTI)

    Larry Demick

    2011-08-01T23:59:59.000Z

    This paper summarizes the potential market for process heat produced by a high temperature gas-cooled reactor (HTGR), the environmental benefits reduced CO2 emissions will have on these markets, and the typical economics of projects using these applications. It gives examples of HTGR technological applications to industrial processes in the typical co-generation supply of process heat and electricity, the conversion of coal to transportation fuels and chemical process feedstock, and the production of ammonia as a feedstock for the production of ammonia derivatives, including fertilizer. It also demonstrates how uncertainties in capital costs and financial factors affect the economics of HTGR technology by analyzing the use of HTGR technology in the application of HTGR and high temperature steam electrolysis processes to produce hydrogen.

  3. A Project Management and Systems Engineering Structure for a Generation IV Very High Temperature Reactor

    SciTech Connect (OSTI)

    Ed Gorski; Dennis Harrell; Finis Southworth

    2004-09-01T23:59:59.000Z

    The Very High Temperature Reactor (VHTR) will be an advanced, very high temperature (approximately 1000o C. coolant outlet temperature), gas cooled nuclear reactor and is the nearest term of six Generation IV reactor technologies for nuclear assisted hydrogen production. In 2001, the Generation IV International Forum (GIF), a ten nation international forum working together with the Department of Energys (DOE) Nuclear Energy Research Advisory Committee (NERAC), agreed to proceed with the development of a technology roadmap and identified the next generation of nuclear reactor systems for producing new sources of power. Since a new reactor has not been licensed in the United States since the 1970s, the risks are too large for a single utility to assume in the development of an unprecedented Generation IV reactor. The government must sponsor and invest in the research to resolve major first of a kind (FOAK) issues through a full-scale demonstration prior to industry implementation. DOEs primary mission for the VHTR is to demonstrate nuclear reactor assisted cogeneration of electricity and hydrogen while meeting the Generation IV goals for safety, sustainability, proliferation resistance and physical security and economics. The successful deployment of the VHTR as a demonstration project will aid in restarting the now atrophied U.S. nuclear power industry infrastructure. It is envisioned that VHTR project participants will include DOE Laboratories, industry partners such as designers, constructors, manufacturers, utilities, and Generation IV international countries. To effectively mange R&D, engineering, procurement, construction, and operation for this multi-organizational and technologically complex project, systems engineering will be used extensively to ensure delivery of the final product. Although the VHTR is an unprecedented FOAK system, the R&D, when assessed using the Office of Science and Technology Gate Model, falls primarily in the 3rd - Exploratory Development, 4th Advanced Development, and 5th- Engineering Development stages of maturity rather than in the basic and viability stages. Therefore the R&D must be controlled and project driven from the top down to address specific issues of feasibility, proof of design or support of engineering. The design evolution must be through the systems approach including an iterative process of high-level requirements definition, engineering to focus R&D to verify feasibility, requirements development and conceptual design, R&D to verify design and refine detailed requirements for final detailed design. This paper will define a framework for project management and application of systems engineering at the Idaho National Engineering and Environmental Laboratory (INEEL). The VHTR Project includes an overall reactor design and construction activity and four major supporting activities: fuel development and qualification, materials selection and qualification, NRC licensing and regulatory support, and the hydrogen production plant.

  4. Investigation on the Core Bypass Flow in a Very High Temperature Reactor

    SciTech Connect (OSTI)

    Hassan, Yassin

    2013-10-22T23:59:59.000Z

    Uncertainties associated with the core bypass flow are some of the key issues that directly influence the coolant mass flow distribution and magnitude, and thus the operational core temperature profiles, in the very high-temperature reactor (VHTR). Designers will attempt to configure the core geometry so the core cooling flow rate magnitude and distribution conform to the design values. The objective of this project is to study the bypass flow both experimentally and computationally. Researchers will develop experimental data using state-of-the-art particle image velocimetry in a small test facility. The team will attempt to obtain full field temperature distribution using racks of thermocouples. The experimental data are intended to benchmark computational fluid dynamics (CFD) codes by providing detailed information. These experimental data are urgently needed for validation of the CFD codes. The following are the project tasks: Construct a small-scale bench-top experiment to resemble the bypass flow between the graphite blocks, varying parameters to address their impact on bypass flow. Wall roughness of the graphite block walls, spacing between the blocks, and temperature of the blocks are some of the parameters to be tested. Perform CFD to evaluate pre- and post-test calculations and turbulence models, including sensitivity studies to achieve high accuracy. Develop the state-of-the art large eddy simulation (LES) using appropriate subgrid modeling. Develop models to be used in systems thermal hydraulics codes to account and estimate the bypass flows. These computer programs include, among others, RELAP3D, MELCOR, GAMMA, and GAS-NET. Actual core bypass flow rate may vary considerably from the design value. Although the uncertainty of the bypass flow rate is not known, some sources have stated that the bypass flow rates in the Fort St. Vrain reactor were between 8 and 25 percent of the total reactor mass flow rate. If bypass flow rates are on the high side, the quantity of cooling flow through the core may be considerably less than the nominal design value, causing some regions of the core to operate at temperatures in excess of the design values. These effects are postulated to lead to localized hot regions in the core that must be considered when evaluating the VHTR operational and accident scenarios.

  5. Depletion Analysis of Modular High Temperature Gas-cooled Reactor Loaded with LEU/Thorium Fuel

    SciTech Connect (OSTI)

    Sonat Sen; Gilles Youinou

    2013-02-01T23:59:59.000Z

    Thorium based fuel has been considered as an option to uranium-based fuel, based on considerations of resource utilization (Thorium is more widely available when compared to Uranium). The fertile isotope of Thorium (Th-232) can be converted to fissile isotope U-233 by neutron capture during the operation of a suitable nuclear reactor such as High Temperature Gas-cooled Reactor (HTGR). However, the fertile Thorium needs a fissile supporter to start and maintain the conversion process such as U-235 or Pu-239. This report presents the results of a study that analyzed the thorium utilization in a prismatic HTGR, namely Modular High Temperature Gas-Cooled Reactor (MHTGR) that was designed by General Atomics (GA). The collected for the modeling of this design come from Chapter 4 of MHTGR Preliminary Safety Information Document that GA sent to Department of Energy (DOE) on 1995. Both full core and unit cell models were used to perform this analysis using SCALE 6.1 and Serpent 1.1.18. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were set to match the spectral index between unit cell and full core domains. It was found that for the purposes of this study an adjusted unit cell model is adequate. Discharge isotopics and one-group cross-sections were delivered to the transmutation analysis team. This report provides documentation for these calculations

  6. The ReactorSTM: Atomically resolved scanning tunneling microscopy under high-pressure, high-temperature catalytic reaction conditions

    SciTech Connect (OSTI)

    Herbschleb, C. T.; Tuijn, P. C. van der; Roobol, S. B.; Navarro, V.; Bakker, J. W.; Liu, Q.; Stoltz, D.; Caas-Ventura, M. E.; Verdoes, G.; Spronsen, M. A. van; Bergman, M.; Crama, L.; Taminiau, I.; Frenken, J. W. M., E-mail: frenken@physics.leidenuniv.nl [Huygens-Kamerlingh Onnes Laboratory, Leiden University, P.O. box 9504, 2300 RA Leiden (Netherlands); Ofitserov, A.; Baarle, G. J. C. van [Leiden Probe Microscopy B.V., J.H. Oortweg 21, 2333 CH Leiden (Netherlands)

    2014-08-15T23:59:59.000Z

    To enable atomic-scale observations of model catalysts under conditions approaching those used by the chemical industry, we have developed a second generation, high-pressure, high-temperature scanning tunneling microscope (STM): the ReactorSTM. It consists of a compact STM scanner, of which the tip extends into a 0.5 ml reactor flow-cell, that is housed in a ultra-high vacuum (UHV) system. The STM can be operated from UHV to 6 bars and from room temperature up to 600 K. A gas mixing and analysis system optimized for fast response times allows us to directly correlate the surface structure observed by STM with reactivity measurements from a mass spectrometer. The in situ STM experiments can be combined with ex situ UHV sample preparation and analysis techniques, including ion bombardment, thin film deposition, low-energy electron diffraction and x-ray photoelectron spectroscopy. The performance of the instrument is demonstrated by atomically resolved images of Au(111) and atom-row resolution on Pt(110), both under high-pressure and high-temperature conditions.

  7. Options for treating high-temperature gas-cooled reactor fuel for repository disposal

    SciTech Connect (OSTI)

    Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

    1992-02-01T23:59:59.000Z

    This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing.

  8. Fabrication of Tungsten-Rhenium Cladding materials via Spark Plasma Sintering for Ultra High Temperature Reactor Applications

    SciTech Connect (OSTI)

    Charit, Indrajit; Butt, Darryl; Frary, Megan; Carroll, Mark

    2012-11-05T23:59:59.000Z

    This research will develop an optimized, cost-effective method for producing high-purity tungsten-rhenium alloyed fuel clad forms that are crucial for the development of a very high-temperature nuclear reactor. The study will provide critical insight into the fundamental behavior (processing-microstructure- property correlations) of W-Re alloys made using this new fabrication process comprising high-energy ball milling (HEBM) and spark plasma sintering (SPS). A broader goal is to re-establish the U.S. lead in the research field of refractory alloys, such as W-Re systems, with potential applications in very high-temperature nuclear reactors. An essential long-term goal for nuclear power is to develop the capability of operating nuclear reactors at temperatures in excess of 1,000K. This capability has applications in space exploration and some special terrestrial uses where high temperatures are needed in certain chemical or reforming processes. Refractory alloys have been identified as being capable of withstanding temperatures in excess of 1,000K and are considered critical for the development of ultra hightemperature reactors. Tungsten alloys are known to possess extraordinary properties, such as excellent high-temperature capability, including the ability to resist leakage of fissile materials when used as a fuel clad. However, there are difficulties with the development of refractory alloys: 1) lack of basic experimental data on thermodynamics and mechanical and physical properties, and 2) challenges associated with processing these alloys.

  9. Preliminary studies of coolant by-pass flows in a prismatic very high temperature reactor using computational fluid dynamics

    SciTech Connect (OSTI)

    Hiroyuki Sato; Richard Johnson; Richard Schultz

    2009-09-01T23:59:59.000Z

    Three dimensional computational fluid dynamic (CFD) calculations of a typical prismatic very high temperature gas-cooled reactor (VHTR) were conducted to investigate the influence of gap geometry on flow and temperature distributions in the reactor core using commercial CFD code FLUENT. Parametric calculations changing the gap width in a whole core length model of fuel and reflector columns were performed. The simulations show the effects of core by-pass flows in the heated core region by comparing results for several gap widths including zero gap width. The calculation results underline the importance of considering inter-column gap width for the evaluation of maximum fuel temperatures and temperature gradients in fuel blocks. In addition, it is shown that temperatures of core outlet flow from gaps and channels are strongly affected by the gap width of by-pass flow in the reactor core.

  10. Computational Fluid Dynamics Analysis of Very High Temperature Gas-Cooled Reactor Cavity Cooling System

    SciTech Connect (OSTI)

    Angelo Frisani; Yassin A. Hassan; Victor M. Ugaz

    2010-11-02T23:59:59.000Z

    The design of passive heat removal systems is one of the main concerns for the modular very high temperature gas-cooled reactors (VHTR) vessel cavity. The reactor cavity cooling system (RCCS) is a key heat removal system during normal and off-normal conditions. The design and validation of the RCCS is necessary to demonstrate that VHTRs can survive to the postulated accidents. The computational fluid dynamics (CFD) STAR-CCM+/V3.06.006 code was used for three-dimensional system modeling and analysis of the RCCS. A CFD model was developed to analyze heat exchange in the RCCS. The model incorporates a 180-deg section resembling the VHTR RCCS experimentally reproduced in a laboratory-scale test facility at Texas A&M University. All the key features of the experimental facility were taken into account during the numerical simulations. The objective of the present work was to benchmark CFD tools against experimental data addressing the behavior of the RCCS following accident conditions. Two cooling fluids (i.e., water and air) were considered to test the capability of maintaining the RCCS concrete walls' temperature below design limits. Different temperature profiles at the reactor pressure vessel (RPV) wall obtained from the experimental facility were used as boundary conditions in the numerical analyses to simulate VHTR transient evolution during accident scenarios. Mesh convergence was achieved with an intensive parametric study of the two different cooling configurations and selected boundary conditions. To test the effect of turbulence modeling on the RCCS heat exchange, predictions using several different turbulence models and near-wall treatments were evaluated and compared. The comparison among the different turbulence models analyzed showed satisfactory agreement for the temperature distribution inside the RCCS cavity medium and at the standpipes walls. For such a complicated geometry and flow conditions, the tested turbulence models demonstrated that the realizable k-epsilon model with two-layer all y+ wall treatment performs better than the other k-epsilon and k-omega turbulence models when compared to the experimental results and the Reynolds stress transport turbulence model results. A scaling analysis was developed to address the distortions introduced by the CFD model in simulating the physical phenomena inside the RCCS system with respect to the full plant configuration. The scaling analysis demonstrated that both the experimental facility and the CFD model achieve a satisfactory resemblance of the main flow characteristics inside the RCCS cavity region, and convection and radiation heat exchange phenomena are properly scaled from the actual plant.

  11. NGNP: High Temperature Gas-Cooled Reactor Key Definitions, Plant Capabilities, and Assumptions

    SciTech Connect (OSTI)

    Phillip Mills

    2012-02-01T23:59:59.000Z

    This document is intended to provide a Next Generation Nuclear Plant (NGNP) Project tool in which to collect and identify key definitions, plant capabilities, and inputs and assumptions to be used in ongoing efforts related to the licensing and deployment of a high temperature gas-cooled reactor (HTGR). These definitions, capabilities, and assumptions are extracted from a number of sources, including NGNP Project documents such as licensing related white papers [References 1-11] and previously issued requirement documents [References 13-15]. Also included is information agreed upon by the NGNP Regulatory Affairs group's Licensing Working Group and Configuration Council. The NGNP Project approach to licensing an HTGR plant via a combined license (COL) is defined within the referenced white papers and reference [12], and is not duplicated here.

  12. Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters

    SciTech Connect (OSTI)

    Frank Darmann; Robert Lombaerde; Franco Moriconi; Albert Nelson

    2011-10-31T23:59:59.000Z

    Zenergy Power has successfully designed, built, tested, and installed in the US electrical grid a saturable reactor Fault Current Limiter. Beginning in 2007, first as SC Power Systems and from 2008 as Zenergy Power, Inc., ZP used DOE matching grant and ARRA funds to help refine the design of the saturated reactor fault current limiter. ZP ultimately perfected the design of the saturated reactor FCL to the point that ZP could reliably design a suitable FCL for most utility applications. Beginning with a very basic FCL design using 1G HTS for a coil housed in a LN2 cryostat for the DC bias magnet, the technology progressed to a commercial system that was offered for sale internationally. Substantial progress was made in two areas. First, the cryogenics cooling system progressed from a sub-cooled liquid nitrogen container housing the HTS coils to cryostats utilizing dry conduction cooling and reaching temperatures down to less than 20 degrees K. Large, round cryostats with ??warm bore? diameters of 1.7 meters enabled the design of large tanks to hold the AC components. Second, the design of the AC part of the FCL was refined from a six legged ??spider? design to a more compact and lighter design with better fault current limiting capability. Further refinement of the flux path and core shape led to an efficient saturated reactor design requiring less Ampere-turns to saturate the core. In conclusion, the development of the saturable reactor FCL led to a more efficient design not requiring HTS magnets and their associated peripheral equipment, which yielded a more economical product in line with the electric utility industry expectations. The original goal for the DOE funding of the ZP project ??Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters? was to stimulate the HTS wire industry with, first 1G, then 2G, HTS wire applications. Over the approximately 5 years of ZP??s product development program, the amount of HTS wire employed per FCL and its cost as a percentage of the total FCL product content had not dropped substantially from an unsustainable level of more than 50% of the total cost of the FCL, nor had the availability increased (today the availability of 2G wire for commercial applications outside of specific partnerships with the leading 2G wire manufacturers is extremely limited). ZP had projected a very significant commercial potential for FCLs with higher performance and lower costs compared to the initial models built with 1G wire, which would come about from the widespread availability of low-cost, high-performance 2G HTS wire. The potential for 2G wires at greatly reduced performance-based prices compared to 1G HTS conductor held out the potential for the commercial production of FCLs at price and performance levels attractive to the utility industry. However, the price of HTS wire did not drop as expected and today the available quantities of 2G wire are limited, and the price is higher than the currently available supplies of 1G wire. The commercial option for ZP to provide a reliable and reasonably priced FCL to the utility industry is to employ conventional resistive conductor DC electromagnets to bias the FCL. Since the premise of the original funding was to stimulate the HTS wire industry and ZP concluded that copper-based magnets were more economical for the foreseeable future, DOE and ZP decided to mutually terminate the project.

  13. Nuclear design of small-sized high temperature gas-cooled reactor for developing countries

    SciTech Connect (OSTI)

    Goto, M.; Seki, Y.; Inaba, Y.; Ohashi, H.; Sato, H.; Fukaya, Y.; Tachibana, Y. [Japan Atomic Energy Agency, 4002, Oarai-machi, Higashi Ibaraki-gun, Ibaraki-ken 311-1394 (Japan)

    2012-07-01T23:59:59.000Z

    Japan Atomic Energy Agency (JAEA) has started a conceptual design of a small-sized HTGR with 50 MW thermal power (HTR50S), which is a first-of-a-kind commercial or demonstration plant of a small-sized HTGR to be deployed in developing countries such as Kazakhstan in the 2020's. The nuclear design of the HTR50S is performed by upgrading the proven technology of the High Temperature Engineering Test Reactor (HTTR) to reduce the cost for the construction. In the HTTR design, twelve kinds of fuel enrichment was used to optimize the power distribution, which is required to make the maximum fuel temperature below the thermal limitation during the burn-up period. However, manufacture of many kinds of fuel enrichment causes increase of the construction cost. To solve this problem, the present study challenges the nuclear design by reducing the number of fuel enrichment to as few as possible. The nuclear calculations were performed with SRAC code system whose validity was proven by the HTTR burn-up data. The calculation results suggested that the optimization of the power distribution was reasonably achieved and the maximum fuel temperature was kept below the limitation by using three kinds of fuel enrichment. (authors)

  14. Very High Temperature Reactor (VHTR) Survey of Materials Research and Development Needs to Support Early Deployment

    SciTech Connect (OSTI)

    Eric Shaber; G. Baccaglini; S. Ball; T. Burchell; B. Corwin; T. Fewell; M. Labar; P. MacDonald; P. Rittenhouse; Russ Vollam; F. Southworth

    2003-01-01T23:59:59.000Z

    The VHTR reference concept is a helium-cooled, graphite moderated, thermal neutron spectrum reactor with an outlet temperature of 1000 C or higher. It is expected that the VHTR will be purchased in the future as either an electricity producing plant with a direct cycle gas turbine or a hydrogen producing (or other process heat application) plant. The process heat version of the VHTR will require that an intermediate heat exchanger (IHX) and primary gas circulator be located in an adjoining power conversion vessel. A third VHTR mission - actinide burning - can be accomplished with either the hydrogen-production or gas turbine designs. The first ''demonstration'' VHTR will produce both electricity and hydrogen using the IHX to transfer the heat to either a hydrogen production plant or the gas turbine. The plant size, reactor thermal power, and core configuration will be designed to assure passive decay heat removal without fuel damage during accidents. The fuel cycle will be a once-through very high burnup low-enriched uranium fuel cycle. The purpose of this report is to identify the materials research and development needs for the VHTR. To do this, we focused on the plant design described in Section 2, which is similar to the GT-MHR plant design (850 C core outlet temperature). For system or component designs that present significant material challenges (or far greater expense) there may be some viable design alternatives or options that can reduce development needs or allow use of available (cheaper) materials. Nevertheless, we were not able to assess those alternatives in the time allotted for this report and, to move forward with this material research and development assessment, the authors of this report felt that it was necessary to use a GT-MHR type design as the baseline design.

  15. Gas-cooled reactor programs: high-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1981

    SciTech Connect (OSTI)

    Not Available

    1982-06-01T23:59:59.000Z

    Information is presented concerning HTGR chemistry; fueled graphite development; irradiation services for General Atomic Company; prestressed concrete pressure vessel development; HTGR structural materials; graphite development; high-temperature reactor physics studies; shielding studies; component flow test loop studies; core support performance test; and application and project assessments.

  16. Deterministic Modeling of the High Temperature Test Reactor with DRAGON-HEXPEDITE

    SciTech Connect (OSTI)

    J. Ortensi; M.A. Pope; R.M. Ferrer; J.J. Cogliati; J.D. Bess; A.M. Ougouag

    2010-10-01T23:59:59.000Z

    The Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine the INLs current prismatic reactor analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 fuel column thin annular core, and the fully loaded core critical condition with 30 fuel columns. Special emphasis is devoted to physical phenomena and artifacts in HTTR that are similar to phenomena and artifacts in the NGNP base design. The DRAGON code is used in this study since it offers significant ease and versatility in modeling prismatic designs. DRAGON can generate transport solutions via Collision Probability (CP), Method of Characteristics (MOC) and Discrete Ordinates (Sn). A fine group cross-section library based on the SHEM 281 energy structure is used in the DRAGON calculations. The results from this study show reasonable agreement in the calculation of the core multiplication factor with the MC methods, but a consistent bias of 23% with the experimental values is obtained. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement partially stems from the fact that during the experiments the control rods were adjusted to maintain criticality, whereas in the model, the rod positions were fixed. In addition, this work includes a brief study of a cross section generation approach that seeks to decouple the domain in order to account for neighbor effects. This spectral interpenetration is a dominant effect in annular HTR physics. This analysis methodology should be further explored in order to reduce the error that is systematically propagated in the traditional generation of cross sections.

  17. Implications of Graphite Radiation Damage on the Neutronic, Operational, and Safety Aspects of Very High Temperature Reactors

    SciTech Connect (OSTI)

    Hawari, Ayman I

    2011-08-30T23:59:59.000Z

    In both the prismatic and pebble bed designs of Very High Temperature Reactors (VHTR), the graphite moderator is expected to reach exposure levels of 1021 to 1022 n/cm2 over the lifetime of the reactor. This exposure results in damage to the graphite structure. In this work, molecular dynamic and ab initio molecular static calculations will be used to: 1) simulate radiation damage in graphite under various irradiation and temperature conditions, 2) generate the thermal neutron scattering cross sections for damaged graphite, and 3) examine the resulting microstructure to identify damage formations that may produce the high-temperature Wigner effect. The impact of damage on the neutronic, operational and safety behavior of the reactor will be assessed using reactor physics calculations. In addition, tests will be performed on irradiated graphite samples to search for the high-temperature Wigner effect, and phonon density of states measurements will be conducted to quantify the effect on thermal neutron scattering cross sections using these samples.

  18. Benchmark analysis of high temperature engineering test reactor core using McCARD code

    SciTech Connect (OSTI)

    Jeong, Chang Joon; Jo, Chang Keun; Lee, Hyun Chul; Noh, Jae Man [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong, Daejeon, 305-353 (Korea, Republic of)

    2013-07-01T23:59:59.000Z

    A benchmark calculation has been performed for a startup core physics test of Japan's High Temperature Engineering Test Reactor (HTTR). The calculation is carried out by the McCARD code, which adopts the Monte Carlo method. The cross section library is ENDF-B/VII.0. The fuel cell is modeled by the reactivity-equivalent physical transform (RPT) method. Effective multiplication factors with different numbers of fuel columns have been analyzed. The calculation shows that the HTTR becomes critical with 19 fuel columns with an excess reactivity of 0.84% ?k/k. The discrepancies between the measurements and Monte Carlo calculations are 2.2 and 1.4 % ?k/k for 24 and 30 columns, respectively. The reasons for the discrepancy are thought to be the current version of cross section library and the impurity in the graphite which is represented by the boron concentration. In the future, the depletion results will be proposed for further benchmark calculations. (authors)

  19. Development and Transient Analysis of a Helical-coil Steam Generator for High Temperature Reactors

    SciTech Connect (OSTI)

    Nathan V. Hoffer; Nolan A. Anderson; Piyush Sabharwall

    2011-08-01T23:59:59.000Z

    A high temperature gas-cooled reactor (HTGR) is under development by the Next Generation Nuclear Plant (NGNP) Project at the Idaho National Laboratory (INL). Its design emphasizes electrical power production which may potentially be coupled with process heat for hydrogen production and other industrial applications. NGNP is considering a helical-coil steam generator for the primary heat transport loop heat exchanger based on its increased heat transfer and compactness when compared to other steam generators. The safety and reliability of the helical-coil steam generator is currently under evaluation as part of the development of NGNP. Transients, such as loss of coolant accidents (LOCA), are of interest in evaluating the safety of steam generators. In this study, a complete steam generator inlet pipe break (double ended pipe break) LOCA was simulated by an exponential loss of primary side pressure. For this analysis, a model of the helical-coil steam generator was developed using RELAP5-3D, an INL inhouse systems analysis code. The steam generator model behaved normally during the transient simulating the complete steam generator inlet pipe break LOCA. Further analysis is required to comprehensively evaluate the safety and reliability of the helical-coil steam generator design in the NGNP setting.

  20. Use and Storage of Test and Operations Data from the High Temperature Test Reactor Acquired by the US Government from the Japan Atomic Energy Agency

    SciTech Connect (OSTI)

    Hans Gougar

    2010-02-01T23:59:59.000Z

    This document describes the use and storage of data from the High Temperature Test Reactor (HTTR) acquired from the Japan Atomic Energy Agency (JAEA) by the U.S. Government for high temperature reactor research under the Next Generation Nuclear Plant (NGNP) Project.

  1. Experimental and Analytic Study on the Core Bypass Flow in a Very High Temperature Reactor

    SciTech Connect (OSTI)

    Richard Schultz

    2012-04-01T23:59:59.000Z

    Core bypass flow has been one of key issues in the very high temperature reactor (VHTR) design for securing core thermal margins and achieving target temperatures at the core exit. The bypass flow in a prismatic VHTR core occurs through the control element holes and the radial and axial gaps between the graphite blocks for manufacturing and refueling tolerances. These gaps vary with the core life cycles because of the irradiation swelling/shrinkage characteristic of the graphite blocks such as fuel and reflector blocks, which are main components of a core's structure. Thus, the core bypass flow occurs in a complicated multidimensional way. The accurate prediction of this bypass flow and counter-measures to minimize it are thus of major importance in assuring core thermal margins and securing higher core efficiency. Even with this importance, there has not been much effort in quantifying and accurately modeling the effect of the core bypass flow. The main objectives of this project were to generate experimental data for validating the software to be used to calculate the bypass flow in a prismatic VHTR core, validate thermofluid analysis tools and their model improvements, and identify and assess measures for reducing the bypass flow. To achieve these objectives, tasks were defined to (1) design and construct experiments to generate validation data for software analysis tools, (2) determine the experimental conditions and define the measurement requirements and techniques, (3) generate and analyze the experimental data, (4) validate and improve the thermofluid analysis tools, and (5) identify measures to control the bypass flow and assess its performance in the experiment.

  2. Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing

    SciTech Connect (OSTI)

    D. L. Knudson; J. L. Rempe

    2012-02-01T23:59:59.000Z

    New materials are being considered for fuel, cladding and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine and return irradiated samples for each measurement make this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated under pressurized water reactor coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INLs High Temperature Test Laboratory.

  3. Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing

    SciTech Connect (OSTI)

    D. L. Knudson; J. L. Rempe

    2012-02-01T23:59:59.000Z

    New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine, and return irradiated samples for each measurement makes this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated in pressurized water reactor (PWR) coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL's High Temperature Test Laboratory (HTTL).

  4. CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

    SciTech Connect (OSTI)

    Hassan, Yassin; Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

    2014-07-14T23:59:59.000Z

    The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during stead-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the stead-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

  5. Concept of a thermonuclear reactor based on gravity retention of high-temperature plasma

    E-Print Network [OSTI]

    S. I. Fisenko; I. S. Fisenko

    2007-05-27T23:59:59.000Z

    In the present paper the realization of the obtained results in relation to the dense high- temperature plasma of multivalent ions including experimental data interpretation is discussed.

  6. advanced high-temperature reactor: Topics by E-print Network

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

    of different instabilities that lead to many competing orders. In high-temperature superconduct- ors, besides the superconducting phase, many competing or- ders, such as...

  7. Technology Development Roadmap for the Advanced High Temperature Reactor Secondary Heat Exchanger

    SciTech Connect (OSTI)

    P. Sabharwall; M. McCllar; A. Siahpush; D. Clark; M. Patterson; J. Collins

    2012-09-01T23:59:59.000Z

    This Technology Development Roadmap (TDRM) presents the path forward for deploying large-scale molten salt secondary heat exchangers (MS-SHX) and recognizing the benefits of using molten salt as the heat transport medium for advanced high temperature reactors (AHTR). This TDRM will aid in the development and selection of the required heat exchanger for: power production (the first anticipated process heat application), hydrogen production, steam methane reforming, methanol to gasoline production, or ammonia production. This TDRM (a) establishes the current state of molten salt SHX technology readiness, (b) defines a path forward that systematically and effectively tests this technology to overcome areas of uncertainty, (c) demonstrates the achievement of an appropriate level of maturity prior to construction and plant operation, and (d) identifies issues and prioritizes future work for maturing the state of SHX technology. This study discusses the results of a preliminary design analysis of the SHX and explains the evaluation and selection methodology. An important engineering challenge will be to prevent the molten salt from freezing during normal and off-normal operations because of its high melting temperature (390C for KF ZrF4). 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 need for efficiency, compactness, and safety challenge the capabilities of existing heat exchanger technology. The description of potential heat exchanger configurations or designs (such as printed circuit, spiral or helical coiled, ceramic, plate and fin, and plate type) were covered in an earlier report (Sabharwall et al. 2011). Significant future work, much of which is suggested in this report, is needed before the benefits and full potential of the AHTR can be realized. The execution of this TDRM will focuses research efforts on the near-term qualification, selection, or maturation strategy as detailed in this report. Development of the integration methodology feasibility study, along with research and development (R&D) needs, are ongoing tasks that will be covered in the future reports as work progresses. Section 2 briefly presents the integration of AHTR technology with conventional chemical industrial processes., See Idaho National Laboratory (INL) TEV-1160 (2011) for further details

  8. Separation Requirements for a Hydrogen Production Plant and High-Temperature Nuclear Reactor

    SciTech Connect (OSTI)

    Curtis Smith; Scott Beck; Bill Galyean

    2005-09-01T23:59:59.000Z

    This report provides the methods, models, and results of an evaluation for locating a hydrogen production facility near a nuclear power plant. In order to answer the risk-related questions for this combined nuclear and chemical facility, we utilized standard probabilistic safety assessment methodologies to answer three questions: what can happen, how likely is it, and what are the consequences? As part of answering these questions, we developed a model suitable to determine separation distances for hydrogen process structures and the nuclear plant structures. Our objective of the model-development and analysis is to answer key safety questions related to the placement of one or more hydrogen production plants in the vicinity of a high-temperature nuclear reactor. From a thermal-hydraulic standpoint we would like the two facilities to be quite close. However, safety and regulatory implications force the separation distance to be increased, perhaps substantially. Without answering these safety questions, the likelihood for obtaining a permit to construct and build such as facility in the U.S. would be questionable. The quantitative analysis performed for this report provides us with a scoping mechanism to determine key parameters related to the development of a nuclear-based hydrogen production facility. From our calculations, we estimate that when the separation distance is less than 100m, the core damage frequency is large enough (greater than 1E-6/yr) to become problematic in a risk-informed environment. However, a variety of design modifications, for example blast-deflection barriers, were explored to determine the impact of potential mitigating strategies. We found that these mitigating cases may significantly reduce risk and should be explored as the design for the hydrogen production facility evolves.

  9. Transmutation Analysis of Enriched Uranium and Deep Burn High Temperature Reactors

    SciTech Connect (OSTI)

    Michael A. Pope

    2012-07-01T23:59:59.000Z

    High temperature reactors (HTRs) have been under consideration for production of electricity, process heat, and for destruction of transuranics for decades. As part of the transmutation analysis efforts within the Fuel Cycle Research and Development (FCR&D) campaign, a need was identified for detailed discharge isotopics from HTRs for use in the VISION code. A conventional HTR using enriched uranium in UCO fuel was modeled having discharge burnup of 120 GWd/MTiHM. Also, a deep burn HTR (DB-HTR) was modeled burning transuranic (TRU)-only TRU-O2 fuel to a discharge burnup of 648 GWd/MTiHM. For each of these cases, unit cell depletion calculations were performed with SCALE/TRITON. Unit cells were used to perform this analysis using SCALE 6.1. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were first set by using Serpent calculations to match a spectral index between unit cell and whole core domains. In the case of the DB-HTR, the unit cell which was arrived at in this way conserved the ratio of fuel to moderator found in a single block of fuel. In the conventional HTR case, a larger moderator-to-fuel ratio than that of a single block was needed to simulate the whole core spectrum. Discharge isotopics (for 500 nuclides) and one-group cross-sections (for 1022 nuclides) were delivered to the transmutation analysis team. This report provides documentation for these calculations. In addition to the discharge isotopics, one-group cross-sections were provided for the full list of 1022 nuclides tracked in the transmutation library.

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

    SciTech Connect (OSTI)

    Philip Casey Durst; Mark Schanfein

    2012-08-01T23:59:59.000Z

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

  11. Studies Related to the Oregon State University High Temperature Test Facility: Scaling, the Validation Matrix, and Similarities to the Modular High Temperature Gas-Cooled Reactor

    SciTech Connect (OSTI)

    Richard R. Schultz; Paul D. Bayless; Richard W. Johnson; William T. Taitano; James R. Wolf; Glenn E. McCreery

    2010-09-01T23:59:59.000Z

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5 year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant project. Because the NRC interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC). Since DOE has incorporated the HTTF as an ingredient in the NGNP thermal-fluids validation program, several important outcomes should be noted: 1. The reference prismatic reactor design, that serves as the basis for scaling the HTTF, became the modular high temperature gas-cooled reactor (MHTGR). The MHTGR has also been chosen as the reference design for all of the other NGNP thermal-fluid experiments. 2. The NGNP validation matrix is being planned using the same scaling strategy that has been implemented to design the HTTF, i.e., the hierarchical two-tiered scaling methodology developed by Zuber in 1991. Using this approach a preliminary validation matrix has been designed that integrates the HTTF experiments with the other experiments planned for the NGNP thermal-fluids verification and validation project. 3. Initial analyses showed that the inherent power capability of the OSU infrastructure, which only allowed a total operational facility power capability of 0.6 MW, is inadequate to permit steady-state operation at reasonable conditions. 4. To enable the HTTF to operate at a more representative steady-state conditions, DOE recently allocated funding via a DOE subcontract to HTTF to permit an OSU infrastructure upgrade such that 2.2 MW will become available for HTTF experiments. 5. Analyses have been performed to study the relationship between HTTF and MHTGR via the hierarchical two-tiered scaling methodology which has been used successfully in the past, e.g., APEX facility scaling to the Westinghouse AP600 plant. These analyses have focused on the relationship between key variables that will be measured in the HTTF to the counterpart variables in the MHTGR with a focus on natural circulation, using nitrogen as a working fluid, and core heat transfer. 6. Both RELAP5-3D and computational fluid dynamics (CD-Adapcos STAR-CCM+) numerical models of the MHTGR and the HTTF have been constructed and analyses are underway to study the relationship between the reference reactor and the HTTF. The HTTF is presently being designed. It has -scaling relationship to the MHTGR in both the height and the diameter. Decisions have been made to design the reactor cavity cooling system (RCCS) simulation as a boundary condition for the HTTF to ensure that (a) the boundary condition is well defined and (b) the boundary condition can be modified easily to achieve the desired heat transfer sink for HTTF experimental operations.

  12. The development and operational testing of an experimental reactor for gas-liquid-solid reaction systems at high temperatures and pressures

    E-Print Network [OSTI]

    Hess, Richard Kenneth

    1985-01-01T23:59:59.000Z

    THE DEVELOPMENT AND OPERATIONAL TESTING OF AN EXPERIMENTAL REACTOR FOR GAS-LIQUID-SOLID REACTION SYSTEMS AT HIGH TEMPERATURES AND PRESSURES A Thesis by RICHARD KENNETH HESS Submitted to the Graduate College of Texas A&M University in partial... fulfillment of the requirements for the degree of MASTER OF SCIENCE August 1985 Major Subject: Chemical Engineering THE DEVELOPMENT AND OPERATIONAL TESTING OF AN EXPERIMENTAL REACTOR FOR GAS-LIQUID-SOLID REACTION SYSTEMS AT HIGH TEMPERATURES...

  13. An Advanced Integrated Diffusion/Transport Method for the Design, Analysis and Optimization of the Very-High-Temperature Reactors

    SciTech Connect (OSTI)

    Farzad Rahnema; Dingkang Zhang; Abderrafi Ougouag; Frederick Gleicher

    2011-04-04T23:59:59.000Z

    The main objective of this research is to develop an integrated diffusion/transport (IDT) method to substantially improve the accuracy of nodal diffusion methods for the design and analysis of Very High Temperature Reactors (VHTR). Because of the presence of control rods in the reflector regions in the Pebble Bed Reactor (PBR-VHTR), traditional nodal diffusion methods do not accurately model these regions, within which diffusion theory breaks down in the vicinity of high neutron absorption and steep flux gradients. The IDT method uses a local transport solver based on a new incident flux response expansion method in the controlled nodes. Diffusion theory is used in the rest of the core. This approach improves the accuracy of the core solution by generating transport solutions of controlled nodes while maintaining computational efficiency by using diffusion solutions in nodes where such a treatment is sufficient. The transport method is initially developed and coupled to the reformulated 3-D nodal diffusion model in the CYNOD code for PBR core design and fuel cycle analysis. This method is also extended to the prismatic VHTR. The new method accurately captures transport effects in highly heterogeneous regions with steep flux gradients. The calculations of these nodes with transport theory avoid errors associated with spatial homogenization commonly used in diffusion methods in reactor core simulators

  14. Fission Product Impact Reduction via Protracted In-core Retention in Very High Temperature Reactor (VHTR) Transmutation Scenarios

    E-Print Network [OSTI]

    Alajo, Ayodeji Babatunde

    2011-08-08T23:59:59.000Z

    of HTGR by improvements in thermal efficiency and deployment for high-temperature applications such as hydrogen production, sea-water desalination and industrial process heat supply [17]. The VHTR is a graphite-moderated helium-cooled reactor...-based transmutation concept takes advantage of the higher number of steps it takes for a neutron to slow-down to thermal energies in graphite than the steps required in conventional LWR. The reduced slowing-down rate in graphite media favors the attainment...

  15. Modelling of trickle-bed reactors at high temperatures and pressures

    E-Print Network [OSTI]

    Collins, George Michael

    1983-01-01T23:59:59.000Z

    reactors in the petroleum industry has inspired interest in predicting the performance of these reactors for the purposes of design and operation. A trickle-bed reactor is a three-phase reactor with a fixed catalyst bed and concurrent downward flow... of gas and liquid phases. Trickle-flow involves a continuous gas phase with a liquid phase trickling in a thin film or rivulets over the catalyst bed. Optimally, the catalyst pellets are entirely covered by a thin layer of liquid, through which...

  16. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

    SciTech Connect (OSTI)

    Michael A. Pope

    2011-10-01T23:59:59.000Z

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

  17. Aerosol Resuspension Model for MELCOR for Fusion and Very High Temperature Reactor Applications

    SciTech Connect (OSTI)

    B.J. Merrill

    2011-01-01T23:59:59.000Z

    Dust is generated in fusion reactors from plasma erosion of plasma facing components within the reactors vacuum vessel (VV) during reactor operation. This dust collects in cooler regions on interior surfaces of the VV. Because this dust can be radioactive, toxic, and/or chemically reactive, it poses a safety concern, especially if mobilized by the process of resuspension during an accident and then transported as an aerosol though out the reactor confinement building, and possibly released to the environment. A computer code used at the Idaho National Laboratory (INL) to model aerosol transport for safety consequence analysis is the MELCOR code. A primary reason for selecting MELCOR for this application is its aerosol transport capabilities. The INL Fusion Safety Program (FSP) organization has made fusion specific modifications to MELCOR. Recent modifications include the implementation of aerosol resuspension models in MELCOR 1.8.5 for Fusion. This paper presents the resuspension models adopted and the initial benchmarking of these models.

  18. Applications of Nd:YAG laser micromanufacturing in High Temperature Gas Reactor research

    SciTech Connect (OSTI)

    I. J. van Rooyen; C. A. Smal; J. Steyn; H. Greyling

    2012-08-01T23:59:59.000Z

    Two innovative applications of Nd:YAG laser micromachining techniques are demonstrated in this publication. Research projects to determine the fission product transport mechanisms in TRISO coated particles necessitate heat treatment studies as well as the manufacturing of a unique sealed system for experimentation at very high temperatures. This article describes firstly the design and creation of an alumina jig designed to contain 500 {mu}m diameter ZrO2 spheres intended for annealing experiments at temperatures up to 1600 C. Functional requirements of this jig are the precision positioning of spheres for laser ablation, welding and post weld heat treatment in order to ensure process repeatability and accurate indexing of individual spheres. The design challenges and the performance of the holding device are reported. Secondly the manufacture of a sealing system using laser micromachining is reported. ZrO2 micro plugs isolate the openings of micro-machined cavities to produce a gas-tight seal fit for application in a high temperature environment. The technique is described along with a discussion of the problems experienced during the sealing process. Typical problems experienced were seating dimensions and the relative small size ({approx} 200 {mu}m) of these plugs that posed handling challenges. Manufacturing processes for both the tapered seating cavity and the plug are demonstrated. In conclusion, this article demonstrates the application of Nd-YAG micromachining in an innovative way to solve practical research problems.

  19. Configuration adjustment potential of the Very High Temperature Reactor prismatic cores with advanced actinide fuels

    E-Print Network [OSTI]

    Ames, David E, II

    2006-10-30T23:59:59.000Z

    Minor actinides represent the long-term radiotoxicity of nuclear wastes. As one of their potential incineration options, partitioning and transmutation in fission reactors are seriously considered worldwide. If implemented, ...

  20. Configuration adjustment potential of the Very High Temperature Reactor prismatic cores with advanced actinide fuels

    E-Print Network [OSTI]

    Ames, David E, II

    2006-10-30T23:59:59.000Z

    Minor actinides represent the long-term radiotoxicity of nuclear wastes. As one of their potential incineration options, partitioning and transmutation in fission reactors are seriously considered worldwide. If implemented, these technologies could...

  1. High-temperature gas-cooled reactor (HTGR): long term program plan

    SciTech Connect (OSTI)

    Not Available

    1980-10-09T23:59:59.000Z

    The FY 1980 effort was to investigate four technology options identified by program participants as potentially viable candidates for near-term demonstration: the Gas Turbine system (HTGR-GT), reflecting its perceived compatibility with the dry-cooling market, two systems addressing the process heat market, the Reforming (HTGR-R) and Steam Cycle (HTGR-SC) systems, and a more developmental reactor system, The Nuclear Heat Source Demonstration Reactor (NHSDR), which was to serve as a basis for both the HTGR-GT and HTGR-R systems as well as the further potential for developing advanced applications such as steam-coal gasification and water splitting.

  2. Integration of High Temperature Gas-cooled Reactor Technology with Oil Sands Processes

    SciTech Connect (OSTI)

    L.E. Demick

    2011-10-01T23:59:59.000Z

    This paper summarizes an evaluation of siting an HTGR plant in a remote area supplying steam, electricity and high temperature gas for recovery and upgrading of unconventional crude oil from oil sands. The area selected for this evaluation is the Alberta Canada oil sands. This is a very fertile and active area for bitumen recovery and upgrading with significant quantities piped to refineries in Canada and the U.S Additionally data on the energy consumption and other factors that are required to complete the evaluation of HTGR application is readily available in the public domain. There is also interest by the Alberta oil sands producers (OSP) in identifying alternative energy sources for their operations. It should be noted, however, that the results of this evaluation could be applied to any similar oil sands area.

  3. Long term out-of-pile thermocouple tests in conditions representative for nuclear gas-cooled high temperature reactors

    SciTech Connect (OSTI)

    Laurie, M. [European Commission, Joint Research Centre, Inst. for Energy, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Fourrez, S. [THERMOCOAX SAS, BP 26, Planquivon, F-61438 Flers Cedex (France); Fuetterer, M. A.; Lapetite, J. M. [European Commission, Joint Research Centre, Inst. for Energy, P.O. Box 2, NL-1755 ZG Petten (Netherlands)

    2011-07-01T23:59:59.000Z

    During irradiation tests at high temperature, failure of commercial Inconel 600 sheathed thermocouples is commonly encountered. To understand and remedy this problem, out-of-pile tests were performed with thermocouples in carburizing atmospheres which can be assumed to be at least locally representative for High Temperature Reactors. The objective was to screen those thermocouples which would consecutively be used under irradiation. Two such screening tests have been performed with a set of thermocouples embedded in graphite (mainly conventional Type N thermocouples and thermocouples with innovative sheaths) in a dedicated furnace with helium flushing. Performance indicators such as thermal drift, insulation and loop resistance were monitored and compared to those from conventional Type N thermocouples. Several parameters were investigated: niobium sleeves, bending, thickness, sheath composition, temperature as well as the chemical environment. After the tests, Scanning Electron Microscopy (SEM) examinations were performed to analyze possible local damage in wires and in the sheath. The present paper describes the two experiments, summarizes results and outlines further work, in particular to further analyze the findings and to select suitable thermocouples for qualification under irradiation. (authors)

  4. Magnitude and reactivity consequences of moisture ingress into the modular High-Temperature Gas-Cooled Reactor core

    SciTech Connect (OSTI)

    Smith, O.L. (Oak Ridge National Lab., TN (United States))

    1992-12-01T23:59:59.000Z

    Inadvertent admission of moisture into the primary system of a modular high-temperature gas-cooled reactor has been identified in US Department of Energy-sponsored studies as an important safety concern. The work described here develops an analytical methodology to quantify the pressure and reactivity consequences of steam-generator tube rupture and other moisture-ingress-related incidents. Important neutronic and thermohydraulic processes are coupled with reactivity feedback and safety and control system responses. The rate and magnitude of steam buildup are found to be dominated by major system features such as break size compared with safety valve capacity and reliability and less sensitive to factors such as heat transfer coefficients. The results indicate that ingress transients progress at a slower pace than previously predicted by bounding analyses, with milder power overshoots and more time for operator or automatic corrective actions.

  5. Benchmark Evaluation of Start-Up and Zero-Power Measurements at the High Temperature Engineering Test Reactor

    SciTech Connect (OSTI)

    John D. Bess; Nozomu Fujimoto

    2014-11-01T23:59:59.000Z

    Benchmark models were developed to evaluate six cold-critical and two warm-critical, zero-power measurements of the HTTR. Additional measurements of a fully-loaded subcritical configuration, core excess reactivity, shutdown margins, six isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as acceptable benchmark experiments. Insufficient information is publicly available to develop finely-detailed models of the HTTR as much of the design information is still proprietary. However, the uncertainties in the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias uncertainties incurred through the simplification process used to develop the benchmark models. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the impurity content of the various graphite blocks that comprise the HTTR. Monte Carlo calculations of keff are between approximately 0.9 % and 2.7 % greater than the benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulation of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  6. Benchmark Evaluation of Start-Up and Zero-Power Measurements at the High-Temperature Engineering Test Reactor

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Bess, John D.; Fujimoto, Nozomu

    2014-10-09T23:59:59.000Z

    Benchmark models were developed to evaluate six cold-critical and two warm-critical, zero-power measurements of the HTTR. Additional measurements of a fully-loaded subcritical configuration, core excess reactivity, shutdown margins, six isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as acceptable benchmark experiments. Insufficient information is publicly available to develop finely-detailed models of the HTTR as much of the design information is still proprietary. However, the uncertainties in the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias uncertainties incurred through the simplification process used to develop the benchmark models. Dominant uncertainties in themoreexperimental keff for all core configurations come from uncertainties in the impurity content of the various graphite blocks that comprise the HTTR. Monte Carlo calculations of keff are between approximately 0.9 % and 2.7 % greater than the benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulation of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.less

  7. Benchmark Evaluation of Start-Up and Zero-Power Measurements at the High-Temperature Engineering Test Reactor

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Bess, John D.; Fujimoto, Nozomu

    2014-10-09T23:59:59.000Z

    Benchmark models were developed to evaluate six cold-critical and two warm-critical, zero-power measurements of the HTTR. Additional measurements of a fully-loaded subcritical configuration, core excess reactivity, shutdown margins, six isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as acceptable benchmark experiments. Insufficient information is publicly available to develop finely-detailed models of the HTTR as much of the design information is still proprietary. However, the uncertainties in the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias uncertainties incurred through the simplification process used to develop the benchmark models. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the impurity content of the various graphite blocks that comprise the HTTR. Monte Carlo calculations of keff are between approximately 0.9 % and 2.7 % greater than the benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulation of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

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

    SciTech Connect (OSTI)

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

    2014-03-01T23:59:59.000Z

    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.

  9. Development and validation of scale nuclear analysis methods for high temperature gas-cooled reactors

    SciTech Connect (OSTI)

    Gehin, Jess C [ORNL] [ORNL; Jessee, Matthew Anderson [ORNL] [ORNL; Williams, Mark L [ORNL] [ORNL; Lee, Deokjung [ORNL] [ORNL; Goluoglu, Sedat [ORNL] [ORNL; Ilas, Germina [ORNL] [ORNL; Ilas, Dan [ORNL] [ORNL; Bowman, Steve A [ORNL] [ORNL

    2010-01-01T23:59:59.000Z

    In support of the U.S. Nuclear Regulatory Commission, ORNL is updating the nuclear analysis methods and data in the SCALE code system to support modeling of HTGRs. Development activities include methods used for reactor physics, criticality safety, and radiation shielding. This paper focuses on the nuclear methods in support of reactor physics, which primarily include lattice physics for cross-section processing of both prismatic and pebble-bed designs, Monte Carlo depletion methods and efficiency improvements for double heterogeneous fuels, and validation against relevant experiments. These methods enhancements are being validated using available experimental data from the HTTR and HTR-10 startup and initial criticality experiments. Results obtained with three-dimensional Monte Carlo models of the HTTR initial core critical configurations with SCALE6/KENO show excellent agreement between the continuous energy and multigroup methods and the results are consistent with results obtained by others. A three-dimensional multigroup Monte Carlo model for the initial critical core of the HTR-10 has been developed with SCALE6/KENO based on the benchmark specifications included in the IRPhE Handbook. The core eigenvalue obtained with this model is in very good agreement with the corresponding value obtained with a consistent continuous energy MCNP5 core model.

  10. High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant

    SciTech Connect (OSTI)

    J. M. Beck; L. F. Pincock

    2011-04-01T23:59:59.000Z

    The purpose of this report is to identify possible issues highlighted by these lessons learned that could apply to the NGNP in reducing technical risks commensurate with the current phase of design. Some of the lessons learned have been applied to the NGNP and documented in the Preconceptual Design Report. These are addressed in the background section of this document and include, for example, the decision to use TRISO fuel rather than BISO fuel used in the Peach Bottom reactor; the use of a reactor pressure vessel rather than prestressed concrete found in Fort St. Vrain; and the use of helium as a primary coolant rather than CO2. Other lessons learned, 68 in total, are documented in Sections 2 through 6 and will be applied, as appropriate, in advancing phases of design. The lessons learned are derived from both negative and positive outcomes from prior HTGR experiences. Lessons learned are grouped according to the plant, areas, systems, subsystems, and components defined in the NGNP Preconceptual Design Report, and subsequent NGNP project documents.

  11. High-temperature gas-cooled reactor safety studies for the Division of Accident Evaluation quarterly progress report, January 1-March 31, 1985

    SciTech Connect (OSTI)

    Ball, S.J.; Cleveland, J.C.; Harrington, R.M.; Weber, C.F.; Wilson, J.H.

    1985-10-01T23:59:59.000Z

    Modeling, code development, and analyses of the modular High-Temperature Gas-Cooled Reactor (HTGR) continued with work on the side-by-side design. Fission-product release and transport experiments were completed. A description and assessment report on Oak Ridge National Laboratory HTGR safety codes was issued.

  12. High-Temperature Carbon-Irradiation Issues for the Sombrero ICF Reactor

    E-Print Network [OSTI]

    in any graphitic material including polycrystalline and carbon-fiber composite materials. The exact (609) 243-2418 #12;2 Abstract In order to assess the feasibility of carbon materials for the first and thermal properties of graphites and carbon-fiber-composites (CFC's) under neutron irradiation and high

  13. Status of Physics and Safety Analyses for the Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR)

    SciTech Connect (OSTI)

    Ingersoll, DT

    2005-12-15T23:59:59.000Z

    A study has been completed to develop a new baseline core design for the liquid-salt-cooled very high-temperature reactor (LS-VHTR) that is better optimized for liquid coolant and that satisfies the top-level operational and safety targets, including strong passive safety performance, acceptable fuel cycle parameters, and favorable core reactivity response to coolant voiding. Three organizations participated in the study: Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL), and Argonne National Laboratory (ANL). Although the intent was to generate a new reference LS-VHTR core design, the emphasis was on performing parametric studies of the many variables that constitute a design. The results of the parametric studies not only provide the basis for choosing the optimum balance of design options, they also provide a valuable understanding of the fundamental behavior of the core, which will be the basis of future design trade-off studies. A new 2400-MW(t) baseline design was established that consists of a cylindrical, nonannular core cooled by liquid {sup 7}Li{sub 2}BeF{sub 4} (Flibe) salt. The inlet and outlet coolant temperatures were decreased by 50 C, and the coolant channel diameter was increased to help lower the maximum fuel and vessel temperatures. An 18-month fuel cycle length with 156 GWD/t burnup was achieved with a two-batch shuffling scheme, while maintaining a core power density of 10 MW/m{sup 3} using graphite-coated uranium oxicarbide particle fuel enriched to 15% {sup 235}U and assuming a 25 vol-% packing of the coated particles in the fuel compacts. The revised design appears to have excellent steady-state and transient performance. The previous concern regarding the core's response to coolant voiding has been resolved for the case of Flibe coolant by increasing the coolant channel diameter and the fuel loading. Also, the LSVHTR has a strong decay heat removal performance and appears capable of surviving a loss of forced circulation (LOFC) even with failure to scram. Significant natural convection of the coolant salt occurs, resulting in fuel temperatures below steady-state values and nearly uniform temperature distributions during the transient.

  14. Fuel Summary for Peach Bottom Unit 1 High-Temperature Gas-Cooled Reactor Cores 1 and 2

    SciTech Connect (OSTI)

    Karel I. Kingrey

    2003-04-01T23:59:59.000Z

    This fuel summary report contains background and summary information for the Peach Bottom Unit 1, High-Temperature, Gas-Cooled Reactor Cores 1 and 2. This report contains detailed information about the fuel in the two cores, the Peach Bottom Unit 1 operating history, nuclear parameters, physical and chemical characteristics, and shipping and storage canister related data. The data in this document have been compiled from a large number of sources and are not qualified beyond the qualification of the source documents. This report is intended to provide an overview of the existing data pertaining to spent fuel management and point to pertinent reference source documents. For design applications, the original source documentation must be used. While all referenced sources are available as records or controlled documents at the Idaho National Engineering and Environmental Laboratory (INEEL), some of the sources were marked as informal or draft reports. This is noted where applicable. In some instances, source documents are not consistent. Where they are known, this document identifies those instances and provides clarification where possible. However, as stated above, this document has not been independently qualified and such clarifications are only included for information purposes. Some of the information in this summary is available in multiple source documents. An effort has been made to clearly identify at least one record document as the source for the information included in this report.

  15. Life cycle assessment of hydrogen production from S-I thermochemical process coupled to a high temperature gas reactor

    SciTech Connect (OSTI)

    Giraldi, M. R.; Francois, J. L.; Castro-Uriegas, D. [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac No. 8532, Col. Progreso, C.P. 62550, Jiutepec, Morelos (Mexico)

    2012-07-01T23:59:59.000Z

    The purpose of this paper is to quantify the greenhouse gas (GHG) emissions associated to the hydrogen produced by the sulfur-iodine thermochemical process, coupled to a high temperature nuclear reactor, and to compare the results with other life cycle analysis (LCA) studies on hydrogen production technologies, both conventional and emerging. The LCA tool was used to quantify the impacts associated with climate change. The product system was defined by the following steps: (i) extraction and manufacturing of raw materials (upstream flows), (U) external energy supplied to the system, (iii) nuclear power plant, and (iv) hydrogen production plant. Particular attention was focused to those processes where there was limited information from literature about inventory data, as the TRISO fuel manufacture, and the production of iodine. The results show that the electric power, supplied to the hydrogen plant, is a sensitive parameter for GHG emissions. When the nuclear power plant supplied the electrical power, low GHG emissions were obtained. These results improve those reported by conventional hydrogen production methods, such as steam reforming. (authors)

  16. Balance of Plant System Analysis and Component Design of Turbo-Machinery for High Temperature Gas Reactor Systems

    SciTech Connect (OSTI)

    Ronald G. Ballinger Chunyun Wang Andrew Kadak Neil Todreas

    2004-08-30T23:59:59.000Z

    The Modular Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion system for it to achieve economic competitiveness as a Generation IV nuclear system. The availability of controllable helium turbomachinery and compact heat exchangers are thus the critical enabling technology for the gas turbine cycle. The development of an initial reference design for an indirect helium cycle has been accomplished with the overriding constraint that this design could be built with existing technology and complies with all current codes and standards. Using the initial reference design, limiting features were identified. Finally, an optimized reference design was developed by identifying key advances in the technology that could reasonably be expected to be achieved with limited R&D. This final reference design is an indirect, intercooled and recuperated cycle consisting of a three-shaft arrangement for the turbomachinery system. A critical part of the design process involved the interaction between individual component design and overall plant performance. The helium cycle overall efficiency is significantly influenced by performance of individual components. Changes in the design of one component, a turbine for example, often required changes in other components. To allow for the optimization of the overall design with these interdependencies, a detailed steady state and transient control model was developed. The use of the steady state and transient models as a part of an iterative design process represents a key contribution of this work. A dynamic model, MPBRSim, has been developed. The model integrates the reactor core and the power conversion system simultaneously. Physical parameters such as the heat exchangers; weights and practical performance maps such as the turbine characteristics and compressor characteristics are incorporated into the model. The individual component models as well as the fully integrated model of the power conversion system have been verified with an industry-standard general thermal-fluid code Flownet. With respect to the dynamic model, bypass valve control and inventory control have been used as the primary control methods for the power conversion system. By performing simulation using the dynamic model with the designed control scheme, the combination of bypass and inventory control was optimized to assure system stability within design temperature and pressure limits. Bypass control allows for rapid control system response while inventory control allows for ultimate steady state operation at part power very near the optimum operating point for the system. Load transients simulations show that the indirect, three-shaft arrangement gas turbine power conversion system is stable and controllable. For the indirect cycle the intermediate heat exchanger (IHX) is the interface between the reactor and the turbomachinery systems. As a part of the design effort the IHX was identified as the key component in the system. Two technologies, printed circuit and compact plate-fin, were investigated that have the promise of meeting the design requirements for the system. The reference design incorporates the possibility of using either technology although the compact plate-fin design was chosen for subsequent analysis. The thermal design and parametric analysis with an IHX and recuperator using the plate-fin configuration have been performed. As a three-shaft arrangement, the turbo-shaft sets consist of a pair of turbine/compressor sets (high pressure and low pressure turbines with same-shaft compressor) and a power turbine coupled with a synchronous generator. The turbines and compressors are all axial type and the shaft configuration is horizontal. The core outlet/inlet temperatures are 900/520 C, and the optimum pressure ratio in the power conversion cycle is 2.9. The design achieves a plant net efficiency of approximately 48%.

  17. Decommissioning of the Dragon High Temperature Reactor (HTR) Located at the Former United Kingdom Atomic Energy Authority (UKAEA) Research Site at Winfrith - 13180

    SciTech Connect (OSTI)

    Smith, Anthony A. [Research Sites Restoration Ltd, Winfrith, Dorset (United Kingdom)] [Research Sites Restoration Ltd, Winfrith, Dorset (United Kingdom)

    2013-07-01T23:59:59.000Z

    The Dragon Reactor was constructed at the United Kingdom Atomic Energy Research Establishment at Winfrith in Dorset through the late 1950's and into the early 1960's. It was a High Temperature Gas Cooled Reactor (HTR) with helium gas coolant and graphite moderation. It operated as a fuel testing and demonstration reactor at up to 20 MW (Thermal) from 1964 until 1975, when international funding for this project was terminated. The fuel was removed from the core in 1976 and the reactor was put into Safestore. To meet the UK's Nuclear Decommissioning Authority (NDA) objective to 'drive hazard reduction' [1] it is necessary to decommission and remediate all the Research Sites Restoration Ltd (RSRL) facilities. This includes the Dragon Reactor where the activated core, pressure vessel and control rods and the contaminated primary circuit (including a {sup 90}Sr source) still remain. It is essential to remove these hazards at the appropriate time and return the area occupied by the reactor to a safe condition. (author)

  18. Thermal-Hydraulic Analyses of Heat Transfer Fluid Requirements and Characteristics for Coupling A Hydrogen Production Plant to a High-Temperature Nuclear Reactor

    SciTech Connect (OSTI)

    C. B. Davis; C. H. Oh; R. B. Barner; D. F. Wilson

    2005-06-01T23:59:59.000Z

    The Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the hightemperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant, may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood. Seven possible configurations for a system that transfers heat between the nuclear reactor and the hydrogen and/or electrical generation plants were identified. These configurations included both direct and indirect cycles for the production of electricity. Both helium and liquid salts were considered as the working fluid in the intermediate heat transport loop. Methods were developed to perform thermalhydraulic and cycle-efficiency evaluations of the different configurations and coolants. The thermalhydraulic evaluations estimated the sizes of various components in the intermediate heat transport loop for the different configurations. The relative sizes of components provide a relative indication of the capital cost associated with the various configurations. Estimates of the overall cycle efficiency of the various configurations were also determined. The evaluations determined which configurations and coolants are the most promising from thermal-hydraulic and efficiency points of view. These evaluations also determined which configurations and options do not appear to be feasible at the current time.

  19. High Temperatures & Electricity Demand

    E-Print Network [OSTI]

    High Temperatures & Electricity Demand An Assessment of Supply Adequacy in California Trends.......................................................................................................1 HIGH TEMPERATURES AND ELECTRICITY DEMAND.....................................................................................................................7 SECTION I: HIGH TEMPERATURES AND ELECTRICITY DEMAND ..........................9 BACKGROUND

  20. High solids fermentation reactor

    DOE Patents [OSTI]

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-03-02T23:59:59.000Z

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  1. High solids fermentation reactor

    DOE Patents [OSTI]

    Wyman, Charles E. (Lakewood, CO); Grohmann, Karel (Littleton, CO); Himmel, Michael E. (Littleton, CO); Richard, Christopher J. (Lakewood, CO)

    1993-01-01T23:59:59.000Z

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

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

    E-Print Network [OSTI]

    Fisenko, Stanislav

    2010-01-01T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Stanislav Fisenko; Igor Fisenko

    2010-06-27T23:59:59.000Z

    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.

  4. Updating of ASME Nuclear Code Case N-201 to Accommodate the Needs of Metallic Core Support Structures for High Temperature Gas Cooled Reactors Currently in Development

    SciTech Connect (OSTI)

    Mit Basol; John F. Kielb; John F. MuHooly; Kobus Smit

    2007-05-02T23:59:59.000Z

    On September 29, 2005, ASME Standards Technology, LLC (ASME ST-LLC) executed a multi-year, cooperative agreement with the United States DOE for the Generation IV Reactor Materials project. The project's objective is to update and expand appropriate materials, construction, and design codes for application in future Generation IV nuclear reactor systems that operate at elevated temperatures. Task 4 was embarked upon in recognition of the large quantity of ongoing reactor designs utilizing high temperature technology. Since Code Case N-201 had not seen a significant revision (except for a minor revision in September, 2006 to change the SA-336 forging reference for 304SS and 316SS to SA-965 in Tables 1.2(a) and 1.2(b), and some minor editorial changes) since December 1994, identifying recommended updates to support the current high temperature Core Support Structure (CSS) designs and potential new designs was important. As anticipated, the Task 4 effort identified a number of Code Case N-201 issues. Items requiring further consideration range from addressing apparent inconsistencies in definitions and certain material properties between CC-N-201 and Subsection NH, to inclusion of additional materials to provide the designer more flexibility of design. Task 4 developed a design parameter survey that requested input from the CSS designers of ongoing high temperature gas cooled reactor metallic core support designs. The responses to the survey provided Task 4 valuable input to identify the design operating parameters and future needs of the CSS designers. Types of materials, metal temperature, time of exposure, design pressure, design life, and fluence levels were included in the Task 4 survey responses. The results of the survey are included in this report. This research proves that additional work must be done to update Code Case N-201. Task 4 activities provide the framework for the Code Case N-201 update and future work to provide input on materials. Candidate materials such as Type 321 and Type 347 austenitic stainless steels, Modified 9Cr-1Mo steel for core support structure construction, and Alloy 718 for Threaded Structural Fasteners were among the recommended materials for inclusion in the Code Case. This Task 4 Report identifies the need to address design life beyond 3 x 105 hours, especially in consideration of 60-year design life. A proposed update to the latest Code Case N-201 revision (i.e., Code Case N-201-5) including the items resolved in this report is included as Appendix A.

  5. High-temperature gas-cooled reactor safety studies for the division of accident evaluation. Quarterly progress report, October 1-December 31, 1982

    SciTech Connect (OSTI)

    Ball, S.J.; Clapp, N.E. Jr.; Cleveland, J.C.; Conklin, J.C.; Harrington, R.M.; Lindemer, T.B.; Siman-Tov, I.

    1983-08-01T23:59:59.000Z

    Work continued on high-temperature gas-cooled reactor safety code development, including both the Fort St. Vrain and the 2240-MW(t) lead plant versions of the three-dimensional core code ORECA, the BLAST steam generator code, and a simplified core model code called SCORE. Oak Ridge National Laboratory participated in the Nuclear Regulatory Commission siting study for the lead plant with three other laboratories. Investigations continued to determine the status of fission-product source-term methodology applicable to postulated severe accidents.

  6. A high-temperature, ambient-pressure ultra-dry operando reactor cell for Fourier-transform infrared spectroscopy

    SciTech Connect (OSTI)

    Kck, Eva-Maria; Kogler, Michaela; Pramsoler, Reinhold; Kltzer, Bernhard; Penner, Simon, E-mail: simon.penner@uibk.ac.at [Institute of Physical Chemistry, University of Innsbruck, Innrain 80-82, A-6020 Innsbruck (Austria)

    2014-08-15T23:59:59.000Z

    The construction of a newly designed high-temperature, high-pressure FT-IR reaction cell for ultra-dry in situ and operando operation is reported. The reaction cell itself as well as the sample holder is fully made of quartz glass, with no hot metal or ceramic parts in the vicinity of the high-temperature zone. Special emphasis was put on chemically absolute water-free and inert experimental conditions, which includes reaction cell and gas-feeding lines. Operation and spectroscopy up to 1273 K is possible, as well as pressures up to ambient conditions. The reaction cell exhibits a very easy and variable construction and can be adjusted to any available FT-IR spectrometer. Its particular strength lies in its possibility to access and study samples under very demanding experimental conditions. This includes studies at very high temperatures, e.g., for solid-oxide fuel cell research or studies where the water content of the reaction mixtures must be exactly adjusted. The latter includes all adsorption studies on oxide surfaces, where the hydroxylation degree is of paramount importance. The capability of the reaction cell will be demonstrated for two selected examples where information and in due course a correlation to other methods can only be achieved using the presented setup.

  7. EXTENDING SODIUM FAST REACTOR DRIVER FUEL USE TO HIGHER TEMPERATURES

    SciTech Connect (OSTI)

    Douglas L. Porter

    2011-02-01T23:59:59.000Z

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

  8. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980

    SciTech Connect (OSTI)

    Not Available

    1981-08-01T23:59:59.000Z

    Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies.

  9. The design of a functionally graded composite for service in high temperature lead and lead-bismuth cooled nuclear reactors

    E-Print Network [OSTI]

    Short, Michael Philip

    2010-01-01T23:59:59.000Z

    A material that resists lead-bismuth eutectic (LBE) attack and retains its strength at 700C would be an enabling technology for LBE-cooled reactors. No single alloy currently exists that can economically meet the required ...

  10. Sequential high temperature reduction, low temperature hydrolysis...

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

    high temperature reduction, low temperature hydrolysis for the regeneration of sulfated NOx trap catalysts. Sequential high temperature reduction, low temperature hydrolysis for...

  11. Fission Product Impact Reduction via Protracted In-core Retention in Very High Temperature Reactor (VHTR) Transmutation Scenarios

    E-Print Network [OSTI]

    Alajo, Ayodeji Babatunde

    2011-08-08T23:59:59.000Z

    is performed, the spent fuel can be partitioned and separated into 3 streams: depleted uranium (to be recycled with plutonium in reactors), TRU and FP. The TRU content of spent fuel is potentially a useable material. TRU can be recycled in advanced reactors... percent depleted uranium and 1.1 percent higher actinides [25]. Based on the 4.6w/o fission product content, it can be estimated that 10GWd/MTU burnup corresponds to about 1.0w/o of fission products in the spent fuel. Given the burnup of U.S. legacy...

  12. Gas-cooled reactor programs. High-temperature gas-cooled reactor technology development program. Annual progress report, December 31, 1983

    SciTech Connect (OSTI)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1984-06-01T23:59:59.000Z

    ORNL continues to make significant contributions to the national program. In the HTR fuels area, we are providing detailed statistical information on the fission product retention performance of irradiated fuel. Our studies are also providing basic data on the mechanical, physical, and chemical behavior of HTR materials, including metals, ceramics, graphite, and concrete. The ORNL has an important role in the development of improved HTR graphites and in the specification of criteria that need to be met by commercial products. We are also developing improved reactor physics design methods. Our work in component development and testing centers in the Component Flow Test Loop (CFTL), which is being used to evaluate the performance of the HTR core support structure. Other work includes experimental evaluation of the shielding effectiveness of the lower portions of an HTR core. This evaluation is being performed at the ORNL Tower Shielding Facility. Researchers at ORNL are developing welding techniques for attaching steam generator tubing to the tubesheets and are testing ceramic pads on which the core posts rest. They are also performing extensive testing of aggregate materials obtained from potential HTR site areas for possible use in prestressed concrete reactor vessels. During the past year we continued to serve as a peer reviewer of small modular reactor designs being developed by GA and GE with balance-of-plant layouts being developed by Bechtel Group, Inc. We have also evaluated the national need for developing HTRs with emphasis on the longer term applications of the HTRs to fossil conversion processes.

  13. Thermal Hydraulic Analysis of a Reduced Scale High Temperature Gas-Cooled Reactor Test Facility and its Prototype with MELCOR

    E-Print Network [OSTI]

    Beeny, Bradley Aaron 1988-

    2012-11-12T23:59:59.000Z

    ................ 86 6.4 Area-averaged outer RPV wall temperature during PCC .............................. 88 6.5 Mass-averaged (by ring) core graphite temperatures during DCC ................ 89 6.6 Mass-averaged (by level) core... graphite temperatures during DCC ............... 91 6.7 Area-averaged outer RPV wall temperature during DCC .............................. 92 6.8 Steady-state core structural temperature distribution...

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

    E-Print Network [OSTI]

    Gitau, Ernest Travis Ngure

    2012-10-19T23:59:59.000Z

    the world, adequate methods for safeguarding the reactor must be developed. Current safeguards methods for the pebble-fueled HTGR focus on extensive, redundant containment and surveillance (C/S) measures or a combination of item-type and bulk-type material...

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

    SciTech Connect (OSTI)

    Phillpot, Simon; Tulenko, James

    2011-09-08T23:59:59.000Z

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

  16. TRITIUM PERMEATION AND TRANSPORT IN THE GASOLINE PRODUCTION SYSTEM COUPLED WITH HIGH TEMPERATURE GAS-COOLED REACTORS (HTGRS)

    SciTech Connect (OSTI)

    Chang H. Oh; Eung S. Kim; Mike Patterson

    2011-05-01T23:59:59.000Z

    This paper describes scoping analyses on tritium behaviors in the HTGR-integrated gasoline production system, which is based on a methanol-to-gasoline (MTG) plant. In this system, the HTGR transfers heat and electricity to the MTG system. This system was analyzed using the TPAC code, which was recently developed by Idaho National Laboratory. The global sensitivity analyses were performed to understand and characterize tritium behaviors in the coupled HTGR/MTG system. This Monte Carlo based random sampling method was used to evaluate maximum 17,408 numbers of samples with different input values. According to the analyses, the average tritium concentration in the product gasoline is about 3.0510-3 Bq/cm3, and 62 % cases are within the tritium effluent limit (= 3.7x10-3 Bq/cm3[STP]). About 0.19% of released tritium is finally transported from the core to the gasoline product through permeations. This study also identified that the following four parameters are important concerning tritium behaviors in the HTGR/MTG system: (1) tritium source, (2) wall thickness of process heat exchanger, (3) operating temperature, and (4) tritium permeation coefficient of process heat exchanger. These four parameters contribute about 95 % of the total output uncertainties. This study strongly recommends focusing our future research on these four parameters to improve modeling accuracy and to mitigate tritium permeation into the gasol ine product. If the permeation barrier is included in the future study, the tritium concentration will be significantly reduced.

  17. NSTX High Temperature Sensor Systems

    SciTech Connect (OSTI)

    B.McCormack; H.W. Kugel; P. Goranson; R. Kaita; et al

    1999-11-01T23:59:59.000Z

    The design of the more than 300 in-vessel sensor systems for the National Spherical Torus Experiment (NSTX) has encountered several challenging fusion reactor diagnostic issues involving high temperatures and space constraints. This has resulted in unique miniature, high temperature in-vessel sensor systems mounted in small spaces behind plasma facing armor tiles, and they are prototypical of possible high power reactor first-wall applications. In the Center Stack, Divertor, Passive Plate, and vessel wall regions, the small magnetic sensors, large magnetic sensors, flux loops, Rogowski Coils, thermocouples, and Langmuir Probes are qualified for 600 degrees C operation. This rating will accommodate both peak rear-face graphite tile temperatures during operations and the 350 degrees C bake-out conditions. Similar sensor systems including flux loops, on other vacuum vessel regions are qualified for 350 degrees C operation. Cabling from the sensors embedded in the graphite tiles follows narrow routes to exit the vessel. The detailed sensor design and installation methods of these diagnostic systems developed for high-powered ST operation are discussed.

  18. High Temperature Heat Exchanger Project

    SciTech Connect (OSTI)

    Anthony E. Hechanova, Ph.D.

    2008-09-30T23:59:59.000Z

    The UNLV Research Foundation assembled a research consortium for high temperature heat exchanger design and materials compatibility and performance comprised of university and private industry partners under the auspices of the US DOE-NE Nuclear Hydrogen Initiative in October 2003. The objectives of the consortium were to conduct investigations of candidate materials for high temperature heat exchanger componets in hydrogen production processes and design and perform prototypical testing of heat exchangers. The initial research of the consortium focused on the intermediate heat exchanger (located between the nuclear reactor and hydrogen production plan) and the components for the hydrogen iodine decomposition process and sulfuric acid decomposition process. These heat exchanger components were deemed the most challenging from a materials performance and compatibility perspective

  19. Molten Salts for High Temperature Reactors: University of Wisconsin Molten Salt Corrosion and Flow Loop Experiments -- Issues Identified and Path Forward

    SciTech Connect (OSTI)

    Piyush Sabharwall; Matt Ebner; Manohar Sohal; Phil Sharpe; Thermal Hydraulics Group

    2010-03-01T23:59:59.000Z

    Considerable amount of work is going on regarding the development of high temperature liquid salts technology to meet future process needs of Next Generation Nuclear Plant. This report identifies the important characteristics and concerns of high temperature molten salts (with lesson learned at University of Wisconsin-Madison, Molten Salt Program) and provides some possible recommendation for future work

  20. Temperature control method for series-connected reactors

    SciTech Connect (OSTI)

    Abrams, L.M.

    1984-07-03T23:59:59.000Z

    A method is claimed for controlling the temperature and composition of a vapor feedstream into a second reactor connected in series flow arrangement with a first reactor. The effluent stream from the first reactor containing vapor and liquid fractions is first cooled against a vapor stream and then further cooled against a suitable external fluid, then is phase separated to provide vapor and liquid fractions. The separated vapor fraction is reheated against the first reactor effluent stream and passed at an intermediate temperature into the second reactor. The first reactor is preferably an ebullated bed type catalytic reactor and the second reactor is preferably a fixed bed type catalytic reactor which is operated at an inlet temperature 20/sup 0/-200/sup 0/ F. lower than the first reactor effluent stream temperature. If desired, the effluent stream from the first reactor can be initially phase separated into vapor and liquid factions, and the vapor fraction only passed to the first heat exchange step for cooling to a first lower temperature.

  1. BROOKHAVEN NATIONAL LABORATORY'S HIGH FLUX BEAM REACTOR

    E-Print Network [OSTI]

    Ohta, Shigemi

    1 BROOKHAVEN NATIONAL LABORATORY'S HIGH FLUX BEAM REACTOR Compiled by S. M. Shapiro I. PICTORIAL with fiberglass insulation and a protective aluminum skin. The reactor vessel is shaped somewhat like a very large at the spherical end. It is located at the center of the reactor building and is surrounded by a lead and steel

  2. High-Temperature Superconductivity

    ScienceCinema (OSTI)

    Peter Johnson

    2010-01-08T23:59:59.000Z

    Like astronomers tweaking images to gain a more detailed glimpse of distant stars, physicists at Brookhaven National Laboratory have found ways to sharpen images of the energy spectra in high-temperature superconductors ? materials that carry electrical c

  3. High temperature pressure gauge

    DOE Patents [OSTI]

    Echtler, J. Paul (Pittsburgh, PA); Scandrol, Roy O. (Library, PA)

    1981-01-01T23:59:59.000Z

    A high temperature pressure gauge comprising a pressure gauge positioned in fluid communication with one end of a conduit which has a diaphragm mounted in its other end. The conduit is filled with a low melting metal alloy above the diaphragm for a portion of its length with a high temperature fluid being positioned in the remaining length of the conduit and in the pressure gauge.

  4. High Temperature Capacitor Development

    SciTech Connect (OSTI)

    John Kosek

    2009-06-30T23:59:59.000Z

    The absence of high-temperature electronics is an obstacle to the development of untapped energy resources (deep oil, gas and geothermal). US natural gas consumption is projected to grow from 22 trillion cubic feet per year (tcf) in 1999 to 34 tcf in 2020. Cumulatively this is 607 tcf of consumption by 2020, while recoverable reserves using current technology are 177 tcf. A significant portion of this shortfall may be met by tapping deep gas reservoirs. Tapping these reservoirs represents a significant technical challenge. At these depths, temperatures and pressures are very high and may require penetrating very hard rock. Logistics of supporting 6.1 km (20,000 ft) drill strings and the drilling processes are complex and expensive. At these depths up to 50% of the total drilling cost may be in the last 10% of the well depth. Thus, as wells go deeper it is increasingly important that drillers are able to monitor conditions down-hole such as temperature, pressure, heading, etc. Commercial off-the-shelf electronics are not specified to meet these operating conditions. This is due to problems associated with all aspects of the electronics including the resistors and capacitors. With respect to capacitors, increasing temperature often significantly changes capacitance because of the strong temperature dependence of the dielectric constant. Higher temperatures also affect the equivalent series resistance (ESR). High-temperature capacitors usually have low capacitance values because of these dielectric effects and because packages are kept small to prevent mechanical breakage caused by thermal stresses. Electrolytic capacitors do not operate at temperatures above 150oC due to dielectric breakdown. The development of high-temperature capacitors to be used in a high-pressure high-temperature (HPHT) drilling environment was investigated. These capacitors were based on a previously developed high-voltage hybridized capacitor developed at Giner, Inc. in conjunction with a unique high-temperature electrolyte developed during the course of the program. During this program the feasibility of operating a high voltage hybridized capacitor at 230oC was demonstrated. Capacitor specifications were established in conjunction with potential capacitor users. A method to allow for capacitor operation at both ambient and elevated temperatures was demonstrated. The program was terminated prior to moving into Phase II due to a lack of cost-sharing funds.

  5. Thermal hydraulic design of a salt-cooled highly efficient environmentally friendly reactor

    E-Print Network [OSTI]

    Whitman, Joshua (Joshua J.)

    2009-01-01T23:59:59.000Z

    A 1 OOOMWth liquid-salt cooled thermal spectrum reactor was designed with a long fuel cycle, and high core exit temperature. These features are desirable in a reactor designed to provide process heat applications such as ...

  6. OPERATING TEMPERATURE WINDOWS FOR FUSION REACTOR STRUCTURAL MATERIALS

    E-Print Network [OSTI]

    California at Los Angeles, University of

    OPERATING TEMPERATURE WINDOWS FOR FUSION REACTOR STRUCTURAL MATERIALS S.J. Zinkle1 and N.M. Ghoniem reactor structural materials: four reduced-activation structural materials (oxide-dispersion- strengthened operating temperature limit of structural materials is determined by one of four factors, all of which

  7. High-temperature Pump Monitoring - High-temperature ESP Monitoring...

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

    7 4.4.4 High-temperature Pump Monitoring - High-temperature ESP Monitoring Presentation Number: 018 Investigator: Dhruva, Brindesh (Schlumberger Technology Corp.) Objectives: To...

  8. High temperature storage battery

    SciTech Connect (OSTI)

    Sammells, A.F.

    1988-06-07T23:59:59.000Z

    A high temperature electrochemical cell is described comprising: a solid-state divalent cation conducting electrolyte; a positive electrode in contact with the electrolyte; a solid-state negative electrode contacting a divalent cation conducting molten salt mediating agent providing ionic mediation between the solid-state negative electrode and the solid-state electrolyte.

  9. Nuclear fuels for very high temperature applications

    SciTech Connect (OSTI)

    Lundberg, L.B.; Hobbins, R.R.

    1992-08-01T23:59:59.000Z

    The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO{sub 2} or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures.

  10. Nuclear fuels for very high temperature applications

    SciTech Connect (OSTI)

    Lundberg, L.B.; Hobbins, R.R.

    1992-01-01T23:59:59.000Z

    The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO{sub 2} or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures.

  11. High temperature thermometric phosphors

    DOE Patents [OSTI]

    Allison, Stephen W. (Knoxville, TN); Cates, Michael R. (Oak Ridge, TN); Boatner, Lynn A. (Oak Ridge, TN); Gillies, George T. (Earlysville, VA)

    1999-03-23T23:59:59.000Z

    A high temperature phosphor consists essentially of a material having the general formula LuPO.sub.4 :Dy.sub.(x),Eu.sub.y) wherein: 0.1 wt %.ltoreq.x.ltoreq.20 wt % and 0.1 wt %.ltoreq.y.ltoreq.20 wt %. The high temperature phosphor is in contact with an article whose temperature is to be determined. The article having the phosphor in contact with it is placed in the environment for which the temperature of the article is to be determined. The phosphor is excited by a laser causing the phosphor to fluoresce. The emission from the phosphor is optically focused into a beam-splitting mirror which separates the emission into two separate emissions, the emission caused by the dysprosium dopant and the emission caused by the europium dopent. The separated emissions are optically filtered and the intensities of the emission are detected and measured. The ratio of the intensity of each emission is determined and the temperature of the article is calculated from the ratio of the intensities of the separate emissions.

  12. High temperature thermometric phosphors

    DOE Patents [OSTI]

    Allison, S.W.; Cates, M.R.; Boatner, L.A.; Gillies, G.T.

    1999-03-23T23:59:59.000Z

    A high temperature phosphor consists essentially of a material having the general formula LuPO{sub 4}:Dy{sub x},Eu{sub y} wherein: 0.1 wt % {<=} x {<=} 20 wt % and 0.1 wt % {<=} y {<=} 20 wt %. The high temperature phosphor is in contact with an article whose temperature is to be determined. The article having the phosphor in contact with it is placed in the environment for which the temperature of the article is to be determined. The phosphor is excited by a laser causing the phosphor to fluoresce. The emission from the phosphor is optically focused into a beam-splitting mirror which separates the emission into two separate emissions, the emission caused by the dysprosium dopant and the emission caused by the europium dopant. The separated emissions are optically filtered and the intensities of the emission are detected and measured. The ratio of the intensity of each emission is determined and the temperature of the article is calculated from the ratio of the intensities of the separate emissions. 2 figs.

  13. High temperature adsorption measurements

    SciTech Connect (OSTI)

    Bertani, R.; Parisi, L.; Perini, R.; Tarquini, B.

    1996-01-24T23:59:59.000Z

    Adsorption phenomena are a rich and rather new field of study in geothermal research, in particular at very high temperature. ENEL is interested in the exploitation of geothermal regions with superheated steam, and it is important to understand the behavior of water-rock interaction. We have analyzed in the 170-200 C temperature range four samples of Monteverdi cuttings; the next experimental effort will be at 220 C and over in 1996. The first results of the 1995 runs are collected in this paper. We can highlight four main items: 1. At relative pressures over 0.6 the capillarity forces are very important. 2. There is no significant temperature effect. 3. Adsorbed water can be present, and it is able to multiply by a factor of 15 the estimated reserve of super-heated steam only. 4. Pores smaller than 15 do not contribute to the adsorbed mass.

  14. High Temperature Membrane Working Group

    Broader source: Energy.gov [DOE]

    This presentation provides an overview of the High Temperature Membrane Working Group Meeting in May 2007.

  15. RECHARGEABLE HIGH-TEMPERATURE BATTERIES

    E-Print Network [OSTI]

    Cairns, Elton J.

    2014-01-01T23:59:59.000Z

    F. Eshman, High-Performance Batteries for Electric-VehicleS. Sudar, High Performance Batteries for Electric-VehicleHIGH-TEMPERATURE BATTERIES Elton J. Cairns January 1981 TWO-

  16. Fabrication and Characterization of Uranium-based High Temperature...

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

    Fabrication and Characterization of Uranium-based High Temperature Reactor Fuel June 01, 2013 The Uranium Fuel Development Laboratory is a modern R&D scale lab for the fabrication...

  17. Assessment of the Use of Nitrogen Trifluoride for Purifying Coolant and Heat Transfer Salts in the Fluoride Salt-Cooled High-Temperature Reactor

    SciTech Connect (OSTI)

    Scheele, Randall D.; Casella, Andrew M.

    2010-09-28T23:59:59.000Z

    This report provides an assessment of the use of nitrogen trifluoride for removing oxide and water-caused contaminants in the fluoride salts that will be used as coolants in a molten salt cooled reactor.

  18. High temperature interfacial superconductivity

    DOE Patents [OSTI]

    Bozovic, Ivan (Mount Sinai, NY); Logvenov, Gennady (Port Jefferson Station, NY); Gozar, Adrian Mihai (Port Jefferson, NY)

    2012-06-19T23:59:59.000Z

    High-temperature superconductivity confined to nanometer-scale interfaces has been a long standing goal because of potential applications in electronic devices. The spontaneous formation of a superconducting interface in bilayers consisting of an insulator (La.sub.2CuO.sub.4) and a metal (La.sub.1-xSr.sub.xCuO.sub.4), neither of which is superconducting per se, is described. Depending upon the layering sequence of the bilayers, T.sub.c may be either .about.15 K or .about.30 K. This highly robust phenomenon is confined to within 2-3 nm around the interface. After exposing the bilayer to ozone, T.sub.c exceeds 50 K and this enhanced superconductivity is also shown to originate from a 1 to 2 unit cell thick interfacial layer. The results demonstrate that engineering artificial heterostructures provides a novel, unconventional way to fabricate stable, quasi two-dimensional high T.sub.c phases and to significantly enhance superconducting properties in other superconductors. The superconducting interface may be implemented, for example, in SIS tunnel junctions or a SuFET.

  19. High-Temperature-High-Volume Lifting for Enhanced Geothermal...

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

    High-Temperature-High-Volume Lifting for Enhanced Geothermal Systems High-Temperature-High-Volume Lifting for Enhanced Geothermal Systems High-Temperature-High-Volume Lifting for...

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

    SciTech Connect (OSTI)

    Simos, N.

    2011-05-01T23:59:59.000Z

    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.

  1. Philosophy 26 High Temperature Superconductivity

    E-Print Network [OSTI]

    Callender, Craig

    Philosophy 26 High Temperature Superconductivity By Ohm's Law, resistance will dim. Low temperature superconductivity was discovered in 1911 by Heike was explained by BCS theory. BCS theory explains superconductivity microscopically

  2. High Temperature, High Pressure Devices for Zonal Isolation in...

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

    High Temperature, High Pressure Devices for Zonal Isolation in Geothermal Wells High Temperature, High Pressure Devices for Zonal Isolation in Geothermal Wells High Temperature,...

  3. Hotline IV ?High Temperature ESP

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

    Hotline IV - High Temperature ESP Brindesh Dhruva (principal Inv.) Michael Dowling (presenter) Schlumberger Track Name May 18, 2010 This presentation does not contain any...

  4. High Temperature Materials Interim Data Qualification Report

    SciTech Connect (OSTI)

    Nancy Lybeck

    2010-08-01T23:59:59.000Z

    ABSTRACT Projects for the very high temperature reactor (VHTR) Technology Development Office provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are qualified for use, stored in a readily accessible electronic form, and analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the High Temperature Materials characterization data stream, the processing of these data within NDMAS, and reports the interim FY2010 qualification status of the data. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category assigned by the data generator. The High Temperature Materials data are being collected under NQA-1 guidelines, and will be qualified data. For NQA-1 qualified data, the qualification activities include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing to confirm that the data are an accurate representation of the system or object being measured, and (3) documenting that the data were collected under an NQA-1 or equivalent Quality Assurance program. Currently, data from two test series within the High Temperature Materials data stream have been entered into the NDMAS vault: 1. Tensile Tests for Sm (i.e., Allowable Stress) Confirmatory Testing 1,403,994 records have been inserted into the NDMAS database. Capture testing is in process. 2. Creep-Fatigue Testing to Support Determination of Creep-Fatigue Interaction Diagram 918,854 records have been processed and inserted into the NDMAS database. Capture testing is in process.

  5. Technologies for Upgrading Light Water Reactor Outlet Temperature

    SciTech Connect (OSTI)

    Daniel S. Wendt; Piyush Sabharwall; Vivek Utgikar

    2013-07-01T23:59:59.000Z

    Nuclear energy could potentially be utilized in hybrid energy systems to produce synthetic fuels and feedstocks from indigenous carbon sources such as coal and biomass. First generation nuclear hybrid energy system (NHES) technology will most likely be based on conventional light water reactors (LWRs). However, these LWRs provide thermal energy at temperatures of approximately 300C, while the desired temperatures for many chemical processes are much higher. In order to realize the benefits of nuclear hybrid energy systems with the current LWR reactor fleets, selection and development of a complimentary temperature upgrading technology is necessary. This paper provides an initial assessment of technologies that may be well suited toward LWR outlet temperature upgrading for powering elevated temperature industrial and chemical processes during periods of off-peak power demand. Chemical heat transformers (CHTs) are a technology with the potential to meet LWR temperature upgrading requirements for NHESs. CHTs utilize chemical heat of reaction to change the temperature at which selected heat sources supply or consume thermal energy. CHTs could directly utilize LWR heat output without intermediate mechanical or electrical power conversion operations and the associated thermodynamic losses. CHT thermal characteristics are determined by selection of the chemical working pair and operating conditions. This paper discusses the chemical working pairs applicable to LWR outlet temperature upgrading and the CHT operating conditions required for providing process heat in NHES applications.

  6. High-temperature ceramic receivers

    SciTech Connect (OSTI)

    Jarvinen, P. O.

    1980-01-01T23:59:59.000Z

    An advanced ceramic dome cavity receiver is discussed which heats pressurized gas to temperatures above 1800/sup 0/F (1000/sup 0/C) for use in solar Brayton power systems of the dispersed receiver/dish or central receiver type. Optical, heat transfer, structural, and ceramic material design aspects of the receiver are reported and the development and experimental demonstration of a high-temperature seal between the pressurized gas and the high-temperature silicon carbide dome material is described.

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

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

    High Flux Isotope Reactor named Nuclear Historic Landmark The High Flux Isotope Reactor vessel at Oak Ridge National Laboratory resides in a pool of water illuminated by the blue...

  8. Calculation of Heating Values for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Peterson, Joshua L [ORNL] [ORNL; Ilas, Germina [ORNL] [ORNL

    2012-01-01T23:59:59.000Z

    Calculating the amount of energy released by a fission reaction (fission Q value) and the heating rate distribution in a nuclear reactor is an important part of the safety analysis. However, these calculations can become very complex. One of the codes that can be used for this type of analyses is the Monte Carlo transport code MCNP5. Currently it is impossible to calculate the Q value and heating rate disposition for delayed beta and delayed gamma particles directly from MCNP5. The purpose of this paper is to outline a rigorous method for indirectly calculating the Q values and heating rates in the High Flux Isotope Reactor (HFIR), based on previous similar studies carried out for very high-temperature reactor configurations. This method has been applied in this study to calculate heating rates for the beginning of cycle (BOC) and end-of-cycle (EOC) states of HFIR. In addition, the BOC results obtained for HFIR are compared with corresponding results for the Advanced Test Reactor. The fission Q value for HFIR was calculated as 200.2 MeV for the BOC and 201.3 MeV for the EOC. It was also determined that 95.1% and 95.4% of the heat was deposited within the HFIR fuel plates for the BOC and EOC models, respectively. This methodology can also be used for heating rate calculations for HFIR experiments.

  9. Model biases in high-burnup fast reactor simulations

    SciTech Connect (OSTI)

    Touran, N.; Cheatham, J.; Petroski, R. [TerraPower LLC, 11235 S.E. 6th St, Bellevue, WA 98004 (United States)

    2012-07-01T23:59:59.000Z

    A new code system called the Advanced Reactor Modeling Interface (ARMI) has been developed that loosely couples multiscale, multiphysics nuclear reactor simulations to provide rapid, user-friendly, high-fidelity full systems analysis. Incorporating neutronic, thermal-hydraulic, safety/transient, fuel performance, core mechanical, and economic analyses, ARMI provides 'one-click' assessments of many multi-disciplined performance metrics and constraints that historically require iterations between many diverse experts. The capabilities of ARMI are implemented in this study to quantify neutronic biases of various modeling approximations typically made in fast reactor analysis at an equilibrium condition, after many repetitive shuffles. Sensitivities at equilibrium that result in very high discharge burnup are considered ( and >20% FIMA), as motivated by the development of the Traveling Wave Reactor. Model approximations discussed include homogenization, neutronic and depletion mesh resolution, thermal-hydraulic coupling, explicit control rod insertion, burnup-dependent cross sections, fission product model, burn chain truncation, and dynamic fuel performance. The sensitivities of these approximations on equilibrium discharge burnup, k{sub eff}, power density, delayed neutron fraction, and coolant temperature coefficient are discussed. (authors)

  10. High temperature structural insulating material

    DOE Patents [OSTI]

    Chen, W.Y.

    1984-07-27T23:59:59.000Z

    A high temperature structural insulating material useful as a liner for cylinders of high temperature engines through the favorable combination of high service temperature (above about 800/sup 0/C), low thermal conductivity (below about 0.2 W/m/sup 0/C), and high compressive strength (above about 250 psi). The insulating material is produced by selecting hollow ceramic beads with a softening temperature above about 800/sup 0/C, a diameter within the range of 20-200 ..mu..m, and a wall thickness in the range of about 2 to 4 ..mu..m; compacting the beads and a compatible silicate binder composition under pressure and sintering conditions to provide the desired structural form with the structure having a closed-cell, compact array of bonded beads.

  11. High temperature structural insulating material

    DOE Patents [OSTI]

    Chen, Wayne Y. (Munster, IN)

    1987-01-01T23:59:59.000Z

    A high temperature structural insulating material useful as a liner for cylinders of high temperature engines through the favorable combination of high service temperature (above about 800.degree. C.), low thermal conductivity (below about 0.2 W/m.degree. C.), and high compressive strength (above about 250 psi). The insulating material is produced by selecting hollow ceramic beads with a softening temperature above about 800.degree. C., a diameter within the range of 20-200 .mu.m, and a wall thickness in the range of about 2-4 .mu.m; compacting the beads and a compatible silicate binder composition under pressure and sintering conditions to provide the desired structural form with the structure having a closed-cell, compact array of bonded beads.

  12. Method of nuclear reactor control using a variable temperature load dependent set point

    SciTech Connect (OSTI)

    Kelly, J.J.; Rambo, G.E.

    1982-04-27T23:59:59.000Z

    A method and apparatus for controlling a nuclear reactor in response to a variable average reactor coolant temperature set point is disclosed. The set point is dependent upon percent of full power load demand. A manually-actuated ''droop mode'' of control is provided whereby the reactor coolant temperature is allowed to drop below the set point temperature a predetermined amount wherein the control is switched from reactor control rods exclusively to feedwater flow.

  13. Thorium Molten Salt Reactor : from high breeding to simplified reprocessing

    E-Print Network [OSTI]

    Paris-Sud XI, Universit de

    Thorium Molten Salt Reactor : from high breeding to simplified reprocessing L. Mathieu, D. Heuer, A- ceptable. The Thorium Molten Salt Reactor (TMSR) may contribute to solve these problems. The thorium cycle

  14. Cross section generation strategy for high conversion light water reactors

    E-Print Network [OSTI]

    Herman, Bryan R. (Bryan Robert)

    2011-01-01T23:59:59.000Z

    High conversion water reactors (HCWR), such as the Resource-renewable Boiling Water Reactor (RBWR), are being designed with axial heterogeneity of alternating fissile and blanket zones to achieve a conversion ratio of ...

  15. Temperature controlled high voltage regulator

    DOE Patents [OSTI]

    Chiaro, Jr., Peter J. (Clinton, TN); Schulze, Gerald K. (Knoxville, TN)

    2004-04-20T23:59:59.000Z

    A temperature controlled high voltage regulator for automatically adjusting the high voltage applied to a radiation detector is described. The regulator is a solid state device that is independent of the attached radiation detector, enabling the regulator to be used by various models of radiation detectors, such as gas flow proportional radiation detectors.

  16. High temperature lightweight foamed cements

    DOE Patents [OSTI]

    Sugama, Toshifumi (Mastic Beach, NY)

    1989-01-01T23:59:59.000Z

    Cement slurries are disclosed which are suitable for use in geothermal wells since they can withstand high temperatures and high pressures. The formulation consists of cement, silica flour, water, a retarder, a foaming agent, a foam stabilizer, and a reinforcing agent. A process for producing these cements is also disclosed.

  17. High temperature lightweight foamed cements

    DOE Patents [OSTI]

    Sugama, Toshifumi.

    1989-10-03T23:59:59.000Z

    Cement slurries are disclosed which are suitable for use in geothermal wells since they can withstand high temperatures and high pressures. The formulation consists of cement, silica flour, water, a retarder, a foaming agent, a foam stabilizer, and a reinforcing agent. A process for producing these cements is also disclosed. 3 figs.

  18. High temperature Seebeck coefficient metrology

    SciTech Connect (OSTI)

    Martin, J. [Materials Science and Engineering Laboratory, National Institute of Standards and Technology, Gaithersburg, Maryland 20899 (United States); Tritt, T. [Department of Physics and Astronomy, Clemson University, Clemson, South Carolina 29634 (United States); Uher, C. [Department of Physics, University of Michigan, Ann Arbor, Michigan 48109 (United States)

    2010-12-15T23:59:59.000Z

    We present an overview of the challenges and practices of thermoelectric metrology on bulk materials at high temperature (300 to 1300 K). The Seebeck coefficient, when combined with thermal and electrical conductivity, is an essential property measurement for evaluating the potential performance of novel thermoelectric materials. However, there is some question as to which measurement technique(s) provides the most accurate determination of the Seebeck coefficient at high temperature. This has led to the implementation of nonideal practices that have further complicated the confirmation of reported high ZT materials. To ensure meaningful interlaboratory comparison of data, thermoelectric measurements must be reliable, accurate, and consistent. This article will summarize and compare the relevant measurement techniques and apparatus designs required to effectively manage uncertainty, while also providing a reference resource of previous advances in high temperature thermoelectric metrology.

  19. High temperature Seebeck coefficient metrology

    SciTech Connect (OSTI)

    Martin, J.; Tritt, T.; Uher, Ctirad

    2010-01-01T23:59:59.000Z

    We present an overview of the challenges and practices of thermoelectric metrology on bulk materials at high temperature (300 to 1300 K). The Seebeck coefficient, when combined with thermal and electrical conductivity, is an essential propertymeasurement for evaluating the potential performance of novel thermoelectricmaterials. However, there is some question as to which measurement technique(s) provides the most accurate determination of the Seebeck coefficient at high temperature. This has led to the implementation of nonideal practices that have further complicated the confirmation of reported high ZT materials. To ensure meaningful interlaboratory comparison of data, thermoelectricmeasurements must be reliable, accurate, and consistent. This article will summarize and compare the relevant measurement techniques and apparatus designs required to effectively manage uncertainty, while also providing a reference resource of previous advances in high temperature thermoelectric metrology.

  20. Robust controller design for temperature tracking problems in jacketed batch reactors

    E-Print Network [OSTI]

    Palanki, Srinivas

    Robust controller design for temperature tracking problems in jacketed batch reactors Vishak for temperature tracking problems in batch reactors in the presence of parametric uncertainty. The controller has]. Control is achieved by manipulating the heat content from the jacket to the reactor. In the past

  1. NGNP/HTE full-power operation at reduced high-temperature heat exchanger temperatures.

    SciTech Connect (OSTI)

    VIlim, R.; Nuclear Engineering Division

    2009-03-12T23:59:59.000Z

    Operation of the Next Generation Nuclear Plant (NGNP) with reduced reactor outlet temperature at full power was investigated for the High Temperature Electrolysis (HTE) hydrogen-production application. The foremost challenge for operation at design temperature is achieving an acceptably long service life for heat exchangers. In both the Intermediate Heat Exchanger (IHX) and the Process Heat Exchanger (PHX) (referred to collectively as high temperature heat exchangers) a pressure differential of several MPa exists with temperatures at or above 850 C. Thermal creep of the heat exchanger channel wall may severely limit heat exchanger life depending on the alloy selected. This report investigates plant performance with IHX temperatures reduced by lowering reactor outlet temperature. The objective is to lower the temperature in heat transfer channels to the point where existing materials can meet the 40 year lifetime needed for this component. A conservative estimate for this temperature is believed to be about 700 C. The reactor outlet temperature was reduced from 850 C to 700 C while maintaining reactor power at 600 MWt and high pressure compressor outlet at 7 MPa. We included a previously reported design option for reducing temperature at the PHX. Heat exchanger lengths were adjusted to reflect the change in performance resulting from coolant property changes and from resizing related to operating-point change. Turbomachine parameters were also optimized for the new operating condition. An integrated optimization of the complete system including heat transfer equipment was not performed. It is estimated, however, that by performing a pinch analysis the combined plant efficiency can be increased from 35.5 percent obtained in this report to a value between 38.5 and 40.1 percent. Then after normalizing for a more than three percent decrease in commodities inventory compared to the reference plant, the commodities-normalized efficiency lies between 40.0 and 41.3. This compares with a value of 43.9 for the reference plant. This latter plant has a reactor outlet temperature of 850 C and the two high temperature heat exchangers. The reduction in reactor outlet temperature from 850 C to 700 C reduces the tritium permeability rate in the IHX metal by a factor of three and thermal creep by five orders of magnitude. The design option for reducing PHX temperature from 800 C to 200 C reduces the permeability there by three orders of magnitude. In that design option this heat exchanger is the single 'choke-point' for tritium migration from the nuclear to the chemical plant.

  2. High flux reactor PIK to be at PNPI. Scientific program

    E-Print Network [OSTI]

    Titov, Anatoly

    High flux reactor PIK to be at PNPI. Scientific program V.V.Fedorov Petersburg Nuclear Physics Venus Pavilion fire-lookout tower #12;General view of the reactor PIK buildings #12;The project of PIK of 60-th, but till now it does not become out of date and now used for all modern reactors. In 1991

  3. High temperature superconductor current leads

    DOE Patents [OSTI]

    Hull, John R. (Hinsdale, IL); Poeppel, Roger B. (Glen Ellyn, IL)

    1995-01-01T23:59:59.000Z

    An electrical lead having one end for connection to an apparatus in a cryogenic environment and the other end for connection to an apparatus outside the cryogenic environment. The electrical lead includes a high temperature superconductor wire and an electrically conductive material distributed therein, where the conductive material is present at the one end of the lead at a concentration in the range of from 0 to about 3% by volume, and at the other end of the lead at a concentration of less than about 20% by volume. Various embodiments are shown for groups of high temperature superconductor wires and sheaths.

  4. High temperature superconductor current leads

    DOE Patents [OSTI]

    Hull, J.R.; Poeppel, R.B.

    1995-06-20T23:59:59.000Z

    An electrical lead is disclosed having one end for connection to an apparatus in a cryogenic environment and the other end for connection to an apparatus outside the cryogenic environment. The electrical lead includes a high temperature superconductor wire and an electrically conductive material distributed therein, where the conductive material is present at the one end of the lead at a concentration in the range of from 0 to about 3% by volume, and at the other end of the lead at a concentration of less than about 20% by volume. Various embodiments are shown for groups of high temperature superconductor wires and sheaths. 9 figs.

  5. Thermocouples For High Temperature In-Pile Testing

    SciTech Connect (OSTI)

    J. L. Rempe

    2005-11-01T23:59:59.000Z

    Many advanced nuclear reactor designs require new fuel, cladding and structural materials. Data are needed to characeterize the performance of these new materials in high temperature, oxidizing and radiation conditions. To obtain this data, robust instrumentation is needed htat can survive proposed test conditions. Traditional methods for measuring temperature in-pile degrade at temperatures above 1080 degrees C. Hence, a project was intiated to develop specialized thermocouples for high temperature in-pile applications (see Rempe and Wilkins, 2005). This paper summarizes efforts to develop, fabricate and evaluate these specialized thermocouples.

  6. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    SciTech Connect (OSTI)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31T23:59:59.000Z

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

  7. HIGH TEMPERATURE GEOTHERMAL RESERVOIR ENGINEERING

    E-Print Network [OSTI]

    Schroeder, R.C.

    2009-01-01T23:59:59.000Z

    on the Cerro P r i e t o Geothermal F i e l d , Mexicali,e C e r r o P r i e t o Geothermal F i e l d , Baja C a l i1979 HIGH TEMPERATURE GEOTHERMAL RESERVOIR ENGINEERING R.

  8. Geothermal high temperature instrumentation applications

    SciTech Connect (OSTI)

    Normann, R.A. [Sandia National Labs., Albuquerque, NM (United States); Livesay, B.J. [Livesay Consultants (United States)

    1998-06-11T23:59:59.000Z

    A quick look at the geothermal industry shows a small industry producing about $1 billion in electric sales annually. The industry is becoming older and in need of new innovative solutions to instrumentation problems. A quick look at problem areas is given along with basic instrumentation requirements. The focus of instrumentation is on high temperature electronics.

  9. Fusion Engineering and Design 54 (2001) 167180 Options for the use of high temperature superconductor

    E-Print Network [OSTI]

    California at San Diego, University of

    superconductor in tokamak fusion reactor designs L. Bromberg a, *, M. Tekula b , L.A. El-Guebaly c , R. Miller d temperature superconductors (HTS) in long term tokamak fusion reactors is analyzed in this paper. The well-documented physical properties of high temperature superconductors are used in the evaluation. Short-sam- ple wires

  10. Self isolating high frequency saturable reactor

    DOE Patents [OSTI]

    Moore, James A. (Powell, TN)

    1998-06-23T23:59:59.000Z

    The present invention discloses a saturable reactor and a method for decoupling the interwinding capacitance from the frequency limitations of the reactor so that the equivalent electrical circuit of the saturable reactor comprises a variable inductor. The saturable reactor comprises a plurality of physically symmetrical magnetic cores with closed loop magnetic paths and a novel method of wiring a control winding and a RF winding. The present invention additionally discloses a matching network and method for matching the impedances of a RF generator to a load. The matching network comprises a matching transformer and a saturable reactor.

  11. Acid Doped Membranes for High Temperature PEMFC

    Broader source: Energy.gov [DOE]

    Presentation on Acid Doped Membranes for High Temperature PEMFC to the High Temperature Membrane Working Group, May 25, 2004 in Philadelphia, PA.

  12. High-Temperature Thermoelectric Materials Characterization for...

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

    High-Temperature Thermoelectric Materials Characterization for Automotive Waste Heat Recovery: Success Stories from the High Temperature Materials Laboratory (HTML) User Program...

  13. High Temperature Thermoelectric Materials Characterization for...

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

    High Temperature Thermoelectric Materials Characterization for Automotive Waste Heat Recovery: Success Stories from the High Temperature Materials Laboratory (HTML) User Program...

  14. Nanostructured High Temperature Bulk Thermoelectric Energy Conversion...

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

    High Temperature Bulk Thermoelectric Energy Conversion for Efficient Waste Heat Recovery Nanostructured High Temperature Bulk Thermoelectric Energy Conversion for Efficient Waste...

  15. Manufacturing Barriers to High Temperature PEM Commercialization...

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

    Barriers to High Temperature PEM Commercialization Manufacturing Barriers to High Temperature PEM Commercialization Presented at the NREL Hydrogen and Fuel Cell Manufacturing R&D...

  16. Deposition method for producing silicon carbide high-temperature semiconductors

    DOE Patents [OSTI]

    Hsu, George C. (La Crescenta, CA); Rohatgi, Naresh K. (W. Corine, CA)

    1987-01-01T23:59:59.000Z

    An improved deposition method for producing silicon carbide high-temperature semiconductor material comprising placing a semiconductor substrate composed of silicon carbide in a fluidized bed silicon carbide deposition reactor, fluidizing the bed particles by hydrogen gas in a mildly bubbling mode through a gas distributor and heating the substrate at temperatures around 1200.degree.-1500.degree. C. thereby depositing a layer of silicon carbide on the semiconductor substrate.

  17. Progress in Understanding Low-Temperature Organic Compound Oxidation Using a Jet-Stirred Reactor

    E-Print Network [OSTI]

    1 Progress in Understanding Low-Temperature Organic Compound Oxidation Using a Jet-Stirred Reactor Lorraine, CNRS, ENSIC, BP 20451, 1 rue Grandville, 54000 Nancy, France Abstract The jet-stirred reactor compounds: rapid compression machines, shock tubes, and heated continuous flow reactors, such as flow tubes

  18. High temperature turbine engine structure

    DOE Patents [OSTI]

    Carruthers, William D. (Mesa, AZ); Boyd, Gary L. (Tempe, AZ)

    1994-01-01T23:59:59.000Z

    A high temperature ceramic/metallic turbine engine includes a metallic housing which journals a rotor member of the turbine engine. A ceramic disk-like shroud portion of the engine is supported on the metallic housing portion and maintains a close running clearance with the rotor member. A ceramic spacer assembly maintains the close running clearance of the shroud portion and rotor member despite differential thermal movements between the shroud portion and metallic housing portion.

  19. High temperature turbine engine structure

    DOE Patents [OSTI]

    Carruthers, William D. (Mesa, AZ); Boyd, Gary L. (Tempe, AZ)

    1993-01-01T23:59:59.000Z

    A high temperature ceramic/metallic turbine engine includes a metallic housing which journals a rotor member of the turbine engine. A ceramic disk-like shroud portion of the engine is supported on the metallic housing portion and maintains a close running clearance with the rotor member. A ceramic spacer assembly maintains the close running clearance of the shroud portion and rotor member despite differential thermal movements between the shroud portion and metallic housing portion.

  20. High temperature turbine engine structure

    DOE Patents [OSTI]

    Carruthers, William D. (Mesa, AZ); Boyd, Gary L. (Tempe, AZ)

    1992-01-01T23:59:59.000Z

    A high temperature ceramic/metallic turbine engine includes a metallic housing which journals a rotor member of the turbine engine. A ceramic disk-like shroud portion of the engine is supported on the metallic housing portion and maintains a close running clearance with the rotor member. A ceramic spacer assembly maintains the close running clearance of the shroud portion and rotor member despite differential thermal movements between the shroud portion and metallic housing portion.

  1. Aging study of boiling water reactor high pressure injection systems

    SciTech Connect (OSTI)

    Conley, D.A.; Edson, J.L.; Fineman, C.F. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1995-03-01T23:59:59.000Z

    The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200{degrees}C (2,200{degrees}F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission`s Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed.

  2. Liquid Fuel Production from Biomass via High Temperature Steam Electrolysis

    SciTech Connect (OSTI)

    Grant L. Hawkes; Michael G. McKellar

    2009-11-01T23:59:59.000Z

    A process model of syngas production using high temperature electrolysis and biomass gasification is presented. Process heat from the biomass gasifier is used to heat steam for the hydrogen production via the high temperature steam electrolysis process. Hydrogen from electrolysis allows a high utilization of the biomass carbon for syngas production. Oxygen produced form the electrolysis process is used to control the oxidation rate in the oxygen-fed biomass gasifier. Based on the gasifier temperature, 94% to 95% of the carbon in the biomass becomes carbon monoxide in the syngas (carbon monoxide and hydrogen). Assuming the thermal efficiency of the power cycle for electricity generation is 50%, (as expected from GEN IV nuclear reactors), the syngas production efficiency ranges from 70% to 73% as the gasifier temperature decreases from 1900 K to 1500 K. Parametric studies of system pressure, biomass moisture content and low temperature alkaline electrolysis are also presented.

  3. High temperature ceramic composition for hydrogen retention

    DOE Patents [OSTI]

    Webb, R.W.

    1974-01-01T23:59:59.000Z

    A ceramic coating for H retention in fuel elements is described. The coating has relatively low thermal neutron cross section, is not readily reduced by H at 1500 deg F, is adherent to the fuel element base metal, and is stable at reactor operating temperatures. (JRD)

  4. Neutron and X-ray experiments at high temperature P. Aldebert (*)

    E-Print Network [OSTI]

    Boyer, Edmond

    neutron scattering have appeared as power- ful tools to get information, mainly structural temperature scattering devices compared to X-rays. At the present time thermal neutron high flux reactors be investigated by neutron scattering.

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

    SciTech Connect (OSTI)

    Michael G. McKellar; Edwin A. Harvego

    2010-05-01T23:59:59.000Z

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

  6. Thermal disconnect for high-temperature batteries

    DOE Patents [OSTI]

    Jungst, Rudolph George (Albuquerque, NM); Armijo, James Rudolph (Albuquerque, NM); Frear, Darrel Richard (Austin, TX)

    2000-01-01T23:59:59.000Z

    A new type of high temperature thermal disconnect has been developed to protect electrical and mechanical equipment from damage caused by operation at extreme temperatures. These thermal disconnects allow continuous operation at temperatures ranging from 250.degree. C. to 450.degree. C., while rapidly terminating operation at temperatures 50.degree. C. to 150.degree. C. higher than the continuous operating temperature.

  7. Faraday imaging at high temperatures

    DOE Patents [OSTI]

    Hackel, L.A.; Reichert, P.

    1997-03-18T23:59:59.000Z

    A Faraday filter rejects background light from self-luminous thermal objects, but transmits laser light at the passband wavelength, thus providing an ultra-narrow optical bandpass filter. The filter preserves images so a camera looking through a Faraday filter at a hot target illuminated by a laser will not see the thermal radiation but will see the laser radiation. Faraday filters are useful for monitoring or inspecting the uranium separator chamber in an atomic vapor laser isotope separation process. Other uses include viewing welds, furnaces, plasma jets, combustion chambers, and other high temperature objects. These filters are can be produced at many discrete wavelengths. A Faraday filter consists of a pair of crossed polarizers on either side of a heated vapor cell mounted inside a solenoid. 3 figs.

  8. Faraday imaging at high temperatures

    DOE Patents [OSTI]

    Hackel, Lloyd A. (Livermore, CA); Reichert, Patrick (Hayward, CA)

    1997-01-01T23:59:59.000Z

    A Faraday filter rejects background light from self-luminous thermal objects, but transmits laser light at the passband wavelength, thus providing an ultra-narrow optical bandpass filter. The filter preserves images so a camera looking through a Faraday filter at a hot target illuminated by a laser will not see the thermal radiation but will see the laser radiation. Faraday filters are useful for monitoring or inspecting the uranium separator chamber in an atomic vapor laser isotope separation process. Other uses include viewing welds, furnaces, plasma jets, combustion chambers, and other high temperature objects. These filters are can be produced at many discrete wavelengths. A Faraday filter consists of a pair of crossed polarizers on either side of a heated vapor cell mounted inside a solenoid.

  9. High-temperature thermocouples and related methods

    DOE Patents [OSTI]

    Rempe, Joy L. (Idaho Falls, ID); Knudson, Darrell L. (Firth, ID); Condie, Keith G. (Idaho Falls, ID); Wilkins, S. Curt (Idaho Falls, ID)

    2011-01-18T23:59:59.000Z

    A high-temperature thermocouple and methods for fabricating a thermocouple capable of long-term operation in high-temperature, hostile environments without significant signal degradation or shortened thermocouple lifetime due to heat induced brittleness.

  10. Stability analysis of supercritical water cooled reactors

    E-Print Network [OSTI]

    Zhao, Jiyun, Ph. D. Massachusetts Institute of Technology

    2005-01-01T23:59:59.000Z

    The Supercritical Water-Cooled Reactor (SCWR) is a concept for an advanced reactor that will operate at high pressure (25MPa) and high temperature (500C average core exit). The high coolant temperature as it leaves the ...

  11. Materials Characterization Capabilities at the High Temperature...

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

    Lightweighting Materials Materials Characterization Capabilities at the High Temperature Materials Laboratory: Focus Lightweighting Materials 2011 DOE Hydrogen and Fuel Cells...

  12. Agenda: High Temperature Membrane Working Group Meeting

    Broader source: Energy.gov [DOE]

    Agenda for the High Temperature Membrane Working Group (HTMWG) meeting on May 18, 2009, in Arlington, Virginia

  13. Innovative fuel designs for high power density pressurized water reactor

    E-Print Network [OSTI]

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

    2006-01-01T23:59:59.000Z

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

  14. TEMPERATURE MONITORING OPTIONS AVAILABLE AT THE IDAHO NATIONAL LABORATORY ADVANCED TEST REACTOR

    SciTech Connect (OSTI)

    J.E. Daw; J.L. Rempe; D.L. Knudson; T. Unruh; B.M. Chase; K.L Davis

    2012-03-01T23:59:59.000Z

    As part of the Advanced Test Reactor National Scientific User Facility (ATR NSUF) program, the Idaho National Laboratory (INL) has developed in-house capabilities to fabricate, test, and qualify new and enhanced sensors for irradiation testing. To meet recent customer requests, an array of temperature monitoring options is now available to ATR users. The method selected is determined by test requirements and budget. Melt wires are the simplest and least expensive option for monitoring temperature. INL has recently verified the melting temperature of a collection of materials with melt temperatures ranging from 100 to 1000 C with a differential scanning calorimeter installed at INLs High Temperature Test Laboratory (HTTL). INL encapsulates these melt wires in quartz or metal tubes. In the case of quartz tubes, multiple wires can be encapsulated in a single 1.6 mm diameter tube. The second option available to ATR users is a silicon carbide temperature monitor. The benefit of this option is that a single small monitor (typically 1 mm x 1 mm x 10 mm or 1 mm diameter x 10 mm length) can be used to detect peak irradiation temperatures ranging from 200 to 800 C. Equipment has been installed at INLs HTTL to complete post-irradiation resistivity measurements on SiC monitors, a technique that has been found to yield the most accurate temperatures from these monitors. For instrumented tests, thermocouples may be used. In addition to Type-K and Type-N thermocouples, a High Temperature Irradiation Resistant ThermoCouple (HTIR-TC) was developed at the HTTL that contains commercially-available doped molybdenum paired with a niobium alloy thermoelements. Long duration high temperature tests, in furnaces and in the ATR and other MTRs, demonstrate that the HTIR-TC is accurate up to 1800 C and insensitive to thermal neutron interactions. Thus, degradation observed at temperatures above 1100 C with Type K and N thermocouples and decalibration due to transmutation with tungsten-rhenium and platinum rhodium thermocouples can be avoided. INL is also developing an Ultrasonic Thermometry (UT) capability. In addition to small size, UTs offer several potential advantages over other temperature sensors. Measurements may be made near the melting point of the sensor material, potentially allowing monitoring of temperatures up to 3000 C. In addition, because no electrical insulation is required, shunting effects are avoided. Most attractive, however, is the ability to introduce acoustic discontinuities to the sensor, as this enables temperature measurements at several points along the sensor length. As discussed in this paper, the suite of temperature monitors offered by INL is not only available to ATR users, but also to users at other MTRs.

  15. Decommissioning of the high flux beam reactor at Brookhaven Lab

    SciTech Connect (OSTI)

    Hu, J.P. [National Synchrotron Light Source, Brookhaven Laboratory, Upton, NY 11973 (United States); Reciniello, R.N. [Radiological Control Div., Brookhaven Laboratory, Upton, NY 11973 (United States); Holden, N.E. [National Nuclear Data Center, Brookhaven Laboratory, Upton, NY 11973 (United States)

    2011-07-01T23:59:59.000Z

    The high-flux beam reactor (HFBR) at the Brookhaven National Laboratory was a heavy water cooled and moderated reactor that achieved criticality on Oct. 31, 1965. It operated at a power level of 40 megawatts. An equipment upgrade in 1982 allowed operations at 60 megawatts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 megawatts. The HFBR was shut down in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of groundwater from wells located adjacent to the reactor's spent fuel pool. The reactor remained shut down for almost three years for safety and environmental reviews. In November 1999 the United States Dept. of Energy decided to permanently shut down the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome, which still contains the irradiated reactor vessel, is presently under 24/7 surveillance for safety. Detailed dosimetry performed for the HFBR decommissioning during 1996-2009 is described in the paper. (authors)

  16. High Temperature Superconducting Underground Cable

    SciTech Connect (OSTI)

    Farrell, Roger, A.

    2010-02-28T23:59:59.000Z

    The purpose of this Project was to design, build, install and demonstrate the technical feasibility of an underground high temperature superconducting (HTS) power cable installed between two utility substations. In the first phase two HTS cables, 320 m and 30 m in length, were constructed using 1st generation BSCCO wire. The two 34.5 kV, 800 Arms, 48 MVA sections were connected together using a superconducting joint in an underground vault. In the second phase the 30 m BSCCO cable was replaced by one constructed with 2nd generation YBCO wire. 2nd generation wire is needed for commercialization because of inherent cost and performance benefits. Primary objectives of the Project were to build and operate an HTS cable system which demonstrates significant progress towards commercial progress and addresses real world utility concerns such as installation, maintenance, reliability and compatibility with the existing grid. Four key technical areas addressed were the HTS cable and terminations (where the cable connects to the grid), cryogenic refrigeration system, underground cable-to-cable joint (needed for replacement of cable sections) and cost-effective 2nd generation HTS wire. This was the worlds first installation and operation of an HTS cable underground, between two utility substations as well as the first to demonstrate a cable-to-cable joint, remote monitoring system and 2nd generation HTS.

  17. Computational Analysis of Fluid Flow in Pebble Bed Modular Reactor

    E-Print Network [OSTI]

    Gandhir, Akshay

    2012-10-19T23:59:59.000Z

    High Temperature Gas-cooled Reactor (HTGR) is a Generation IV reactor under consideration by Department of Energy and in the nuclear industry. There are two categories of HTGRs, namely, Pebble Bed Modular Reactor (PBMR) and Prismatic reactor. Pebble...

  18. Temperature dependent scattering cross section effects on nuclear reactor control

    E-Print Network [OSTI]

    Biggs, Charles Leon

    1968-01-01T23:59:59.000Z

    Reactor e e o e o a a e e ~ a o e ~ a e o o 43 Ato. . . Dsnsitiec of 'Materials in the Conceived Fast Nuclear Reactor ~ ~ o e o e e a o a o e e o ~ ~ ~ ~ 6, IDPut SPecifications oi' the AIYi-6 Criticality arch e o o e a a ~ a e e o ~ ~ ~ e e e ~ o e... both reactors depended upon axially expanding fuel elements for inherent control, other methods should be considered, Due to ths magnitude and sign of the reactivity cosfi'ioients, inherent control is especially of. interest in large fast nuclear...

  19. Method of and apparatus for removing silicon from a high temperature sodium coolant

    DOE Patents [OSTI]

    Yunker, W.H.; Christiansen, D.W.

    1983-11-25T23:59:59.000Z

    This patent discloses a method of and system for removing silicon from a high temperature liquid sodium coolant system for a nuclear reactor. The sodium is cooled to a temperature below the silicon saturation temperature and retained at such reduced temperature while inducing high turbulence into the sodium flow for promoting precipitation of silicon compounds and ultimate separation of silicon compound particles from the liquid sodium.

  20. High temperature superconducting fault current limiter

    DOE Patents [OSTI]

    Hull, J.R.

    1997-02-04T23:59:59.000Z

    A fault current limiter for an electrical circuit is disclosed. The fault current limiter includes a high temperature superconductor in the electrical circuit. The high temperature superconductor is cooled below its critical temperature to maintain the superconducting electrical properties during operation as the fault current limiter. 15 figs.

  1. High temperature superconducting fault current limiter

    DOE Patents [OSTI]

    Hull, John R. (Hinsdale, IL)

    1997-01-01T23:59:59.000Z

    A fault current limiter (10) for an electrical circuit (14). The fault current limiter (10) includes a high temperature superconductor (12) in the electrical circuit (14). The high temperature superconductor (12) is cooled below its critical temperature to maintain the superconducting electrical properties during operation as the fault current limiter (10).

  2. Measurement of thermodynamic temperature of high temperature fixed points

    SciTech Connect (OSTI)

    Gavrilov, V. R.; Khlevnoy, B. B.; Otryaskin, D. A.; Grigorieva, I. A.; Samoylov, M. L.; Sapritsky, V. I. [All-Russian Research Institute for Optical and Physical Measurements (VNIIOFI), 46 Ozernaya St., Moscow 119361 (Russian Federation)] [All-Russian Research Institute for Optical and Physical Measurements (VNIIOFI), 46 Ozernaya St., Moscow 119361 (Russian Federation)

    2013-09-11T23:59:59.000Z

    The paper is devoted to VNIIOFI's measurements of thermodynamic temperature of the high temperature fixed points Co-C, Pt-C and Re-C within the scope of the international project coordinated by the Consultative Committee for Thermometry working group 5 'Radiation Thermometry'. The melting temperatures of the fixed points were measured by a radiance mode radiation thermometer calibrated against a filter radiometer with known irradiance spectral responsivity via a high temperature black body. This paper describes the facility used for the measurements, the results and estimated uncertainties.

  3. Deep Trek High Temperature Electronics Project

    SciTech Connect (OSTI)

    Bruce Ohme

    2007-07-31T23:59:59.000Z

    This report summarizes technical progress achieved during the cooperative research agreement between Honeywell and U.S. Department of Energy to develop high-temperature electronics. Objects of this development included Silicon-on-Insulator (SOI) wafer process development for high temperature, supporting design tools and libraries, and high temperature integrated circuit component development including FPGA, EEPROM, high-resolution A-to-D converter, and a precision amplifier.

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

    E-Print Network [OSTI]

    Steen, Celeste Marie

    1990-01-01T23:59:59.000Z

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

  5. High thermal power density heat transfer apparatus providing electrical isolation at high temperature using heat pipes

    SciTech Connect (OSTI)

    Morris, J. F.

    1985-03-19T23:59:59.000Z

    This invention is directed to transferring heat from an extremely high temperature source to an electrically isolated lower temperature receiver. The invention is particularly concerned with supplying thermal power to a thermionic converter from a nuclear reactor with electric isolation. Heat from a high temperature heat pipe is transferred through a vacuum or a gap filled with electrically nonconducting gas to a cooler heat pipe. The heat pipe is used to cool the nuclear reactor while the heat pipe is connected thermally and electrically to a thermionic converter. If the receiver requires greater thermal power density, geometries are used with larger heat pipe areas for transmitting and receiving energy than the area for conducting the heat to the thermionic converter. In this way the heat pipe capability for increasing thermal power densities compensates for the comparatively low thermal power densities through the electrically nonconducting gap between the two heat pipes.

  6. Power conversion unit studies for the next generation nuclear plant coupled to a high-temperature steam electrolysis facility

    E-Print Network [OSTI]

    Barner, Robert Buckner

    2007-04-25T23:59:59.000Z

    -cooled Fast Reactor (GFR), Lead-cooled Fast Reactor (LFR), Molten Salt Reactor (MSR), Sodium-cooled Fast Reactor (SFR), Supercritical-water-cooled Reactor (SCWR) and the Very-high-temperature Reactor (VHTR). An international effort to develop these new... and the hydrogen production plant4,5. Davis et al. investigated the possibility of helium and molten salts in the IHTL2. The thermal efficiency of the power conversion unit is paramount to the success of this next generation technology. Current light water...

  7. Sputtering properties of copper-lithium alloys at reactor-level temperatures and surface erosion rates

    SciTech Connect (OSTI)

    Krauss, A.R.; Gruen, D.M.; Lam, N.Q.; DeWald, A.B.

    1984-01-01T23:59:59.000Z

    Previous experiments on copper-lithium alloys at temperatures up to 250/sup 0/C and with erosion rates of .01 to .1 monolayer per second have shown that in the electric and magnetic field environment of a magnetic-confinement fusion reactor, it is possible to maintain a lithium overlayer which will significantly reduce the copper erosion rate. We have extended these experiments to the reactor-relevant regime of 350 to 400/sup 0/C, with erosion rates approaching one monolayer per second. By comparison with the lower flux experiments, it is found that radiation damage effects start to dominate both the surface concentration and depth profile of the lithium. The subsurface region of enhanced lithium concentration is broadened, while the surface concentration is not depleted as rapidly per incident ion as in the low flux case. The time-dependent lithium depth profile is calculated using a computer code developed at Argonne which includes both Gibbsian segregation and radiation-induced effects. The experimental results are compared with these calculations. It is found that the sputtering behavior of the copper-lithium alloy is highly dependent on the mass and energy spectrum of the incident particles, the sample temperature, subsurface structure, and the partial sputtering yields of the alloy components.

  8. Measurement of cryogenic moderator temperature effects in a small heterogeneous thermal reactor

    SciTech Connect (OSTI)

    Hoovler, G.S.; Ball, R.M.; Lewis, R.H.

    1994-12-31T23:59:59.000Z

    Past papers have described a critical experiment (CX) built at Sandia National Laboratories to investigate the neutronic behavior of the particle-bed reactor (PBK). Among the experiments previously reported were tests to measure the reactivity effect of uniform temperature variations between 20 and 80{degree}C. This paper describes additional experiments designed to examine the effects of cryogenic moderator temperatures on core reactivity and neutron spectrum. The general importance of temperature effects to the design of the PBR have been previously discussed. A unique feature of the PBR is that the moderator may be at cryogenic temperatures during reactor startup. Because temperature effects in small, heterogeneous thermal reactors can be significant and because we found no integral measurements with cryogenic moderators in such systems, an experiment with a cryogenic moderator was designed and performed in the CX as an extension to the isothermal measurements previously reported.

  9. Materials Characterization Capabilities at the High Temperature...

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

    2010 -- Washington D.C. lm028laracurzio2010o.pdf More Documents & Publications Materials Characterization Capabilities at the High Temperature Materials Laboratory and HTML...

  10. Materials Characterization Capabilities at the High Temperature...

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

    Review and Peer Evaluation lm028laracurzio2011o.pdf More Documents & Publications Materials Characterization Capabilities at the High Temperature Materials Laboratory and HTML...

  11. Materials Characterization Capabilities at the High Temperature...

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

    and Peer Evaluation Meeting lm028laracurzio2012o.pdf More Documents & Publications Materials Characterization Capabilities at the High Temperature Materials Laboratory and HTML...

  12. Materials Characterization Capabilities at the High Temperature...

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

    May 18-22, 2009 -- Washington D.C. lm01laracurzio.pdf More Documents & Publications Materials Characterization Capabilities at the High Temperature Materials Laboratory and HTML...

  13. Materials Characterization Capabilities at the High Temperature...

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

    Laboratory: Focus on Carbon Fiber and Composites Materials Characterization Capabilities at the High Temperature Materials Laboratory: Focus on Carbon Fiber and Composites 2011 DOE...

  14. Materials Characterization Capabilities at the High Temperature...

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

    Characterization Capabilities at the High Temperature Materials Laboratory: Focus on Carbon Fiber and Composites Project ID: LM027 DOE 2011 Vehicle Technologies Annual Merit...

  15. Intertwined Orders in High Temperature Superconductors

    E-Print Network [OSTI]

    Ostoja-Starzewski, Martin

    Intertwined Orders in High Temperature Superconductors ! Eduardo Fradkin University of Illinois · Electronic liquid crystal phases have also been seen heavy fermions and iron superconductors 7 #12

  16. Polyelectrolyte Materials for High Temperature Fuel Cells

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

    High 3M (3M) Temperature Fuel Cells John B. Kerr Lawrence Berkeley National Laboratory (LBNL) Collaborators: Los Alamos National Laboratory (LANL). February 13, 2007 This...

  17. Quantitative Modeling of High Temperature Magnetization Dynamics

    SciTech Connect (OSTI)

    Zhang, Shufeng

    2009-03-01T23:59:59.000Z

    Final Technical Report Project title: Quantitative Modeling of High Temperature Magnetization Dynamics DOE/Office of Science Program Manager Contact: Dr. James Davenport

  18. Photonic crystals for high temperature applications

    E-Print Network [OSTI]

    Yeng, Yi Xiang

    2014-01-01T23:59:59.000Z

    This thesis focuses on the design, optimization, fabrication, and experimental realization of metallic photonic crystals (MPhCs) for high temperature applications, for instance thermophotovoltaic (TPV) energy conversion ...

  19. High Temperature Materials Interim Data Qualification Report FY 2011

    SciTech Connect (OSTI)

    Nancy Lybeck

    2011-08-01T23:59:59.000Z

    Projects for the very high temperature reactor (VHTR) Technology Development Office provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR program has established the Next Generation Nuclear Plant (NGNP) Data Management and Analysis System (NDMAS) to ensure that VHTR data are qualified for use, stored in a readily accessible electronic form, and analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the High Temperature Materials characterization data stream, the processing of these data within NDMAS, and reports the interim fiscal year (FY) 2011 qualification status of the data. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category assigned by the data generator. The High Temperature Materials data are being collected under the Nuclear Quality Assurance (NQA)-1 guidelines and will be qualified data. For NQA-1 qualified data, the qualification activities include: (1) capture testing to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing to confirm that the data are an accurate representation of the system or object being measured, and (3) documenting that the data were collected under an NQA-1 or equivalent Quality Assurance program. Currently, data from seven test series within the High Temperature Materials data stream have been entered into the NDMAS vault, including tensile tests, creep tests, and cyclic tests. Of the 5,603,682 records currently in the vault, 4,480,444 have been capture passed, and capture testing is in process for the remaining 1,123,238.

  20. High temperature synthetic cement retarder

    SciTech Connect (OSTI)

    Eoff, L.S.; Buster, D.

    1995-11-01T23:59:59.000Z

    A synthetic cement retarder which provides excellent retardation and compressive strength development has been synthesized. The response properties and temperature ranges of the synthetic retarder far exceed those of commonly used retarders such as lignosulfonates. The chemical nature of the new retarder is discussed and compared to another synthetic retarder.

  1. Investigations into High Temperature Components and Packaging

    SciTech Connect (OSTI)

    Marlino, L.D.; Seiber, L.E.; Scudiere, M.B.; M.S. Chinthavali, M.S.; McCluskey, F.P.

    2007-12-31T23:59:59.000Z

    The purpose of this report is to document the work that was performed at the Oak Ridge National Laboratory (ORNL) in support of the development of high temperature power electronics and components with monies remaining from the Semikron High Temperature Inverter Project managed by the National Energy Technology Laboratory (NETL). High temperature electronic components are needed to allow inverters to operate in more extreme operating conditions as required in advanced traction drive applications. The trend to try to eliminate secondary cooling loops and utilize the internal combustion (IC) cooling system, which operates with approximately 105 C water/ethylene glycol coolant at the output of the radiator, is necessary to further reduce vehicle costs and weight. The activity documented in this report includes development and testing of high temperature components, activities in support of high temperature testing, an assessment of several component packaging methods, and how elevated operating temperatures would impact their reliability. This report is organized with testing of new high temperature capacitors in Section 2 and testing of new 150 C junction temperature trench insulated gate bipolar transistor (IGBTs) in Section 3. Section 4 addresses some operational OPAL-GT information, which was necessary for developing module level tests. Section 5 summarizes calibration of equipment needed for the high temperature testing. Section 6 details some additional work that was funded on silicon carbide (SiC) device testing for high temperature use, and Section 7 is the complete text of a report funded from this effort summarizing packaging methods and their reliability issues for use in high temperature power electronics. Components were tested to evaluate the performance characteristics of the component at different operating temperatures. The temperature of the component is determined by the ambient temperature (i.e., temperature surrounding the device) plus the temperature increase inside the device due the internal heat that is generated due to conduction and switching losses. Capacitors and high current switches that are reliable and meet performance specifications over an increased temperature range are necessary to realize electronics needed for hybrid-electric vehicles (HEVs), fuel cell (FC) and plug-in HEVs (PHEVs). In addition to individual component level testing, it is necessary to evaluate and perform long term module level testing to ascertain the effects of high temperature operation on power electronics.

  2. Corrosion Resistant Coatings for High Temperature Applications

    SciTech Connect (OSTI)

    Besman, T.M.; Cooley, K.M.; Haynes, J.A.; Lee, W.Y.; Vaubert, V.M.

    1998-12-01T23:59:59.000Z

    Efforts to increase efficiency of energy conversion devices have required their operation at ever higher temperatures. This will force the substitution of higher-temperature structural ceramics for lower temperature materials, largely metals. Yet, many of these ceramics will require protection from high temperature corrosion caused by combustion gases, atmospheric contaminants, or the operating medium. This paper discusses examples of the initial development of such coatings and materials for potential application in combustion, aluminum smelting, and other harsh environments.

  3. Fundamental Thermal Fluid Physics of High Temperature Flows in Advanced Reactor Systems - Nuclear Energy Research Initiative Program Interoffice Work Order (IWO) MSF99-0254 Final Report for Period 1 August 1999 to 31 December 2002

    SciTech Connect (OSTI)

    McEligot, D.M.; Condie, K.G.; Foust, T.D.; McCreery, G.E.; Pink, R.J.; Stacey, D.E. (INEEL); Shenoy, A.; Baccaglini, G. (General Atomics); Pletcher, R.H. (Iowa State U.); Wallace, J.M.; Vukoslavcevic, P. (U. Maryland); Jackson, J.D. (U. Manchester, UK); Kunugi, T. (Kyoto U., Japan); Satake, S.-i. (Tokyo U. Science, Japan)

    2002-12-31T23:59:59.000Z

    The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of advanced reactors for higher efficiency and enhanced safety and for deployable reactors for electrical power generation, process heat utilization and hydrogen generation. While key applications would be advanced gas-cooled reactors (AGCRs) using the closed Brayton cycle (CBC) for higher efficiency (such as the proposed Gas Turbine - Modular Helium Reactor (GT-MHR) of General Atomics [Neylan and Simon, 1996]), results of the proposed research should also be valuable in reactor systems with supercritical flow or superheated vapors, e.g., steam. Higher efficiency leads to lower cost/kwh and reduces life-cycle impacts of radioactive waste (by reducing waters/kwh). The outcome will also be useful for some space power and propulsion concepts and for some fusion reactor concepts as side benefits, but they are not the thrusts of the investigation. The objective of the project is to provide fundamental thermal fluid physics knowledge and measurements necessary for the development of the improved methods for the applications.

  4. CRAD, Maintenance- Oak Ridge National Laboratory High Flux Isotope Reactor

    Broader source: Energy.gov [DOE]

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

  5. CRAD, Management- Oak Ridge National Laboratory High Flux Isotope Reactor

    Broader source: Energy.gov [DOE]

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

  6. CRAD, Training- Oak Ridge National Laboratory High Flux Isotope Reactor

    Broader source: Energy.gov [DOE]

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

  7. CRAD, Engineering- Oak Ridge National Laboratory High Flux Isotope Reactor

    Broader source: Energy.gov [DOE]

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

  8. Recrystallization of high temperature superconductors

    SciTech Connect (OSTI)

    Kouzoudis, D.

    1996-05-09T23:59:59.000Z

    Currently one of the most widely used high {Tc} superconductors is the Bi-based compounds Bi{sub 2}Sr{sub 2}CaCu{sub 2}O{sub z} and Bi{sub 2}Sr{sub 2}Ca{sub 2}Cu{sub 3}O{sub z} (known as BSCCO 2212 and 2223 compounds) with {Tc} values of about 85 K and 110 K respectively. Lengths of high performance conductors ranging from 100 to 1000 m long are routinely fabricated and some test magnets have been wound. An additional difficulty here is that although Bi-2212 and Bi-2223 phases exist over a wide range of stoichiometries, neither has been prepared in phase-pure form. So far the most successful method of constructing reliable and robust wires or tapes is the so called powder-in-tube (PIT) technique [1, 2, 3, 4, 5, 6, 7] in which oxide powder of the appropriate stoichiometry and phase content is placed inside a metal tube, deformed into the desired geometry (round wire or flat tape), and annealed to produce the desired superconducting properties. Intermediate anneals are often incorporated between successive deformation steps. Silver is the metal used in this process because it is the most compatible with the reacting phase. In all of the commercial processes for BSCCO, Ag seems to play a special catalytic role promoting the growth of high performance aligned grains that grow in the first few micrometers near the Ag/BSCCO interface. Adjacent to the Ag, the grain alignment is more perfect and the current density is higher than in the center of the tape. It is known that Ag lowers the melting point of several of the phases but the detailed mechanism for growth of these high performance grains is not clearly understood. The purpose of this work is to study the nucleation and growth of the high performance material at this interface.

  9. High-Temperature-High-Volume Lifting for Enhanced Geothermal...

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

    Turnquist GE Global Research High Temperature Tools and Sensors, Down-hole Pumps and Drilling May 19, 2010 This presentation does not contain any proprietary confidential, or...

  10. High-Temperature-High-Volume Lifting For Enhanced Geothermal...

    Open Energy Info (EERE)

    include high-temperature drive system materials, journal and thrust bearings, and corrosion and erosion-resistant lifting pump components. Finally, in Phase 3, the overall...

  11. High temperature solar selective coatings

    DOE Patents [OSTI]

    Kennedy, Cheryl E

    2014-11-25T23:59:59.000Z

    Improved solar collectors (40) comprising glass tubing (42) attached to bellows (44) by airtight seals (56) enclose solar absorber tubes (50) inside an annular evacuated space (54. The exterior surfaces of the solar absorber tubes (50) are coated with improved solar selective coatings {48} which provide higher absorbance, lower emittance and resistance to atmospheric oxidation at elevated temperatures. The coatings are multilayered structures comprising solar absorbent layers (26) applied to the meta surface of the absorber tubes (50), typically stainless steel, topped with antireflective Savers (28) comprising at least two layers 30, 32) of refractory metal or metalloid oxides (such as titania and silica) with substantially differing indices of refraction in adjacent layers. Optionally, at least one layer of a noble metal such as platinum can be included between some of the layers. The absorbent layers cars include cermet materials comprising particles of metal compounds is a matrix, which can contain oxides of refractory metals or metalloids such as silicon. Reflective layers within the coating layers can comprise refractory metal silicides and related compounds characterized by the formulas TiSi. Ti.sub.3SiC.sub.2, TiAlSi, TiAN and similar compounds for Zr and Hf. The titania can be characterized by the formulas TiO.sub.2, Ti.sub.3O.sub.5. TiOx or TiO.sub.xN.sub.1-x with x 0 to 1. The silica can be at least one of SiO.sub.2, SiO.sub.2x or SiO.sub.2xN.sub.1-x with x=0 to 1.

  12. High temperature hot water systems: A primer

    SciTech Connect (OSTI)

    Govan, F.A. [NMD and Associates, Cincinnati, OH (United States)

    1998-01-01T23:59:59.000Z

    The fundamental principles of high temperature water (HTW) system technology and its advantages for thermal energy distribution are presented. Misconceptions of this technology are also addressed. The paper describes design principles, applications, HTW properties, HTW system advantages, selecting the engineer, load diversification, design temperatures, system pressurization, pump considerations, constant vs. VS pumps, HTW generator types, and burners and controls.

  13. High-Temperature Water Splitting | Department of Energy

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

    Temperature Water Splitting High-Temperature Water Splitting High-temperature water splitting (a "thermochemical" process) is a long-term technology in the early stages of...

  14. S and 4 Reactor: Operating Lifetime and Estimates of Temperature and Burnup Reactivity Coefficients

    SciTech Connect (OSTI)

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

    2006-01-20T23:59:59.000Z

    The S and 4 reactor has a sectored, Mo-14%Re solid core for avoidance of single point failures in reactor cooling and Closed Brayton Cycle (CBC) energy conversion. The reactor is loaded with UN fuel, cooled with a He-Xe gas mixture at {approx}1200 K and operates at steady thermal power of 550 kW. Following a launch abort accident, the axial and radial BeO reflectors easily disassemble upon impact so that the bare reactor is subcriticial when submerged in wet sand or seawater and the core voids are filled with seawater. Spectral Shift Absorber (SSA) additives have been shown to increase the UN fuel enrichment and significantly reduce the total mass of the reactor. This paper investigates the effects of SSA additions on the temperature and burnup reactivity coefficients and the operational lifetime of the S and 4 reactor. SSAs slightly decrease the temperature reactivity feedback coefficient, but significantly increase the operating lifetime by decreasing the burnup reactivity coefficient. With no SSAs, fuel enrichment is only 58.5 wt% and the estimated operating lifetime is the shortest (7.6 years) with the highest temperature and burnup reactivity feedback coefficients (-0.2709 cent /K and -1.3470 $/atom%). With europium-151 and gadolinium-155 additions, the enrichment (91.5 and 94 wt%) and operating lifetime (9.9 and 9.8 years) of the S and 4 reactor are the highest while the temperature and burnup reactivity coefficients (-0.2382 and -0.2447 cent /K; -0.9073 and 0.8502 $/atom%) are the lowest.

  15. SUSY and symmetry nonrestoration at high temperature

    SciTech Connect (OSTI)

    Bajc, Borut [J. Stefan Institute, 1001 Ljubljana (Slovenia)

    1999-07-15T23:59:59.000Z

    The status of internal symmetry breaking at high temperature in super-symmetric models is shortly reviewed. This possibility could solve some well known cosmological problems, such as the domain wall, monopole and false vacuum problems.

  16. Fabrication and Design Aspects of High-Temperature Compact Diffusion Bonded Heat Exchangers

    SciTech Connect (OSTI)

    Mylavarapu, Sai K. [Ohio State University; Sun, Xiaodong [Ohio State University; Christensen, Richard N. [Ohio State University; Glosup, Richard E. [Ohio State University; Unocic, Raymond R [ORNL

    2012-01-01T23:59:59.000Z

    The very high temperature reactor (VHTR), using gas-cooled reactor technology, is one of the six reactor concepts selected by the Generation IV International Forum and is anticipated to be the reactor type for the next generation nuclear plant (NGNP). In this type of reactor with an indirect power cycle system, a high-temperature and high integrity intermediate heat exchanger (IHX) with high effectiveness is required to efficiently transfer the core thermal output to secondary fluid for electricity production, process heat, or hydrogen cogeneration. The current Technology Readiness Level status issued by NGNP to all components associated with the IHX for reactor core outlet temperatures of 750-800oC is 3 on a scale of 1 to 10 with 10 being the most ready. At present, there is no proven high-temperature IHX concept for VHTRs. Amongst the various potential IHX concepts available, diffusion bonded heat exchangers (henceforth called printed circuit heat exchangers, or PCHEs) appear promising for NGNP applications. The design and fabrication of this key component of NGNP is the primary focus of this paper. In the current study, two PCHEs were fabricated using Alloy 617 plates and will be experimentally investigated for their thermal-hydraulic performance in a high-temperature helium test facility (HTHF). The HTHF was primarily designed and constructed to test the thermal-hydraulic performance of PCHEs The test facility is primarily of Alloy 800H construction and is designed to facilitate experiments at temperatures and pressures up to 800oC and 3 MPa, respectively. The PCHE fabrication related processes, i.e., photochemical machining and diffusion bonding are briefly discussed for Alloy 617 plates. Diffusion bonding of Alloy 617 plates with and without a Ni interlayer is discussed. Furthermore, preliminary microstructural and mechanical characterization studies of representative diffusion bonded Alloy 617 specimens are presented.

  17. The High Flux Beam Reactor at Brookhaven National Laboratory

    SciTech Connect (OSTI)

    Shapiro, S.M.

    1994-12-31T23:59:59.000Z

    Brookhaven National Laboratory`s High Flux Beam Reactor (HFBR) was built because of the need of the scientist to always want `more`. In the mid-50`s the Brookhaven Graphite reactor was churning away producing a number of new results when the current generation of scientists, led by Donald Hughes, realized the need for a high flux reactor and started down the political, scientific and engineering path that led to the BFBR. The effort was joined by a number of engineers and scientists among them, Chemick, Hastings, Kouts, and Hendrie, who came up with the novel design of the HFBR. The two innovative features that have been incorporated in nearly all other research reactors built since are: (i) an under moderated core arrangement which enables the thermal flux to peak outside the core region where beam tubes can be placed, and (ii) beam tubes that are tangential to the core which decrease the fast neutron background without affecting the thermal beam intensity. Construction began in the fall of 1961 and four years later, at a cost of $12 Million, criticality was achieved on Halloween Night, 1965. Thus began 30 years of scientific accomplishments.

  18. Design of high temperature high speed electromagnetic axial thrust bearing

    E-Print Network [OSTI]

    Mohiuddin, Mohammad Waqar

    2002-01-01T23:59:59.000Z

    DESIGN OF HIGH TEMPERATURE HIGH SPEED ELECTROMAGNETIC AXIAL THRUST BEARING A Thesis by MOHAMMAD WAQAR MOHIUDDIN Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment of the requirements for the degree... of MASTER OF SCIENCE December 2002 Major Subject: Mechanical Engineering DESIGN OF HIGH TEMPERATURE HIGH SPEED ELECTROMAGNETIC AXIAL THRUST BEARING A Thesis by MOHAMMAD WAQAR MOHIUDDIN Submitted to Texas A&M University in partial fulfillment...

  19. High temperature thermometric phosphors for use in a temperature sensor

    DOE Patents [OSTI]

    Allison, Stephen W. (Knoxville, TN); Cates, Michael R. (Oak Ridge, TN); Boatner, Lynn A. (Oak Ridge, TN); Gillies, George T. (Earlysville, VA)

    1998-01-01T23:59:59.000Z

    A high temperature phosphor consists essentially of a material having the general formula LuPO.sub.4 :Dy.sub.(x),Eu.sub.(y), wherein: 0.1 wt %.ltoreq.x.ltoreq.20 wt % and 0.1 wt %.ltoreq.y.ltoreq.20 wt %. The high temperature phosphor is in contact with an article whose temperature is to be determined. The article having the phosphor in contact with it is placed in the environment for which the temperature of the article is to be determined. The phosphor is excited by a laser causing the phosphor to fluoresce. The emission from the phosphor is optically focused into a beam-splitting mirror which separates the emission into two separate emissions, the emission caused by the dysprosium dopant and the emission caused by the europium dopent. The separated emissions are optically filtered and the intensities of the emission are detected and measured. The ratio of the intensity of each emission is determined and the temperature of the article is calculated from the ratio of the intensities of the separate emissions.

  20. High temperature thermometric phosphors for use in a temperature sensor

    DOE Patents [OSTI]

    Allison, S.W.; Cates, M.R.; Boatner, L.A.; Gillies, G.T.

    1998-03-24T23:59:59.000Z

    A high temperature phosphor consists essentially of a material having the general formula LuPO{sub 4}:Dy{sub (x)},Eu{sub (y)}, wherein: 0.1 wt %{<=}x{<=}20 wt % and 0.1 wt %{<=}y{<=}20 wt %. The high temperature phosphor is in contact with an article whose temperature is to be determined. The article having the phosphor in contact with it is placed in the environment for which the temperature of the article is to be determined. The phosphor is excited by a laser causing the phosphor to fluoresce. The emission from the phosphor is optically focused into a beam-splitting mirror which separates the emission into two separate emissions, the emission caused by the dysprosium dopant and the emission caused by the europium dopant. The separated emissions are optically filtered and the intensities of the emission are detected and measured. The ratio of the intensity of each emission is determined and the temperature of the article is calculated from the ratio of the intensities of the separate emissions. 2 figs.

  1. High temperature crystalline superconductors from crystallized glasses

    DOE Patents [OSTI]

    Shi, Donglu (Downers Grove, IL)

    1992-01-01T23:59:59.000Z

    A method of preparing a high temperature superconductor from an amorphous phase. The method involves preparing a starting material of a composition of Bi.sub.2 Sr.sub.2 Ca.sub.3 Cu.sub.4 Ox or Bi.sub.2 Sr.sub.2 Ca.sub.4 Cu.sub.5 Ox, forming an amorphous phase of the composition and heat treating the amorphous phase for particular time and temperature ranges to achieve a single phase high temperature superconductor.

  2. E-Print Network 3.0 - argonne high flux reactor Sample Search...

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

    for: argonne high flux reactor Page: << < 1 2 3 4 5 > >> 1 Thirteenth National School on Neutron and X-ray Scattering Summary: Neutron Source and High Flux Isotope Reactor...

  3. Apparatus and method for high temperature viscosity and temperature measurements

    DOE Patents [OSTI]

    Balasubramaniam, Krishnan (Mississippi State, MS); Shah, Vimal (Houston, TX); Costley, R. Daniel (Mississippi State, MS); Singh, Jagdish P. (Mississippi State, MS)

    2001-01-01T23:59:59.000Z

    A probe for measuring the viscosity and/or temperature of high temperature liquids, such as molten metals, glass and similar materials comprises a rod which is an acoustical waveguide through which a transducer emits an ultrasonic signal through one end of the probe, and which is reflected from (a) a notch or slit or an interface between two materials of the probe and (b) from the other end of the probe which is in contact with the hot liquid or hot melt, and is detected by the same transducer at the signal emission end. To avoid the harmful effects of introducing a thermally conductive heat sink into the melt, the probe is made of relatively thermally insulative (non-heat-conductive) refractory material. The time between signal emission and reflection, and the amplitude of reflections, are compared against calibration curves to obtain temperature and viscosity values.

  4. HIGH TEMPERATURE HIGH PRESSURE THERMODYNAMIC MEASUREMENTS FOR COAL MODEL COMPOUNDS

    SciTech Connect (OSTI)

    Vinayak N. Kabadi

    1999-02-20T23:59:59.000Z

    It is well known that the fluid phase equilibria can be represented by a number of {gamma}-models , but unfortunately most of them do not function well under high temperature. In this calculation, we mainly investigate the performance of UNIQUAC and NRTL models under high temperature, using temperature dependent parameters rather than using the original formulas. the other feature of this calculation is that we try to relate the excess Gibbs energy G{sup E}and enthalpy of mixing H{sup E}simultaneously. In other words, we will use the high temperature and pressure G{sup E} and H{sup E}data to regress the temperature dependant parameters to find out which model and what kind of temperature dependant parameters should be used.

  5. Measurement of Temperature Profile in the Reactor Cavity Cooling System

    E-Print Network [OSTI]

    Alhashimi, Tariq Yaqoob Sayed

    2014-12-02T23:59:59.000Z

    welded to the upper plenum which has two exhaust chimneys. Blowers are used to drive air through in-line heaters which are connected to the bottom end of the riser ducts. Experiments were conducted to measure the temperature spatial profile in the plenum...

  6. Design modification for the modular helium reactor for higher temperature operation and reliability studies for nuclear hydrogen production processes

    E-Print Network [OSTI]

    Reza, S.M. Mohsin

    2009-05-15T23:59:59.000Z

    Design options have been evaluated for the Modular Helium Reactor (MHR) for higher temperature operation. An alternative configuration for the MHR coolant inlet flow path is developed to reduce the peak vessel temperature (PVT). The coolant inlet...

  7. QED3 Theory of High Temperature Superconductors

    E-Print Network [OSTI]

    Tesanovic, Zlatko

    QED3 Theory of High Temperature Superconductors Zlatko Tesanovi´c The Johns Hopkins University-wave Superconductor to Antiferromagnet via Strange Metal #12;This talk is based on: M. Franz and ZT, Phys. Rev. Lett is The Problem in high Tc superconductors? · Superconducting state appears dx2-y2 "BCS-like". Low energy

  8. Low temperature plasma near a tokamak reactor limiter

    SciTech Connect (OSTI)

    Braams, B.J.; Singer, C.E.

    1985-01-01T23:59:59.000Z

    Analytic and two-dimensional computational solutions for the plasma parameters near a toroidally symmetric limiter are illustrated for the projected parameters of a Tokamak Fusion Core Experiment (TFCX). The temperature near the limiter plate is below 20 eV, except when the density 10 cm inside the limiter contact is 8 x 10/sup 13/cm/sup -3/ or less and the thermal diffusivity in the edge region is 2 x 10/sup 4/cm/sup 2//s or less. Extrapolation of recent experimental data suggests that neither of these conditions is likely to be met near ignition in TFCX, so a low plasma temperature near the limiter should be considered a likely possibility.

  9. Development of a 500 Watt High Temperature Thermoelectric Generator

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

    * TEG thermal and electrical interfaces modified to withstand high temperature environment 8 5 August, 2009 Deer 2009 9 100 Watt High Temperature TEG 100 Watt High...

  10. High Temperature Irradiation-Resistant Thermocouple Performance Improvements

    SciTech Connect (OSTI)

    Joshua Daw; Joy Rempe; Darrell Knudson; John Crepeau; S. Curtis Wilkins

    2009-04-01T23:59:59.000Z

    Traditional methods for measuring temperature in-pile degrade at temperatures above 1100 C. To address this instrumentation need, the Idaho National Laboratory (INL) developed and evaluated the performance of a high temperature irradiation-resistant thermocouple (HTIR-TC) that contains alloys of molybdenum and niobium. Data from high temperature (up to 1500 C) long duration (up to 4000 hours) tests and on-going irradiations at INLs Advanced Test Reactor demonstrate the superiority of these sensors to commercially-available thermocouples. However, several options have been identified that could further enhance their reliability, reduce their production costs, and allow their use in a wider range of operating conditions. This paper presents results from on-going Idaho National Laboratory (INL)/University of Idaho (UI) efforts to investigate options to improve HTIR-TC ductility, reliability, and resolution by investigating specially-formulated alloys of molybdenum and niobium and alternate diameter thermoelements (wires). In addition, on-going efforts to evaluate alternate fabrication approaches, such as drawn and loose assembly techniques will be discussed. Efforts to reduce HTIR-TC fabrication costs, such as the use of less expensive extension cable will also be presented. Finally, customized HTIR-TC designs developed for specific customer needs will be summarized to emphasize the varied conditions under which these sensors may be used.

  11. Silicon carbide oxidation in high temperature steam

    E-Print Network [OSTI]

    Arnold, Ramsey Paul

    2011-01-01T23:59:59.000Z

    The commercial nuclear power industry is continually looking for ways to improve reactor productivity and efficiency and to increase reactor safety. A concern that is closely regulated by the Nuclear Regulatory Commission ...

  12. Fusion reactor high vacuum pumping: Charcoal cryosorber tritium exposure results

    SciTech Connect (OSTI)

    Sedgley, D.W.; Walthers, C.R.; Jenkins, E.M. (Grumman Aerospace Corp., Bethpage, NY (United States))

    1991-01-01T23:59:59.000Z

    Recent experiments, have shown the practically of using activated charcoal (coconut charcoal) at 4{degrees}K to pump helium and hydrogen isotopes for a fusion reactor. Both speed and capacity for deuterium/helium and tritium/helium-3 mixtures were shown to be satisfactory. The long term effects of tritium on the charcoal/cement system developed by Grumman and LLNL were not known and a program was undertaken to see what, if any, effect long term tritium exposure has on the cryosorber. Several charcoal on aluminum test samples were subjected to six months exposure of tritium at approximately 77{degrees}K. The tritium was scanned several times with a residual gas analyzer and the speed-capacity performance of the samples was measured before, approximately half way through and after the exposure. Modest effects were noted which would not seriously restrict charcoal's use as a cryosorber for fusion reactor high vacuum pumping applications. 4 refs., 8 figs.

  13. THERMODYNAMIC CONSIDERATIONS FOR THERMAL WATER SPLITTING PROCESSES AND HIGH TEMPERATURE ELECTROLYSIS

    SciTech Connect (OSTI)

    J. E. O'Brien

    2008-11-01T23:59:59.000Z

    A general thermodynamic analysis of hydrogen production based on thermal water splitting processes is presented. Results of the analysis show that the overall efficiency of any thermal water splitting process operating between two temperature limits is proportional to the Carnot efficiency. Implications of thermodynamic efficiency limits and the impacts of loss mechanisms and operating conditions are discussed as they pertain specifically to hydrogen production based on high-temperature electrolysis. Overall system performance predictions are also presented for high-temperature electrolysis plants powered by three different advanced nuclear reactor types, over their respective operating temperature ranges.

  14. High temperature storage loop : final design report.

    SciTech Connect (OSTI)

    Gill, David Dennis; Kolb, William J.

    2013-07-01T23:59:59.000Z

    A three year plan for thermal energy storage (TES) research was created at Sandia National Laboratories in the spring of 2012. This plan included a strategic goal of providing test capability for Sandia and for the nation in which to evaluate high temperature storage (>650%C2%B0C) technology. The plan was to scope, design, and build a flow loop that would be compatible with a multitude of high temperature heat transfer/storage fluids. The High Temperature Storage Loop (HTSL) would be reconfigurable so that it was useful for not only storage testing, but also for high temperature receiver testing and high efficiency power cycle testing as well. In that way, HTSL was part of a much larger strategy for Sandia to provide a research and testing platform that would be integral for the evaluation of individual technologies funded under the SunShot program. DOE's SunShot program seeks to reduce the price of solar technologies to 6/kWhr to be cost competitive with carbon-based fuels. The HTSL project sought to provide evaluation capability for these SunShot supported technologies. This report includes the scoping, design, and budgetary costing aspects of this effort

  15. An examination of the feasibility of a very low temperature nuclear reactor

    E-Print Network [OSTI]

    Dupree, Stephen Allen

    1965-01-01T23:59:59.000Z

    Texas A8cM University 42 46 57 59 6l LIST OF TABLES Number Title Page Three-group Constants for the Proposed Low Temperature Reactor 21 Comparison of Temperature Variation of Factors in the Critical Equation at T = 10 K, Nu = 10 T, eT = 10...~ and the cr1tical radius and mass were found. The temperature coefficient of reactivity was then approx- imated from the critical equation. Three-group time independent diffusion theory is a modified Pl approximation to the Boitzmann transport equation. (8...

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

    SciTech Connect (OSTI)

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

    2013-09-11T23:59:59.000Z

    During the experiments simulating NPR (nuclear power reactor) accidents with a coolant loss fuel elements behavior in a steam-hydrogen medium was studied at the temperature changed with the rate from 1 to 100K/s within the range of 3001500 C. Indications of the thermocouples fixed on the cladding notably differ from real values of the cladding temperatures in the area of measuring junction due to thermal resistance influence of the transition zones cladding-junction and junction-coolant. The estimating method of a measurement error was considered which can provide adequate accounting of the influence factors. The method is based on thermal probing of a thermocouple by electric current flashing through thermoelements under the coolant presence or absence, a response time registration and processing, calculation of thermal inertia value for a thermocouple junction. A formula was derived for calculation of methodical error under stationary mode and within the stage of linear increase in temperature, which will determine the conditions for the cladding depressurization. Some variants of the formula application were considered, and the values of methodical errors were established which reached ?5% of maximum value by the final moment of the stage of linear increase in the temperature.

  17. Design and optimization of a high thermal flux research reactor via Kriging-based algorithm

    E-Print Network [OSTI]

    Kempf, Stephanie Anne

    2011-01-01T23:59:59.000Z

    In response to increasing demands for the services of research reactors, a 5 MW LEU-fueled research reactor core is developed and optimized to provide high thermal flux within specified limits upon thermal hydraulic ...

  18. Wear damage resulting from sliding impact kinematics in pressurized high temperature water: energetical and

    E-Print Network [OSTI]

    Paris-Sud XI, Universit de

    1 Wear damage resulting from sliding impact kinematics in pressurized high temperature water.bec@ec-lyon.fr Abstract Specific wear of Rod Cluster Control Assemblies (RCCA) in Pressurized Water nuclear Reactors (PWR) results from contacts with their guides due to flow induced vibration. Particular sliding impact contact

  19. HYDROGEN SULFIDE -HIGH TEMPERATURE DRILLING CONTINGENCY PLAN

    E-Print Network [OSTI]

    HYDROGEN SULFIDE - HIGH TEMPERATURE DRILLING CONTINGENCY PLAN OCEAN DRILLING PROGRAM TEXAS A&M UNIVERSITY Technical Note 16 Steven P. Howard Ocean Drilling Program Texas A&M University 1000 Discovery Drive College Station, TX 77845-9547 Daniel H. Reudelhuber Ocean Drilling Program Texas A&M University

  20. Advanced Converter Systems for High Temperature Environments

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

    500 1000 1500 2000 2500 Voltage (Volts) Current (nA) . 4.0 Resistance (mOhms) 3.5 3.0 2.5 2.0 1.5 0 20 40 60 80 100 Current (Amps) High temperature package voltage breakdown and...

  1. Fabrication of control rods for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Sease, J.D.

    1998-03-01T23:59:59.000Z

    The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A.

  2. Conversion feasibility studies for the Grenoble high flux reactor

    SciTech Connect (OSTI)

    Mo, S.C.; Matos, J.E.

    1989-01-01T23:59:59.000Z

    Feasibility studies for conversion of the High Flux Reactor (RHF) at Grenoble France have been performed at the Argonne National Laboratory in cooperation with the Institut Laue-Langevin (ILL). The uranium densities required for conversion of the RHF to reduced enrichment fuels were computed to be 7.9 g/cm{sup 3} with 20% enrichment, 4.8 g/cm{sup 3} with 29% enrichment, and 2.8 g/cm{sup 3} with 45% enrichment. Thermal flux reductions at the peak in the heavy water reflector were computed to be 3% with 45% enriched fuel and 7% with 20% enriched fuel. In each case, the reactor's 44 day cycle length was preserved and no changes were made in the fuel element geometry. If the cladding thickness could be reduced from 0.38 mm to 0.30 mm, the required uranium density with 20% enrichment would be about 6.0 g/cm{sup 3} and the thermal flux reduction at the peak in the heavy water reflector would be about 7%. Significantly higher uranium densities are required in the RHF than in heavy water reactors with more conventional designs because the neutron spectrum is much harder in the RHF. Reduced enrichment fuels with the uranium densities required for use in the RHF are either not available or are not licensable at the present time. 6 refs., 6 figs., 3 tabs.

  3. Alternate applications of fusion power: development of a high-temperature blanket for synthetic-fuel production

    SciTech Connect (OSTI)

    Howard, P.A.; Mattas, R.F.; Krajcinovic, D.; DePaz, J.; Gohar, Y.

    1981-11-01T23:59:59.000Z

    This study has shown that utilization of the unique features of a fusion reactor can result in a novel and potentially economical method of decomposing steam into hydrogen and oxygen. Most of the power of fusion reactors is in the form of energetic neutrons. If this power could be used to produce high temperature uncontaminated steam, a large fraction of the energy needed to decomposee the steam could be supplied as thermal energy by the fusion reaction. Proposed high temperature electrolysis processes require steam temperature in excess of 1000/sup 0/C for high efficiency. The design put forth in this study details a system that can accomplish that end.

  4. An investigation of temperature measurement methods in nuclear power plant reactor pressure vessel annealing

    SciTech Connect (OSTI)

    Acton, R.U.; Gill, W.; Sais, D.J.; Schulze, D.H.; Nakos, J.T. [Sandia National Labs., Albuquerque, NM (United States)

    1996-05-01T23:59:59.000Z

    The objective of this project was to provide an assessment of several methods by which the temperature of a commercial nuclear power plant reactor pressure vessel (RPV) could be measured during an annealing process. This project was a coordinated effort between DOE`s Office of Nuclear Energy, Science and Technology; DOE`s Light Water Reactor Technology Center at Sandia National Laboratories; and the Electric Power Research Institute`s Non- Destructive Evaluation Center. Ball- thermocouple probes similar to those described in NUREG/CR-5760, spring-loaded, metal- sheathed thermocouple probes, and 1778 air- suspended thermocouples were investigated in experiments that heated a section of an RPV wall to simulate a thermal annealing treatment. A parametric study of ball material, emissivity, thermal conductivity, and thermocouple function locations was conducted. Also investigated was a sheathed thermocouple failure mode known as shunting (electrical breakdown of insulation separating the thermocouple wires). Large errors were found between the temperature as measured by the probes and the true RPV wall temperature during heat-up and cool-down. At the annealing soak temperature, in this case 454{degrees}C [850`F], all sensors measured the same temperature within about {plus_minus}5% (23.6{degrees}C [42.5{degrees}F]). Because of these errors, actual RPV wall heating and cooling rates differed from those prescribed (by up to 29%). Shunting does not appear to be a problem under these conditions. The large temperature measurement errors led to the development of a thermal model that predicts the RPV wall temperature from the temperature of a ball- probe. Comparisons between the model and the experimental data for ball-probes indicate that the model could be a useful tool in predicting the actual RPV temperature based on the indicated ball- probe temperature. The model does not predict the temperature as well for the spring-loaded and air suspended probes.

  5. High temperature intermetallic binders for HVOF carbides

    SciTech Connect (OSTI)

    Shaw, K.G. [Xform, Inc., Cohoes, NY (United States); Gruninger, M.F.; Jarosinski, W.J. [Praxair Specialty Powders, Indianapolis, IN (United States)

    1994-12-31T23:59:59.000Z

    Gas turbines technology has a long history of employing the desirable high temperature physical attributes of ceramic-metallic (cermet) materials. The most commonly used coatings incorporate combinations of WC-Co and Cr{sub 3}C{sub 2}-NiCr, which have also been successfully utilized in other non-turbine coating applications. Increased turbine operating temperatures and other high temperature service conditions have made apparent the attractive notion of increasing the temperature capability and corrosion resistance of these coatings. In this study the intermetallic binder NiAl has been used to replace the cobalt and NiCr constituents of conventional WC and Cr{sub 3}C{sub 2} cermet powders. The composite carbide thermal spray powders were fabricated for use in the HVOF coating process. The structure of HVOF deposited NiAl-carbide coatings are compared directly to the more familiar WC-Co and Cr{sub 3}C{sub 2}-NiCr coatings using X-ray diffraction, back-scattered electron imaging (BEI) and electron dispersive spectroscopy (EDS). Hardness variations with temperature are reported and compared between the NiAl and Co/NiCr binders.

  6. High Temperature Fuel Cells in the European Union

    Broader source: Energy.gov [DOE]

    Presentation on High Temperature Fuel Cells in the European Union to the High Temperature Membrane Working Group, May 25, 2004 in Philadelphia, PA.

  7. Low Temperature Combustion Demonstrator for High Efficiency Clean...

    Energy Savers [EERE]

    Low Temperature Combustion Demonstrator for High Efficiency Clean Combustion Low Temperature Combustion Demonstrator for High Efficiency Clean Combustion Presentation from the U.S....

  8. Low Temperature Combustion Demonstrator for High Efficiency Clean...

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

    Demonstrator for High Efficiency Clean Combustion Low Temperature Combustion Demonstrator for High Efficiency Clean Combustion Applied low temperature combustion to the Navistar...

  9. Low-Temperature Combustion Demonstrator for High-Efficiency Clean...

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

    Low-Temperature Combustion Demonstrator for High-Efficiency Clean Combustion Low-Temperature Combustion Demonstrator for High-Efficiency Clean Combustion 2010 DOE Vehicle...

  10. Low Temperature Combustion Demonstrator for High Efficiency Clean...

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

    Low Temperature Combustion Demonstrator for High Efficiency Clean Combustion Low Temperature Combustion Demonstrator for High Efficiency Clean Combustion 2009 DOE Hydrogen Program...

  11. Cedarville Elementary & High School Space Heating Low Temperature...

    Open Energy Info (EERE)

    Elementary & High School Space Heating Low Temperature Geothermal Facility Jump to: navigation, search Name Cedarville Elementary & High School Space Heating Low Temperature...

  12. Possible Origin of Improved High Temperature Performance of Hydrotherm...

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

    Origin of Improved High Temperature Performance of Hydrothermally Aged CuBeta Zeolite Catalysts. Possible Origin of Improved High Temperature Performance of Hydrothermally Aged...

  13. Institute of Chemical Engineering and High Temperature Chemical...

    Open Energy Info (EERE)

    Institute of Chemical Engineering and High Temperature Chemical Processes ICEHT Jump to: navigation, search Name: Institute of Chemical Engineering and High Temperature Chemical...

  14. A High Temperature Direct Vehicle Exhaust Flowmeter for Heavy...

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

    A High Temperature Direct Vehicle Exhaust Flowmeter for Heavy Duty Diesel Emission Measurements. A High Temperature Direct Vehicle Exhaust Flowmeter for Heavy Duty Diesel Emission...

  15. High-Temperature, Air-Cooled Traction Drive Inverter Packaging...

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

    High-Temperature, Air-Cooled Traction Drive Inverter Packaging High-Temperature, Air-Cooled Traction Drive Inverter Packaging 2010 DOE Vehicle Technologies and Hydrogen Programs...

  16. Vehicle Technologies Office Merit Review 2014: High-Temperature...

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

    High-Temperature Air-Cooled Power Electronics Thermal Design Vehicle Technologies Office Merit Review 2014: High-Temperature Air-Cooled Power Electronics Thermal Design...

  17. High Resolution and Low-Temperature Photoelectron Spectroscopy...

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

    High Resolution and Low-Temperature Photoelectron Spectroscopy of an Oxygen-Linked Fullerene Dimer Dianion: C120O2-. High Resolution and Low-Temperature Photoelectron Spectroscopy...

  18. Development of a 100-Watt High Temperature Thermoelectric Generator

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

    performance TEG thermal and electrical interfaces modified to withstand high temperature environment Development of a 100 watt High Temperature TE Generator DEER 2008 11 Prototype...

  19. Nanostructured High-Temperature Bulk Thermoelectric Energy Conversion...

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

    High-Temperature Bulk Thermoelectric Energy Conversion for Efficient Automotive Waste Heat Recovery Nanostructured High-Temperature Bulk Thermoelectric Energy Conversion for...

  20. Low and high Temperature Dual Thermoelectric Generation Waste...

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

    and high Temperature Dual Thermoelectric Generation Waste Heat Recovery System for Light-Duty Vehicles Low and high Temperature Dual Thermoelectric Generation Waste Heat Recovery...

  1. Rotational Rehybridization and the High Temperature Phase of...

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

    Rotational Rehybridization and the High Temperature Phase of UC2. Rotational Rehybridization and the High Temperature Phase of UC2. Abstract: The screened hybrid approximation...

  2. High Temperature Thin Film Polymer Dielectric Based Capacitors...

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

    High Temperature Thin Film Polymer Dielectric Based Capacitors for HEV Power Electronic Systems High Temperature Thin Film Polymer Dielectric Based Capacitors for HEV Power...

  3. Development of Advanced High Temperature Fuel Cell Membranes

    Broader source: Energy.gov [DOE]

    Presentation on Development of Advanced High Temperature Fuel Cell Membranes to the High Temperature Membrane Working Group Meeting held in Arlington, Virginia, May 26,2005.

  4. Syngas Enhanced High Efficiency Low Temperature Combustion for...

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

    Enhanced High Efficiency Low Temperature Combustion for Clean Diesel Engines Syngas Enhanced High Efficiency Low Temperature Combustion for Clean Diesel Engines A significant...

  5. High Temperature Polymer Membrane Development at Argonne National Laboratory

    Broader source: Energy.gov [DOE]

    Summary of ANLs high temperature polymer membrane work presented to the High Temperature Membrane Working Group Meeting, Orlando FL, October 17, 2003

  6. High Temperature Thermal Array for Next Generation Solar Thermal...

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

    High Temperature Thermal Array for Next Generation Solar Thermal Power Production High Temperature Thermal Array for Next Generation Solar Thermal Power Production This...

  7. High-temperature directional drilling turbodrill

    SciTech Connect (OSTI)

    Neudecker, J.W.; Rowley, J.C.

    1982-02-01T23:59:59.000Z

    The development of a high-temperature turbodrill for directional drilling of geothermal wells in hard formations is summarized. The turbodrill may be used for straight-hole drilling but was especially designed for directional drilling. The turbodrill was tested on a dynamometer stand, evaluated in laboratory drilling into ambient temperature granite blocks, and used in the field to directionally drill a 12-1/4-in.-diam geothermal well in hot 200/sup 0/C (400/sup 0/F) granite at depths to 10,5000 ft.

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

    E-Print Network [OSTI]

    Steen, Celeste Marie

    1990-01-01T23:59:59.000Z

    in HTGR's and much is known about their performance under irradiation. The oxide fuels are best known for their ability to retain gaseous and heavy metal (U, Pu) fission products through the formation of stable oxides, while the carbide fuels are best...CALCULATED FUEL TEMPERATURES FOR A PROPOSED SPACE BASED REACTOR USING THE LUMPED PARAMETER METHOD A Thesis by CELESTE MARIE STEEN Submitted to the Office of Graduate Studies of Texas AgcM University in partial fulfillment of the requirements...

  9. Thermochemistry of high-temperature corrosion

    SciTech Connect (OSTI)

    Natesan, K.

    1980-01-01T23:59:59.000Z

    Multicomponent gas environments are prevalent in a number of energy systems, especially in those that utilize fossil fuels. The gas environments in these processes contain sulfur-bearing components in addition to oxidants. These complex environments, coupled with the elevated temperatures present in these systems, generally cause significant corrosion of engineering materials. Thermodynamic aspects of high-temperature corrosion processes occuring in complex gas mixtures are discussed, with emphasis on the role of thermochemical diagrams. The interrelationships between the corrosion behavior of materials and gas composition, alloy chemistry, and temperatures are examined. A number of examples from studies on materials behavior in coal-gasification environments are used to elucidate the role of thermochemistry in the understanding of corrosion processes that occur in complex gas mixtures. 11 figures.

  10. Pulsed plasma treatment of polluted gas using wet-/low-temperature corona reactors

    SciTech Connect (OSTI)

    Shimizu, Kazuo; Kinoshita, Katsuhiro; Yanagihara, Kenya; Rajanikanth, B.S.; Katsura, Shinji; Mizuno, Akira [Toyohashi Univ. of Technology, Aichi (Japan). Dept. of Ecological Engineering] [Toyohashi Univ. of Technology, Aichi (Japan). Dept. of Ecological Engineering

    1997-09-01T23:59:59.000Z

    Application of pulsed plasma for gas cleaning is gaining prominence in recent years, mainly from the energy consideration point of view. Normally, the gas treatment is carried out at or above room temperature by the conventional dry-type corona reactor. However, this treatment is still inadequate for the removal of certain stable gases present in the exhaust/flue gas mixture. The authors report here some interesting results of treatment of such stable gases like N{sub 2}O with pulsed plasma at subambient temperature. Also reported in this paper are improvements in DeNO/DeNO{sub x} efficiency using unconventional wet-type reactors, designed and fabricated by us, and operating at different subambient temperatures. DeNO/DeNO{sub x} by the pulsed-plasma process is mainly due to oxidation, but reduction takes place at the same time. When the wet-type reactor was used, the NO{sub 2} product was absorbed by water film and higher DeNO{sub x} efficiency could be achieved. Apart from laboratory tests on simulated gas mixtures, field tests were also carried out on the exhaust gas of an 8-kW diesel engine. A comparative analysis of the various tests are presented, together with a note on the energy consideration.

  11. Approach for control of high-density plasma reactors through optimal pulse shaping*

    E-Print Network [OSTI]

    Raja, Laxminarayan L.

    Approach for control of high-density plasma reactors through optimal pulse shaping* Tyrone L and it relies on a physical model of the plasma reactor used in conjunction with an optimal control algorithm surface in a HDP reactor can be biased separately to enable relatively independent control over plasma

  12. OPTIMAL DESIGN OF A HIGH PRESSURE ORGANOMETALLIC CHEMICAL VAPOR DEPOSITION REACTOR

    E-Print Network [OSTI]

    OPTIMAL DESIGN OF A HIGH PRESSURE ORGANOMETALLIC CHEMICAL VAPOR DEPOSITION REACTOR K.J. BACHMANN vapor deposition (HPOMCVD) reactor for use in thin film crystal growth. The advantages of such a reactor decomposition pressures and increased control over local stoichiometry and defect formation. While we focus here

  13. Thermal fuse for high-temperature batteries

    DOE Patents [OSTI]

    Jungst, Rudolph G. (Albuquerque, NM); Armijo, James R. (Albuquerque, NM); Frear, Darrel R. (Austin, TX)

    2000-01-01T23:59:59.000Z

    A thermal fuse, preferably for a high-temperature battery, comprising leads and a body therebetween having a melting point between approximately 400.degree. C. and 500.degree. C. The body is preferably an alloy of Ag--Mg, Ag--Sb, Al--Ge, Au--In, Bi--Te, Cd--Sb, Cu--Mg, In--Sb, Mg--Pb, Pb--Pd, Sb--Zn, Sn--Te, or Mg--Al.

  14. High temperature hot water distribution system study

    SciTech Connect (OSTI)

    NONE

    1996-12-01T23:59:59.000Z

    The existing High Temperature Hot Water (HTHW) Distribution System has been plagued with design and construction deficiencies since startup of the HTHW system, in October 1988. In October 1989, after one year of service, these deficiencies were outlined in a technical evaluation. The deficiencies included flooded manholes, sump pumps not hooked up, leaking valves, contaminated HTHW water, and no cathodic protection system. This feasibility study of the High Temperature Hot Water (HTHW) Distribution System was performed under Contract No. DACA0l-94-D-0033, Delivery Order 0013, Modification 1, issued to EMC Engineers, Inc. (EMC), by the Norfolk District Corps of Engineers, on 25 April 1996. The purpose of this study was to determine the existing conditions of the High Temperature Hot Water Distribution System, manholes, and areas of containment system degradation. The study focused on two areas of concern, as follows: * Determine existing conditions and areas of containment system degradation (leaks) in the underground carrier pipes and protective conduit. * Document the condition of underground steel and concrete manholes. To document the leaks, a site survey was performed, using state-of-the-art infrared leak detection equipment and tracer gas leak detection equipment. To document the condition of the manholes, color photographs were taken of the insides of 125 manholes, and notes were made on the condition of these manholes.

  15. THE PRODUCTION OF SYNGAS VIA HIGH TEMPERATURE ELECTROLYSIS AND BIO-MASS GASIFICATION

    SciTech Connect (OSTI)

    M. G. McKellar; G. L. Hawkes; J. E. O'Brien

    2008-11-01T23:59:59.000Z

    A process model of syngas production using high temperature electrolysis and biomass gasification is presented. Process heat from the biomass gasifier is used to improve the hydrogen production efficiency of the steam electrolysis process. Hydrogen from electrolysis allows a high utilization of the biomass carbon for syngas production. Based on the gasifier temperature, 94% to 95% of the carbon in the biomass becomes carbon monoxide in the syngas (carbon dioxide and hydrogen). Assuming the thermal efficiency of the power cycle for electricity generation is 50%, (as expected from GEN IV nuclear reactors), the syngas production efficiency ranges from 70% to 73% as the gasifier temperature decreases from 1900 K to 1500 K.

  16. Ultra high temperature diffusion apparatus and operating procedures

    SciTech Connect (OSTI)

    Wyrick, S.B.

    1985-11-15T23:59:59.000Z

    It is the purpose of this paper to present an experimental apparatus which is capable of measuring diffusion coefficients of interdiffusing gases in the temperature range 300K to 2500K. Because of the high temperatures which will be encountered, a special alloy of tantalum (T-111) is used to house the diffusion process. This T-111 diffusion cell is heated via radiation heat from a tungsten heating element powered by a Saban saturable reactor power supply. The diffusion cell heating element are encased in a nickel-plated copper cooling can. This entire assembly is enclosed in an Ultek vacuum chamber to prevent oxidation of the diffusion cell. This report covers the construction and calibration of the diffusion cell, details of the gas loading and sampling system, and complete information on the components required to operate the vacuum furnace. Thus far, several experiments have been run in the temperature range 600K to 800K and the resulting diffusion coefficients agree fairly well with previously published values. 21 refs., 9 figs., 4 tabs.

  17. High Flux Metallic Membranes for Hydrogen Recovery and Membrane Reactors

    SciTech Connect (OSTI)

    Buxbaum, Robert

    2010-06-30T23:59:59.000Z

    We made and tested over 250 new alloys for use as lower cost, higher flux hydrogen extraction membrane materials. Most of these were intermetallic, or contained significant intermetallic content, particularly based on B2 alloy compositions with at least one refractory component; B2 intermetallics resemble BCC alloys, in structure, but the atoms have relatively fixed positions, with one atom at the corners of the cube, the other at the centers. The target materals we were looking for would contain little or no expensive elements, no strongly toxic or radioactive elements, would have high flux to hydrogen, while being fabricable, brazable, and relatively immune to hydrogen embrittlement and corrosion in operation. The best combination of properties of the membrane materials we developed was, in my opinion, a Pd-coated membrane consisting of V -9 atomic % Pd. This material was relatively cheap, had 5 times the flux of Pd under the same pressure differential, was reasonably easy to fabricate and braze, and not bad in terms of embrittlement. Based on all these factors we project, about 1/3 the cost of Pd, on an area basis for a membrane designed to last 20 years, or 1/15 the cost on a flux basis. Alternatives to this membrane replaced significant fractions of the Pd with Ni and or Co. The cost for these membranes was lower, but so was the flux. We produced successful brazed products from the membrane materials, and made them into flat sheets. We tested, unsuccessfully, several means of fabricating thematerials into tubes, and eventually built a membrane reactor using a new, flat-plate design: a disc and doughnut arrangement, a design that seems well- suited to clean hydrogen production from coal. The membranes and reactor were tested successfully at Western Research. A larger equipment company (Chart Industries) produced similar results using a different flat-plate reactor design. Cost projections of the membrane are shown to be attractive.

  18. A HIGH TEMPERATURE GAS RECEIVER UTILIZING SMALL PARTICLES

    E-Print Network [OSTI]

    Hunt, Arlon

    2012-01-01T23:59:59.000Z

    field of high temperature solar process heat. The ultimateof solar applications including industrial process heat and

  19. Density of Gadolinium Nitrate Solutions for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Taylor, Paul Allen [ORNL; Lee, Denise L [ORNL

    2009-05-01T23:59:59.000Z

    In late 1992, the High Flux Isotope Reactor (HFIR) was planning to switch the solution contained in the poison injection tank from cadmium nitrate to gadolinium nitrate. The poison injection system is an emergency system used to shut down the reactor by adding a neutron poison to the cooling water. This system must be able to supply a minimum of 69 pounds of gadolinium to the reactor coolant system in order to guarantee that the reactor would become subcritical. A graph of the density of gadolinium nitrate solutions over a concentration range of 5 to 30 wt% and a temperature range of 15 to 40{sup o}C was prepared. Routine density measurements of the solution in the poison injection tank are made by HFIR personnel, and an adaptation of the original graph is used to determine the gadolinium nitrate concentration. In late 2008, HFIR personnel decided that the heat tracing that was present on the piping for the poison injection system could be removed without any danger of freezing the solution; however, the gadolinium nitrate solution might get as cold as 5{sup o}C. This was outside the range of the current density-concentration correlation, so the range needed to be expanded. This report supplies a new density-concentration correlation that covers the extended temperature range. The correlation is given in new units, which greatly simplifies the calculation that is required to determine the pounds of gadolinium in the tank solution. The procedure for calculating the amount of gadolinium in the HFIR poison injection system is as follows: (1) Calculate the usable volume in the system; (2) Measure the density of the solution; (3) Calculate the gadolinium concentration using the following equation: Gd(lb/ft{sup 3}) = measured density (g/mL) x 34.681 - 34.785; (4) Calculate the amount of gadolinium in the system using the following equation: Amount of Gd(lb) = Gd concentration (lb/ft{sup 3}) x usable volume (ft{sup 3}). The equation in step 3 is exact for a temperature of 5{sup o}C, and overestimates the gadolinium concentration at all higher temperatures. This guarantees that the calculation is conservative, in that the actual concentration will be at least as high as that calculated. If an additional safety factor is desired, it is recommended that an administrative control limit be set that is higher than the required minimum amount of gadolinium.

  20. High Temperature Battery for Drilling Applications

    SciTech Connect (OSTI)

    Josip Caja

    2009-12-31T23:59:59.000Z

    In this project rechargeable cells based on the high temperature electrochemical system Na/beta''-alumina/S(IV) in AlCl3/NaCl were developed for application as an autonomous power source in oil/gas deep drilling wells. The cells operate in the temperature range from 150 C to 250 C. A prototype DD size cell was designed and built based on the results of finite element analysis and vibration testing. The cell consisted of stainless steel case serving as anode compartment with cathode compartment installed in it and a seal closing the cell. Critical element in cell design and fabrication was hermetically sealing the cell. The seal had to be leak tight, thermally and vibration stable and compatible with electrode materials. Cathode compartment was built of beta''-alumina tube which served as an electrolyte, separator and cathode compartment.

  1. High Flux Isotope Reactor (HFIR) | U.S. DOE Office of Science...

    Office of Science (SC) Website

    (SUF) Division SUF Home About User Facilities User Facilities Dev X-Ray Light Sources Neutron Scattering Facilities High Flux Isotope Reactor (HFIR) Lujan Neutron Scattering...

  2. System Analyses of High and Low-Temperature Interface Designs for a Nuclear-Driven High-Temperature Electrolysis Hydrogen Production Plant

    SciTech Connect (OSTI)

    E. A. Harvego; J. E. O'Brien

    2009-07-01T23:59:59.000Z

    As part of the Next Generation Nuclear Plant (NGNP) project, an evaluation of a low-temperature heat-pump interface design for a nuclear-driven high-temperature electrolysis (HTE) hydrogen production plant was performed using the UniSim process analysis software. The lowtemperature interface design is intended to reduce the interface temperature between the reactor power conversion system and the hydrogen production plant by extracting process heat from the low temperature portion of the power cycle rather than from the high-temperature portion of the cycle as is done with the current Idaho National Laboratory (INL) reference design. The intent of this design change is to mitigate the potential for tritium migration from the reactor core to the hydrogen plant, and reduce the potential for high temperature creep in the interface structures. The UniSim model assumed a 600 MWt Very-High Temperature Reactor (VHTR) operating at a primary system pressure of 7.0 MPa and a reactor outlet temperature of 900C. The lowtemperature heat-pump loop is a water/steam loop that operates between 2.6 MPa and 5.0 MPa. The HTE hydrogen production loop operated at 5 MPa, with plant conditions optimized to maximize plant performance (i.e., 800C electrolysis operating temperature, area specific resistance (ASR) = 0.4 ohm-cm2, and a current density of 0.25 amps/cm2). An air sweep gas system was used to remove oxygen from the anode side of the electrolyzer. Heat was also recovered from the hydrogen and oxygen product streams to maximize hydrogen production efficiencies. The results of the UniSim analysis showed that the low-temperature interface design was an effective heat-pump concept, transferring 31.5 MWt from the low-temperature leg of the gas turbine power cycle to the HTE process boiler, while consuming 16.0 MWe of compressor power. However, when this concept was compared with the current INL reference direct Brayton cycle design and with a modification of the reference design to simulate an indirect Brayton cycle (both with heat extracted from the high-temperature portion of the power cycle), the latter two concepts had higher overall hydrogen production rates and efficiencies compared to the low-temperature heatpump concept, but at the expense of higher interface temperatures. Therefore, the ultimate decision on the viability of the low-temperature heat-pump concept involves a tradeoff between the benefits of a lower-temperature interface between the power conversion system and the hydrogen production plant, and the reduced hydrogen production efficiency of the low-temperature heat-pump concept compared to concepts using high-temperature process heat.

  3. High temperature regenerable hydrogen sulfide removal agents

    DOE Patents [OSTI]

    Copeland, Robert J. (Wheat Ridge, CO)

    1993-01-01T23:59:59.000Z

    A system for high temperature desulfurization of coal-derived gases using regenerable sorbents. One sorbent is stannic oxide (tin oxide, SnO.sub.2), the other sorbent is a metal oxide or mixed metal oxide such as zinc ferrite (ZnFe.sub.2 O.sub.4). Certain otherwise undesirable by-products, including hydrogen sulfide (H.sub.2 S) and sulfur dioxide (SO.sub.2) are reused by the system, and elemental sulfur is produced in the regeneration reaction. A system for refabricating the sorbent pellets is also described.

  4. High Temperature Materials Laboratory third annual report

    SciTech Connect (OSTI)

    Tennery, V.J.; Foust, F.M.

    1990-12-01T23:59:59.000Z

    The High Temperature Materials Laboratory has completed its third year of operation as a designated DOE User Facility at the Oak Ridge National Laboratory. Growth of the user program is evidenced by the number of outside institutions who have executed user agreements since the facility began operation in 1987. A total of 88 nonproprietary agreements (40 university and 48 industry) and 20 proprietary agreements (1 university, 19 industry) are now in effect. Sixty-eight nonproprietary research proposals (39 from university, 28 from industry, and 1 other government facility) and 8 proprietary proposals were considered during this reporting period. Research projects active in FY 1990 are summarized.

  5. Modeling forces in high-temperature superconductors

    SciTech Connect (OSTI)

    Turner, L. R.; Foster, M. W.

    1997-11-18T23:59:59.000Z

    We have developed a simple model that uses computed shielding currents to determine the forces acting on a high-temperature superconductor (HTS). The model has been applied to measurements of the force between HTS and permanent magnets (PM). Results show the expected hysteretic variation of force as the HTS moves first toward and then away from a permanent magnet, including the reversal of the sign of the force. Optimization of the shielding currents is carried out through a simulated annealing algorithm in a C++ program that repeatedly calls a commercial electromagnetic software code. Agreement with measured forces is encouraging.

  6. Enabling High Efficiency Low Temperature Combustion by Adaptive...

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

    Low Temperature Combustion by Adaptive In-Situ Jet Cooling Enabling High Efficiency Low Temperature Combustion by Adaptive In-Situ Jet Cooling A new approach, called...

  7. High and Low Temperature Series Estimates for the Critical Temperature of the 3D Ising Model

    E-Print Network [OSTI]

    Adler, Joan

    High and Low Temperature Series Estimates for the Critical Temperature of the 3D Ising Model Zaher Abstract We have analysed low and high temperature series expansions for the threedimensional Ising model on the simple cubic lattice. Our analysis of Butera and Comi's new 32 term high temperature series yields K c

  8. High and Low Temperature Series Estimates for the Critical Temperature of the 3D Ising Model

    E-Print Network [OSTI]

    Adler, Joan

    High and Low Temperature Series Estimates for the Critical Temperature Abstract We have analysed low and high temperature series expansions for the three high temperature series yields Kc = 0.221659 +0.000002-0.000005and from the 32 term low

  9. High temperature insulation for ceramic matrix composites

    SciTech Connect (OSTI)

    Merrill, Gary B. (Monroeville, PA); Morrison, Jay Alan (Orlando, FL)

    2000-01-01T23:59:59.000Z

    A ceramic composition is provided to insulate ceramic matrix composites under high temperature, high heat flux environments. The composite comprises a plurality of hollow oxide-based spheres of varios dimentions, a phosphate binder, and at least one oxide filler powder, whereby the phosphate binder partially fills gaps between the spheres and the filler powders. The spheres are situated in the phosphate binder and the filler powders such that each sphere is in contact with at least one other sphere. The spheres may be any combination of Mullite spheres, Alumina spheres, or stabilized Zirconia spheres. The filler powder may be any combination of Alumina, Mullite, Ceria, or Hafnia. Preferably, the phosphate binder is Aluminum Ortho-Phosphate. A method of manufacturing the ceramic insulating composition and its application to CMC substates are also provided.

  10. High temperature insulation for ceramic matrix composites

    DOE Patents [OSTI]

    Merrill, Gary B.; Morrison, Jay Alan

    2004-01-13T23:59:59.000Z

    A ceramic composition is provided to insulate ceramic matrix composites under high temperature, high heat flux environments. The composition comprises a plurality of hollow oxide-based spheres of various dimensions, a phosphate binder, and at least one oxide filler powder, whereby the phosphate binder partially fills gaps between the spheres and the filler powders. The spheres are situated in the phosphate binder and the filler powders such that each sphere is in contact with at least one other sphere. The spheres may be any combination of Mullite spheres, Alumina spheres, or stabilized Zirconia spheres. The filler powder may be any combination of Alumina, Mullite, Ceria, or Hafnia. Preferably, the phosphate binder is Aluminum Ortho-Phosphate. A method of manufacturing the ceramic insulating composition and its application to CMC substrates are also provided.

  11. High temperature insulation for ceramic matrix composites

    SciTech Connect (OSTI)

    Merrill, Gary B. (Monroeville, PA); Morrison, Jay Alan (Orlando, FL)

    2001-01-01T23:59:59.000Z

    A ceramic composition is provided to insulate ceramic matrix composites under high temperature, high heat flux environments. The composition comprises a plurality of hollow oxide-based spheres of various dimensions, a phosphate binder, and at least one oxide filler powder, whereby the phosphate binder partially fills gaps between the spheres and the filler powders. The spheres are situated in the phosphate binder and the filler powders such that each sphere is in contact with at least one other sphere. The spheres may be any combination of Mullite spheres, Alumina spheres, or stabilized Zirconia spheres. The filler powder may be any combination of Alumina, Mullite, Ceria, or Hafnia. Preferably, the phosphate binder is Aluminum Ortho-Phosphate. A method of manufacturing the ceramic insulating composition and its application to CMC substrates are also provided.

  12. Process Heat Exchanger Options for Fluoride Salt High Temperature Reactor

    SciTech Connect (OSTI)

    Piyush Sabharwall; Eung Soo Kim; Michael McKellar; Nolan Anderson

    2011-04-01T23:59:59.000Z

    The work reported herein is a significant intermediate step in reaching the final goal of commercial-scale deployment and usage of molten salt as the heat transport medium for process heat applications. The primary purpose of this study is to aid in the development and selection of the required heat exchanger for power production and process heat application, which would support large-scale deployment.

  13. HIGH TEMPERATURE GAS-COOLED REACTOR KNOWLEDGE MANAGEMENT

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

    be scrammed without regard to the onset of a core-conduction cool down without active cooling. However, even if all control and shutdown rods are scrammed, the operator must...

  14. Process Heat Exchanger Options for the Advanced High Temperature Reactor

    SciTech Connect (OSTI)

    Piyush Sabharwall; Eung Soo Kim; Michael McKellar; Nolan Anderson

    2011-06-01T23:59:59.000Z

    The work reported herein is a significant intermediate step in reaching the final goal of commercial-scale deployment and usage of molten salt as the heat transport medium for process heat applications. The primary purpose of this study is to aid in the development and selection of the required heat exchanger for power production and process heat application, which would support large-scale deployment.

  15. Fabrication and Characterization of Uranium-based High Temperature Reactor

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville Power AdministrationField8,Dist. Category UC-lFederalFY 2008 FOIAFabricated Metals (2010Fuel |

  16. Safeguards Guidance for Prismatic Fueled High Temperature Gas Reactors (HTGR)

    National Nuclear Security Administration (NNSA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA Approved:AdministrationAnalysis andBHoneywell9/%2ARequest forMod0/%2A21)

  17. Neutronic and thermal calculation of blanket for high power operating condition of fusion reactor

    SciTech Connect (OSTI)

    Sagawa, H.; Shimakawa, S.; Kuroda, T. [Oarai Research Establishement of JAERI, Ibaraki (Japan)] [and others

    1994-12-31T23:59:59.000Z

    Internal (breeding region) structures of ceramic breeder blanket to accommodate high power operating conditions such as a DEMO reactor have been investigated. The conditions considered here are the maximum neutron wall load of 2.8 MW/m{sup 2} at outboard midplane corresponding to a fusion power of 3.0 GW and the coolant temperature of 200{degrees}C. Structure of a blanket is based on the layered pebble bed concept, which has been proposed by Japan since the ITER CDA. Lithium oxide with 50% enriched {sup 6}Li is used in a shape of small spherical pebbles which are filled in a 316SS can avoid its compatibility issue with Be. Beryllium around the breeder can is filled also in a shape of spherical pebbles which works not only as a neutron multiplier but also as a thermal resistant layer to maintain breeder temperature for effective in-situ tritium recovery. Diameters and packing fractions of both pebbles are {<=} 1 mm and 65%, respectively. A layer of block Be between cooling panels is introduced as a neutron multiplier (not as the thermal resistant layer) to enhance tritium breeding performance. Inlet temperature of water coolant is 200{degrees}C to meet the high temperature conditioning requirement to the first wall which is one of walls of the blanket vessel. Neutronics calculations have been carried out by one-dimensional transport code, and thermal calculations have also been carried out by one-dimensional slab code.

  18. High-temperature, high-pressure bonding of nested tubular metallic components

    DOE Patents [OSTI]

    Quinby, Thomas C. (Kingston, TN)

    1980-01-01T23:59:59.000Z

    This invention is a tool for effecting high-temperature, high-compression bonding between the confronting faces of nested, tubular, metallic components. In a typical application, the tool is used to produce tubular target assemblies for irradiation in nuclear reactors or particle accelerators, the target assembly comprising a uranium foil and an aluminum-alloy substrate. The tool preferably is composed throughout of graphite. It comprises a tubular restraining member in which a mechanically expandable tubular core is mounted to form an annulus with the member. The components to be bonded are mounted in nested relation in the annulus. The expandable core is formed of individually movable, axially elongated segments whose outer faces cooperatively define a cylindrical pressing surface and whose inner faces cooperatively define two opposed, inwardly tapered, axial bores. Tapered rams extend respectively into the bores. The loaded tool is mounted in a conventional hot-press provided with evacuation means, heaters for maintaining its interior at bonding temperature, and hydraulic cylinders for maintaining a selected inwardly directed pressure on the tapered rams. With the hot-press evacuated and the loaded tool at the desired temperature, the cylinders are actuated to apply the selected pressure to the rams. The rams in turn expand the segmented core to maintain the nested components in compression against the restraining member. These conditions are maintained until the confronting faces of the nested components are joined in a continuous, uniform bond characterized by high thermal conductivity.

  19. High temperature coatings for gas turbines

    DOE Patents [OSTI]

    Zheng, Xiaoci Maggie

    2003-10-21T23:59:59.000Z

    Coating for high temperature gas turbine components that include a MCrAlX phase, and an aluminum-rich phase, significantly increase oxidation and cracking resistance of the components, thereby increasing their useful life and reducing operating costs. The aluminum-rich phase includes aluminum at a higher concentration than aluminum concentration in the MCrAlX alloy, and an aluminum diffusion-retarding composition, which may include cobalt, nickel, yttrium, zirconium, niobium, molybdenum, rhodium, cadmium, indium, cerium, iron, chromium, tantalum, silicon, boron, carbon, titanium, tungsten, rhenium, platinum, and combinations thereof, and particularly nickel and/or rhenium. The aluminum-rich phase may be derived from a particulate aluminum composite that has a core comprising aluminum and a shell comprising the aluminum diffusion-retarding composition.

  20. Turbine vane with high temperature capable skins

    DOE Patents [OSTI]

    Morrison, Jay A. (Oviedo, FL)

    2012-07-10T23:59:59.000Z

    A turbine vane assembly includes an airfoil extending between an inner shroud and an outer shroud. The airfoil can include a substructure having an outer peripheral surface. At least a portion of the outer peripheral surface is covered by an external skin. The external skin can be made of a high temperature capable material, such as oxide dispersion strengthened alloys, intermetallic alloys, ceramic matrix composites or refractory alloys. The external skin can be formed, and the airfoil can be subsequently bi-cast around or onto the skin. The skin and the substructure can be attached by a plurality of attachment members extending between the skin and the substructure. The skin can be spaced from the outer peripheral surface of the substructure such that a cavity is formed therebetween. Coolant can be supplied to the cavity. Skins can also be applied to the gas path faces of the inner and outer shrouds.

  1. High temperature low friction surface coating

    DOE Patents [OSTI]

    Bhushan, Bharat (Watervliet, NY)

    1980-01-01T23:59:59.000Z

    A high temperature, low friction, flexible coating for metal surfaces which are subject to rubbing contact includes a mixture of three parts graphite and one part cadmium oxide, ball milled in water for four hours, then mixed with thirty percent by weight of sodium silicate in water solution and a few drops of wetting agent. The mixture is sprayed 12-15 microns thick onto an electro-etched metal surface and air dried for thirty minutes, then baked for two hours at 65.degree. C. to remove the water and wetting agent, and baked for an additional eight hours at about 150.degree. C. to produce the optimum bond with the metal surface. The coating is afterwards burnished to a thickness of about 7-10 microns.

  2. Multilayer ultra-high-temperature ceramic coatings

    DOE Patents [OSTI]

    Loehman, Ronald E. (Albuquerque, NM); Corral, Erica L. (Tucson, AZ)

    2012-03-20T23:59:59.000Z

    A coated carbon-carbon composite material with multiple ceramic layers to provide oxidation protection from ultra-high-temperatures, where if the carbon-carbon composite material is uninhibited with B.sub.4C particles, then the first layer on the composite material is selected from ZrB.sub.2 and HfB.sub.2, onto which is coated a layer of SiC coated and if the carbon-carbon composite material is inhibited with B.sub.4C particles, then protection can be achieved with a layer of SiC and a layer of either ZrB.sub.2 and HfB.sub.2 in any order.

  3. Assessment of microelectronics packaging for high temperature, high reliability applications

    SciTech Connect (OSTI)

    Uribe, F.

    1997-04-01T23:59:59.000Z

    This report details characterization and development activities in electronic packaging for high temperature applications. This project was conducted through a Department of Energy sponsored Cooperative Research and Development Agreement between Sandia National Laboratories and General Motors. Even though the target application of this collaborative effort is an automotive electronic throttle control system which would be located in the engine compartment, results of this work are directly applicable to Sandia`s national security mission. The component count associated with the throttle control dictates the use of high density packaging not offered by conventional surface mount. An enabling packaging technology was selected and thermal models defined which characterized the thermal and mechanical response of the throttle control module. These models were used to optimize thick film multichip module design, characterize the thermal signatures of the electronic components inside the module, and to determine the temperature field and resulting thermal stresses under conditions that may be encountered during the operational life of the throttle control module. Because the need to use unpackaged devices limits the level of testing that can be performed either at the wafer level or as individual dice, an approach to assure a high level of reliability of the unpackaged components was formulated. Component assembly and interconnect technologies were also evaluated and characterized for high temperature applications. Electrical, mechanical and chemical characterizations of enabling die and component attach technologies were performed. Additionally, studies were conducted to assess the performance and reliability of gold and aluminum wire bonding to thick film conductor inks. Kinetic models were developed and validated to estimate wire bond reliability.

  4. High-Temperature Solar Selective Coating Development for Power...

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

    High-Temperature Solar Selective Coating Development for Power Tower Receivers - FY13 Q1 High-Temperature Solar Selective Coating Development for Power Tower Receivers - FY13 Q1...

  5. High-Temperature Solar Selective Coating Development for Power...

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

    High-Temperature Solar Selective Coating Development for Power Tower Receivers - FY13 Q2 High-Temperature Solar Selective Coating Development for Power Tower Receivers - FY13 Q2...

  6. Development of a High-Temperature Diagnostics-While-Drilling...

    Energy Savers [EERE]

    Development of a High-Temperature Diagnostics-While-Drilling Tool Development of a High-Temperature Diagnostics-While-Drilling Tool This report documents work performed in the...

  7. Corrosion Studies in High-Temperature Molten Salt Systems for...

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

    Corrosion Studies in High-Temperature Molten Salt Systems for CSP Applications - FY13 Q1 Corrosion Studies in High-Temperature Molten Salt Systems for CSP Applications - FY13 Q1...

  8. Fundamental Corrosion Studies in High-Temperature Molten Salt...

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

    Fundamental Corrosion Studies in High-Temperature Molten Salt Systems for Next-Generation CSP Systems - FY13 Q2 Fundamental Corrosion Studies in High-Temperature Molten Salt...

  9. Modularity of the MIT Pebble Bed Reactor for use by the commercial power industry

    E-Print Network [OSTI]

    Hanlon-Hyssong, Jaime E

    2008-01-01T23:59:59.000Z

    The Modular Pebble Bed Reactor is a small high temperature helium cooled reactor that is being considered for both electric power and hydrogen production. Pebble bed reactors are being developed in South Africa, China and ...

  10. High Temperature High Pressure Thermodynamic Measurements for Coal Model Compounds

    SciTech Connect (OSTI)

    John C. Chen; Vinayak N. Kabadi

    1998-11-12T23:59:59.000Z

    The overall objective of this project is to develop a better thermodynamic model for predicting properties of high-boiling coal derived liquids, especially the phase equilibria of different fractions at elevated temperatures and pressures. The development of such a model requires data on vapor-liquid equilibria (VLE), enthalpy, and heat capacity which would be experimentally determined for binary systems of coal model compounds and compiled into a database. The data will be used to refine existing models such as UNIQUAC and UNIFAC. The flow VLE apparatus designed and built for a previous project was upgraded and recalibrated for data measurements for thk project. The modifications include better and more accurate sampling technique and addition of a digital recorder to monitor temperature, pressure and liquid level inside the VLE cell. VLE data measurements for system benzene-ethylbenzene have been completed. The vapor and liquid samples were analysed using the Perkin-Elmer Autosystem gas chromatography.

  11. High Temperature Integrated Thermoelectric Ststem and Materials

    SciTech Connect (OSTI)

    Mike S. H. Chu

    2011-06-06T23:59:59.000Z

    The final goal of this project is to produce, by the end of Phase II, an all ceramic high temperature thermoelectric module. Such a module design integrates oxide ceramic n-type, oxide ceramic p-type materials as thermoelectric legs and oxide ceramic conductive material as metalizing connection between n-type and p-type legs. The benefits of this all ceramic module are that it can function at higher temperatures (> 700 C), it is mechanically and functionally more reliable and it can be scaled up to production at lower cost. With this all ceramic module, millions of dollars in savings or in new opportunities recovering waste heat from high temperature processes could be made available. A very attractive application will be to convert exhaust heat from a vehicle to reusable electric energy by a thermoelectric generator (TEG). Phase I activities were focused on evaluating potential n-type and p-type oxide compositions as the thermoelectric legs. More than 40 oxide ceramic powder compositions were made and studied in the laboratory. The compositions were divided into 6 groups representing different material systems. Basic ceramic properties and thermoelectric properties of discs sintered from these powders were measured. Powders with different particles sizes were made to evaluate the effects of particle size reduction on thermoelectric properties. Several powders were submitted to a leading thermoelectric company for complete thermoelectric evaluation. Initial evaluation showed that when samples were sintered by conventional method, they had reasonable values of Seebeck coefficient but very low values of electrical conductivity. Therefore, their power factors (PF) and figure of merits (ZT) were too low to be useful for high temperature thermoelectric applications. An unconventional sintering method, Spark Plasma Sintering (SPS) was determined to produce better thermoelectric properties. Particle size reduction of powders also was found to have some positive benefits. Two composition systems, specifically 1.0 SrO - 0.8 x 1.03 TiO2 - 0.2 x 1.03 NbO2.5 and 0.97 TiO2 - 0.03 NbO2.5, have been identified as good base line compositions for n-type thermoelectric compositions in future module design. Tests of these materials at an outside company were promising using that company's processing and material expertise. There was no unique p-type thermoelectric compositions identified in phase I work other than several current cobaltite materials. Ca3Co4O9 will be the primary p-type material for the future module design until alternative materials are developed. BaTiO3 and rare earth titanate based dielectric compositions show both p-type and n-type behavior even though their electrical conductivities were very low. Further research and development of these materials for thermoelectric applications is planned in the future. A preliminary modeling and optimization of a thermoelectric generator (TEG) that uses the n-type 1.0 SrO - 1.03 x 0.8 TiO2 - 1.03 x 0.2 NbO2.5 was performed. Future work will combine development of ceramic powders and manufacturing expertise at TAM, development of SPS at TAM or a partner organization, and thermoelectric material/module testing, modeling, optimization, production at several partner organizations.

  12. High Temperature Membrane Working Group Meeting, May 14, 2007

    Broader source: Energy.gov [DOE]

    This agenda provides information about the High Temperature Membrane Working Group Meeting on May 14, 2007 in Arlington, Va.

  13. Agenda for the High Temperature Membrane Working Group Meeting

    Broader source: Energy.gov [DOE]

    This agenda provides information about the Agenda for the High Temperature Membrane Working Group Meeting on September 14, 2006.

  14. Fabrication and Design Aspects of High-Temperature Compact Diffusion Bonded Heat Exchangers

    SciTech Connect (OSTI)

    Sai K. Mylavarapu; Richard N. Christensen; Raymond R. Unocic; Richard E. Glosup; Mike W. Patterson

    2012-08-01T23:59:59.000Z

    The Very High Temperature Reactor (VHTR) using gas-cooled reactor technology is anticipated to be the reactor type for the Next Generation Nuclear Plant (NGNP). In this reactor concept with an indirect power cycle system, a high-temperature and high integrity Intermediate Heat Exchanger (IHX) with high effectiveness is required to efficiently transfer the core thermal output to a secondary fluid for electricity generation, hydrogen production, and/or industrial process heat applications. At present, there is no proven IHX concept for VHTRs. The current Technology Readiness Level (TRL) status issued by NGNP to all components associated with the IHX for reduced nominal reactor outlet temperatures of 750800 degrees C is 3 on a 110 scale, with 10 indicating omplete technological maturity. Among the various potential IHX concepts available, diffusion bonded heat exchangers (henceforth called printed circuit heat exchangers, or PCHEs) appear promising for NGNP applications. The design and fabrication of this key component of NGNP with Alloy 617, a candidate high-temperature structural material for NGNP applications, are the primary focus of this paper. In the current study, diffusion bonding of Alloy 617 has been demonstrated, although the optimum diffusion bonding process parameters to engineer a quasi interface-free joint are yet to be determined. The PCHE fabrication related processes, i.e., photochemical etching and diffusion bonding are discussed for Alloy 617 plates. In addition, the authors experiences with these non-conventional machining and joining techniques are discussed. Two PCHEs are fabricated using Alloy 617 plates and are being experimentally investigated for their thermal-hydraulic performance in a High-Temperature Helium Facility (HTHF). The HTHF is primarily of Alloy 800H construction and is designed to facilitate experiments at temperatures and pressures up to 800 degrees C and 3 MPa, respectively. Furthermore, some preliminary microstructural and mechanical property characterization studies of representative diffusion bonded Alloy 617 specimens are presented. The characterization studies are restricted and less severe from an NGNP perspective but provide sufficient confidence to ensure safe operation of the heat exchangers in the HTHF. The test results are used to determine the design operating conditions for the PCHEs fabricated.

  15. High Temperature Interactions of Antimony with Nickel

    SciTech Connect (OSTI)

    Marina, Olga A.; Pederson, Larry R.

    2012-07-01T23:59:59.000Z

    In this chapter, the surface and bulk interactions of antimony with the Ni-based anodes in solid oxide fuel cells (SOFC) will be discussed. High fuel flexibility is a significant advantage of SOFCs, allowing the direct use of fossil and bio fuels without a hydrogen separation unit. Synthesis gas derived from coal and biomass consists of a mixture of hydrogen, carbon monoxide, carbon dioxide, and steam, but finite amounts of tars and trace impurities such as S, Se, P, As, Sb, Cd, Pb, Cl, etc, are also always present. While synthesis gas is commonly treated with a series of chemical processes and scrubbers to remove the impurities, complete purification is not economical. Antimony is widely distributed in coals. During coal gasification antimony is volatilized, such that contact with the SOFC anodes and other SOFC parts, e.g., interconnect, current collecting wires, fuel gas supplying tubing, is most likely. This chapter addresses the following topics: high temperature Ni - Sb interactions; alteration phase, Ni3Sb, Ni5Sb2, NiSb, formation; thermochemical modeling; impact of Sb on the electrocatalytic activity of Ni toward the fuel oxidation and the presence of other impurities (sulfur, in particular); converted anode structural instability during long-term SOFC operation; comparison with nickel heterogeneous catalysts.

  16. Integrated High Temperature Coal-to-Hydrogen System with CO2 Separation

    SciTech Connect (OSTI)

    James A. Ruud; Anthony Ku; Vidya Ramaswamy; Wei Wei; Patrick Willson

    2007-05-31T23:59:59.000Z

    A significant barrier to the commercialization of coal-to-hydrogen technologies is high capital cost. The purity requirements for H{sub 2} fuels are generally met by using a series of unit clean-up operations for residual CO removal, sulfur removal, CO{sub 2} removal and final gas polishing to achieve pure H{sub 2}. A substantial reduction in cost can be attained by reducing the number of process operations for H{sub 2} cleanup, and process efficiency can be increased by conducting syngas cleanup at higher temperatures. The objective of this program was to develop the scientific basis for a single high-temperature syngas-cleanup module to produce a pure stream of H{sub 2} from a coal-based system. The approach was to evaluate the feasibility of a 'one box' process that combines a shift reactor with a high-temperature CO{sub 2}-selective membrane to convert CO to CO{sub 2}, remove sulfur compounds, and remove CO{sub 2} in a simple, compact, fully integrated system. A system-level design was produced for a shift reactor that incorporates a high-temperature membrane. The membrane performance targets were determined. System level benefits were evaluated for a coal-to-hydrogen system that would incorporate membranes with properties that would meet the performance targets. The scientific basis for high temperature CO{sub 2}-selective membranes was evaluated by developing and validating a model for high temperature surface flow membranes. Synthesis approaches were pursued for producing membranes that integrated control of pore size with materials adsorption properties. Room temperature reverse-selectivity for CO{sub 2} was observed and performance at higher temperatures was evaluated. Implications for future membrane development are discussed.

  17. Enhanced High Temperature Performance of NOx Storage/Reduction...

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

    Meeting ace026peden2012o.pdf More Documents & Publications Enhanced High Temperature Performance of NOx StorageReduction (NSR) Materials Enhanced High and Low...

  18. NETL - Chemical Looping Reactor

    ScienceCinema (OSTI)

    None

    2014-06-26T23:59:59.000Z

    NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

  19. NETL - Chemical Looping Reactor

    SciTech Connect (OSTI)

    None

    2013-07-24T23:59:59.000Z

    NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

  20. Geochemistry of Aluminum in High Temperature Brines

    SciTech Connect (OSTI)

    Benezeth, P.; Palmer, D.A.; Wesolowski, D.J.

    1999-05-18T23:59:59.000Z

    The objective ofthis research is to provide quantitative data on the equilibrium and thermodynamic properties of aluminum minerals required to model changes in permeability and brine chemistry associated with fluid/rock interactions in the recharge, reservoir, and discharge zones of active geothermal systems. This requires a precise knowledge of the thermodynamics and speciation of aluminum in aqueous brines, spanning the temperature and fluid composition rangesencountered in active systems. The empirical and semi-empirical treatments of the solubility/hydrolysis experimental results on single aluminum mineral phases form the basis for the ultimate investigation of the behavior of complex aluminosilicate minerals. The principal objective in FY 1998 was to complete the solubility measurements on boehmite (AIOOH) inNaC1 media( 1 .O and 5.0 molal ionic strength, IOO-250C). However, additional measurements were also made on boehmite solubility in pure NaOH solutions in order to bolster the database for fitting in-house isopiestic data on this system. Preliminary kinetic Measurements of the dissolution/precipitation of boehmite was also carried out, although these were also not planned in the earlier objective. The 1999 objectives are to incorporate these treatments into existing codes used by the geothermal industry to predict the chemistry ofthe reservoirs; these calculations will be tested for reliability against our laboratory results and field observations. Moreover, based on the success of the experimental methods developed in this program, we intend to use our unique high temperature pH easurement capabilities to make kinetic and equilibrium studies of pH-dependent aluminosilicate transformation reactions and other pH-dependent heterogeneous reactions.

  1. A temperature compensated pressure transducer for high temperature, high pressure applications

    E-Print Network [OSTI]

    Lippka, Sandra Margaret

    1991-01-01T23:59:59.000Z

    and content by: 2/J David G. ansson (Chair ol' Committee) c. Y~ Christian P Burger (i&Iember) Randall Getger ( Member) 5wc Fr~. Walter F. Bradley (Head of Department) May 1991 ABSTRACT A Temperature Compensated Pressure Transducer for High... of the light beam. A compensation schenle is provided through the use of thermally arljusting reflecting surfaces These surfaces can adjust for temperatures up to 1000'F with less than a. I, c error. The final light beam movenlent across the photodiode face...

  2. High performance internal reforming unit for high temperature fuel cells

    DOE Patents [OSTI]

    Ma, Zhiwen (Sandy Hook, CT); Venkataraman, Ramakrishnan (New Milford, CT); Novacco, Lawrence J. (Brookfield, CT)

    2008-10-07T23:59:59.000Z

    A fuel reformer having an enclosure with first and second opposing surfaces, a sidewall connecting the first and second opposing surfaces and an inlet port and an outlet port in the sidewall. A plate assembly supporting a catalyst and baffles are also disposed in the enclosure. A main baffle extends into the enclosure from a point of the sidewall between the inlet and outlet ports. The main baffle cooperates with the enclosure and the plate assembly to establish a path for the flow of fuel gas through the reformer from the inlet port to the outlet port. At least a first directing baffle extends in the enclosure from one of the sidewall and the main baffle and cooperates with the plate assembly and the enclosure to alter the gas flow path. Desired graded catalyst loading pattern has been defined for optimized thermal management for the internal reforming high temperature fuel cells so as to achieve high cell performance.

  3. NGNP High Temperature Materials White Paper

    SciTech Connect (OSTI)

    Lew Lommers; George Honma

    2012-08-01T23:59:59.000Z

    This white paper is one in a series of white papers that address key generic issues of the combined construction and operating license (COL) pre-application program key generic issues for the Next Generation Nuclear Plant reactor using the prismatic block fuel technology. The purpose of the pre-application program interactions with the NRC staff is to reduce the time required for COL application review by identifying and addressing key regulatory issues and, if possible, obtaining agreements for their resolution

  4. Crevice corrosion repassivation temperatures of highly alloyed stainless steels

    SciTech Connect (OSTI)

    Valen, S.; Gartland, P.O. [SINTEF Corrosion Center, Trondheim (Norway)

    1995-10-01T23:59:59.000Z

    An investigation was conducted to study the repassivation temperature of a highly alloyed austenitic (UNS S31254) and of a highly alloyed duplex (UNS S32750) stainless steel (SS). When initiated at a high temperature, repassivation occurred at a temperature level significantly lower than normally associated with initiation of crevice corrosion. Experimental results combined with computer modeling of crevice corrosion explored the mechanistic aspects. In this respect, the similarity between the hysteresis observed by cyclic polarization and cyclic temperature tests was emphasized.

  5. Pressure Resistance Welding of High Temperature Metallic Materials

    SciTech Connect (OSTI)

    N. Jerred; L. Zirker; I. Charit; J. Cole; M. Frary; D. Butt; M. Meyer; K. L. Murty

    2010-10-01T23:59:59.000Z

    Pressure Resistance Welding (PRW) is a solid state joining process used for various high temperature metallic materials (Oxide dispersion strengthened alloys of MA957, MA754; martensitic alloy HT-9, tungsten etc.) for advanced nuclear reactor applications. A new PRW machine has been installed at the Center for Advanced Energy Studies (CAES) in Idaho Falls for conducting joining research for nuclear applications. The key emphasis has been on understanding processing-microstructure-property relationships. Initial studies have shown that sound joints can be made between dissimilar materials such as MA957 alloy cladding tubes and HT-9 end plugs, and MA754 and HT-9 coupons. Limited burst testing of MA957/HT-9 joints carried out at various pressures up to 400oC has shown encouraging results in that the joint regions do not develop any cracking. Similar joint strength observations have also been made by performing simple bend tests. Detailed microstructural studies using SEM/EBSD tools and fatigue crack growth studies of MA754/HT-9 joints are ongoing.

  6. Secondary calcium solid electrolyte high temperature battery

    SciTech Connect (OSTI)

    Sammells, A.F.; Schumacher, B.

    1986-01-01T23:59:59.000Z

    The authors report on recent work directed towards determining the viability of polycrystalline Ca/sup 2 +/ conducting ..beta..''-alumina solid electrolytes as the basis for a new type of high temperature battery. In this battery system the negative electrode consisted of a calcium-silicon alloy whose redox electro-chemistry was mediated to the calcium conducting solid electrolyte via the use of the molten salt eutectic CaCl/sub 2/ (51.4/sup M//0), CaI/sub 2/ (mp 550/sup 0/C). Both the molten salt and the calcium-alloy negative active material were separated from the positive active material via the Ca/sup 2 +/ conducting polycrystalline solid electrolyte. The positive electrode consisted of a solid-state matrix having a somewhat related crystallographic structure to Ca/sup 2 +/ ..beta..''-alumina, but where a significant fraction of the A1/sup 3 +/ sites located within this solid electrolyte's spinel block were replaced by immobile transition metal species. These species were available for participating in solid-state redox electrochemistry upon electrochemical cell cycling.

  7. Development of Strengthened Bundle High Temperature Superconductors

    SciTech Connect (OSTI)

    Lue, J.W.; Lubell, M.S. [Oak Ridge National Lab., TN (United States); Demko, J.A. [Oak Ridge Inst. for Science and Education, TN (United States); Tomsic, M. [Plastronic, Inc., Troy, OH (United States); Sinha, U. [Southwire Company, Carollton, GA (United States)

    1997-12-31T23:59:59.000Z

    In the process of developing high temperature superconducting (HTS) transmission cables, it was found that mechanical strength of the superconducting tape is the most crucial property that needs to be improved. It is also desirable to increase the current carrying capacity of the conductor so that fewer layers are needed to make the kilo-amp class cables required for electric utility usage. A process has been developed by encapsulating a stack of Bi-2223/Ag tapes with a silver or non-silver sheath to form a strengthened bundle superconductor. This process was applied to HTS tapes made by the Continuous Tube Forming and Filling (CTFF) technique pursued by Plastronic Inc. and HTS tapes obtained from other manufacturers. Conductors with a bundle of 2 to 6 HTS tapes have been made. The bundled conductor is greatly strengthened by the non-silver sheath. No superconductor degradation as compared to the sum of the original critical currents of the individual tapes was seen on the finished conductors.

  8. Electronic high voltage generator with a high temperature superconducting coil

    SciTech Connect (OSTI)

    Jin, J.X.; Liu, H.K.; Dou, S.X. [Univ. of Wollongong (Australia)] [and others

    1996-12-31T23:59:59.000Z

    A novel method for generating high voltages from a low voltage DC source, by using a capacitor and inductor in a R, L, C resonant circuit has been developed with the consideration of using a high temperature superconducting (HTS) coil. To generate high voltages the polarity of a low voltage battery source is reversed each half resonant cycle, the control being achieved by an electronic switch. Resistance in the circuit limits the voltages that can be built up. By replacing a copper winding inductor with another inductor which has a HTS winding, the magnitude of achievable voltages is substantially increased. A (Bi,Pb){sub 2}Sr{sub 2}Ca{sub 2}Cu{sub 3}O{sub 10+x} multifilament HTS wire is considered in this work to make the superconducting inductor. The high voltages generated are not capable of supplying low impedance loads, however, possible applications of the generator include electrical partial discharge testing and insulation resistance testing. It could also be used as a testing method for the HTS itself with respect to the critical current and AC loss measurement.

  9. High-flux magnetorheology at elevated temperatures

    E-Print Network [OSTI]

    Ocalan, Murat

    Commercial applications of magnetorheological (MR) fluids often require operation at elevated temperatures as a result of surrounding environmental conditions or intense localized viscous heating. Previous experimental ...

  10. High Temperature 300C Directional Drilling System

    Broader source: Energy.gov [DOE]

    Project objective: provide a directional drilling system that can be used at environmental temperatures of up to 300C; and at depths of 10; 000 meters.

  11. High-Temperature Falling-Particle Receiver

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

    temperatures, nitrate salt fluids become chemically unstable. In contrast, direct absorption receivers using solid particles that fall through a beam of concentrated solar...

  12. STATUS OF HIGH FLUX ISOTOPE REACTOR IRRADIATION OF SILICON CARBIDE/SILICON CARBIDE JOINTS

    SciTech Connect (OSTI)

    Katoh, Yutai [ORNL; Koyanagi, Takaaki [ORNL; Kiggans, Jim [ORNL; Cetiner, Nesrin [ORNL; McDuffee, Joel [ORNL

    2014-09-01T23:59:59.000Z

    Development of silicon carbide (SiC) joints that retain adequate structural and functional properties in the anticipated service conditions is a critical milestone toward establishment of advanced SiC composite technology for the accident-tolerant light water reactor (LWR) fuels and core structures. Neutron irradiation is among the most critical factors that define the harsh service condition of LWR fuel during the normal operation. The overarching goal of the present joining and irradiation studies is to establish technologies for joining SiC-based materials for use as the LWR fuel cladding. The purpose of this work is to fabricate SiC joint specimens, characterize those joints in an unirradiated condition, and prepare rabbit capsules for neutron irradiation study on the fabricated specimens in the High Flux Isotope Reactor (HFIR). Torsional shear test specimens of chemically vapor-deposited SiC were prepared by seven different joining methods either at Oak Ridge National Laboratory or by industrial partners. The joint test specimens were characterized for shear strength and microstructures in an unirradiated condition. Rabbit irradiation capsules were designed and fabricated for neutron irradiation of these joint specimens at an LWR-relevant temperature. These rabbit capsules, already started irradiation in HFIR, are scheduled to complete irradiation to an LWR-relevant dose level in early 2015.

  13. High flux isotope reactor cold source preconceptual design study report

    SciTech Connect (OSTI)

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

    1995-12-01T23:59:59.000Z

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

  14. High Flux Isotope Reactor cold neutron source reference design concept

    SciTech Connect (OSTI)

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

    1998-05-01T23:59:59.000Z

    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.

  15. Level 1 Tornado PRA for the High Flux Beam Reactor

    SciTech Connect (OSTI)

    Bozoki, G.E.; Conrad, C.S.

    1994-05-01T23:59:59.000Z

    This report describes a risk analysis primarily directed at providing an estimate for the frequency of tornado induced damage to the core of the High Flux Beam Reactor (HFBR), and thus it constitutes a Level 1 Probabilistic Risk Assessment (PRA) covering tornado induced accident sequences. The basic methodology of the risk analysis was to develop a ``tornado specific`` plant logic model that integrates the internal random hardware failures with failures caused externally by the tornado strike and includes operator errors worsened by the tornado modified environment. The tornado hazard frequency, as well as earlier prepared structural and equipment fragility data, were used as input data to the model. To keep modeling/calculational complexity as simple as reasonable a ``bounding`` type, slightly conservative, approach was applied. By a thorough screening process a single dominant initiating event was selected as a representative initiator, defined as: ``Tornado Induced Loss of Offsite Power.`` The frequency of this initiator was determined to be 6.37E-5/year. The safety response of the HFBR facility resulted in a total Conditional Core Damage Probability of .621. Thus, the point estimate of the HFBR`s Tornado Induced Core Damage Frequency (CDF) was found to be: (CDF){sub Tornado} = 3.96E-5/year. This value represents only 7.8% of the internal CDF and thus is considered to be a small contribution to the overall facility risk expressed in terms of total Core Damage Frequency. In addition to providing the estimate of (CDF){sub Tornado}, the report documents, the relative importance of various tornado induced system, component, and operator failures that contribute most to (CDF){sub Tornado}.

  16. NOvel Refractory Materials for High Alkali, High Temperature Environments

    SciTech Connect (OSTI)

    Hemrick, J.G.; Griffin, R. (MINTEQ International, Inc.)

    2011-08-30T23:59:59.000Z

    Refractory materials can be limited in their application by many factors including chemical reactions between the service environment and the refractory material, mechanical degradation of the refractory material by the service environment, temperature limitations on the use of a particular refractory material, and the inability to install or repair the refractory material in a cost effective manner or while the vessel was in service. The objective of this project was to address the need for new innovative refractory compositions by developing a family of novel MgO-Al2O3 spinel or other similar magnesia/alumina containing unshaped refractory composition (castables, gunnables, shotcretes, etc) utilizing new aggregate materials, bond systems, protective coatings, and phase formation techniques (in-situ phase formation, altered conversion temperatures, accelerated reactions, etc). This family of refractory compositions would then be tailored for use in high-temperature, highalkaline industrial environments like those found in the aluminum, chemical, forest products, glass, and steel industries. A research team was formed to carry out the proposed work led by Oak Ridge National Laboratory (ORNL) and was comprised of the academic institution Missouri University of Science and Technology (MS&T), and the industrial company MINTEQ International, Inc. (MINTEQ), along with representatives from the aluminum, chemical, glass, and forest products industries. The two goals of this project were to produce novel refractory compositions which will allow for improved energy efficiency and to develop new refractory application techniques which would improve the speed of installation. Also methods of hot installation were sought which would allow for hot repairs and on-line maintenance leading to reduced process downtimes and eliminating the need to cool and reheat process vessels.

  17. Calcite Mineral Scaling Potentials of High-Temperature Geothermal Wells

    E-Print Network [OSTI]

    Karlsson, Brynjar

    #12;i Calcite Mineral Scaling Potentials of High-Temperature Geothermal Wells Alvin I. Remoroza-Temperature Geothermal Wells Alvin I. Remoroza 60 ECTS thesis submitted in partial fulfillment of a Magister Scientiarum #12;iv Calcite Mineral Scaling Potentials of High-Temperature Geothermal Wells 60 ECTS thesis

  18. Vibrational Raman Spectroscopy of High-temperature Superconductors

    E-Print Network [OSTI]

    Nabben, Reinhard

    Vibrational Raman Spectroscopy of High-temperature Superconductors C. Thomsen and G. Kaczmarczyk-temperature Superconductors C. Thomsen and G. Kaczmarczyk Technical University of Berlin, Berlin, Germany 1 INTRODUCTION Raman after the discovery of high- critical-temperature Tc superconductors:2 while reports on Raman scattering

  19. High-temperature piezoresponse force microscopy B. Bhatia,1

    E-Print Network [OSTI]

    King, William P.

    High-temperature piezoresponse force microscopy B. Bhatia,1 J. Karthik,2 D. G. Cahill,1,2 L. W September 2011; published online 24 October 2011) We report high temperature piezoresponse force microscopy resistive heater allows local temperature control up to 1000 C with minimal electrostatic interactions

  20. Quark number susceptibility of high temperature and finite density QCD

    E-Print Network [OSTI]

    Ari Hietanen; Kari Rummukainen

    2007-10-26T23:59:59.000Z

    We utilize lattice simulations of the dimensionally reduced effective field theory (EQCD) to determine the quark number susceptibility of QCD at high temperature ($T>2T_c$). We also use analytic continuation to obtain results at finite density. The results extrapolate well from known perturbative expansion (accurate in extremely high temperatures) to 4d lower temperature lattice data

  1. High temperature pressurized high frequency testing rig and test method

    DOE Patents [OSTI]

    De La Cruz, Jose; Lacey, Paul

    2003-04-15T23:59:59.000Z

    An apparatus is described which permits the lubricity of fuel compositions at or near temperatures and pressures experienced by compression ignition fuel injector components during operation in a running engine. The apparatus consists of means to apply a measured force between two surfaces and oscillate them at high frequency while wetted with a sample of the fuel composition heated to an operator selected temperature. Provision is made to permit operation at or near the flash point of the fuel compositions. Additionally a method of using the subject apparatus to simulate ASTM Testing Method D6079 is disclosed, said method involving using the disclosed apparatus to contact the faces of prepared workpieces under a measured load, sealing the workface contact point into the disclosed apparatus while immersing said contact point between said workfaces in a lubricating media to be tested, pressurizing and heating the chamber and thereby the fluid and workfaces therewithin, using the disclosed apparatus to impart a differential linear motion between the workpieces at their contact point until a measurable scar is imparted to at least one workpiece workface, and then evaluating the workface scar.

  2. Design of High Field Solenoids made of High Temperature Superconductors

    SciTech Connect (OSTI)

    Bartalesi, Antonio; /Pisa U.

    2010-12-01T23:59:59.000Z

    This thesis starts from the analytical mechanical analysis of a superconducting solenoid, loaded by self generated Lorentz forces. Also, a finite element model is proposed and verified with the analytical results. To study the anisotropic behavior of a coil made by layers of superconductor and insulation, a finite element meso-mechanic model is proposed and designed. The resulting material properties are then used in the main solenoid analysis. In parallel, design work is performed as well: an existing Insert Test Facility (ITF) is adapted and structurally verified to support a coil made of YBa{sub 2}Cu{sub 3}O{sub 7}, a High Temperature Superconductor (HTS). Finally, a technological winding process was proposed and the required tooling is designed.

  3. High temperature behavior of metallic inclusions in uranium dioxide

    SciTech Connect (OSTI)

    Yang, R.L.

    1980-08-01T23:59:59.000Z

    The object of this thesis was to construct a temperature gradient furnace to simulate the thermal conditions in the reactor fuel and to study the migration of metallic inclusions in uranium oxide under the influence of temperature gradient. No thermal migration of molybdenum and tungsten inclusions was observed under the experimental conditions. Ruthenium inclusions, however, dissolved and diffused atomically through grain boundaries in slightly reduced uranium oxide. An intermetallic compound (probably URu/sub 3/) was formed by reaction of Ru and UO/sub 2-x/. The diffusivity and solubility of ruthenium in uranium oxide were measured.

  4. Expansion Joint Concepts for High Temperature Insulation Systems

    E-Print Network [OSTI]

    Harrison, M. R.

    1980-01-01T23:59:59.000Z

    As high temperature steam and process piping expands with heat, joints begin to open between the insulation sections, resulting in increased energy loss and possible unsafe surface temperatures. Many different expansion joint designs are presently...

  5. Ultra-High Temperature Distributed Wireless Sensors

    SciTech Connect (OSTI)

    May, Russell; Rumpf, Raymond; Coggin, John; Davis, Williams; Yang, Taeyoung; O'Donnell, Alan; Bresnahan, Peter

    2013-03-31T23:59:59.000Z

    Research was conducted towards the development of a passive wireless sensor for measurement of temperature in coal gasifiers and coal-fired boiler plants. Approaches investigated included metamaterial sensors based on guided mode resonance filters, and temperature-sensitive antennas that modulate the frequency of incident radio waves as they are re-radiated by the antenna. In the guided mode resonant filter metamaterial approach, temperature is encoded as changes in the sharpness of the filter response, which changes with temperature because the dielectric loss of the guided mode resonance filter is temperature-dependent. In the mechanically modulated antenna approach, the resonant frequency of a vibrating cantilever beam attached to the antenna changes with temperature. The vibration of the beam perturbs the electrical impedance of the antenna, so that incident radio waves are phase modulated at a frequency equal to the resonant frequency of the vibrating beam. Since the beam resonant frequency depends on temperature, a Doppler radar can be used to remotely measure the temperature of the antenna. Laboratory testing of the guided mode resonance filter failed to produce the spectral response predicted by simulations. It was concluded that the spectral response was dominated by spectral reflections of radio waves incident on the filter. Laboratory testing of the mechanically modulated antenna demonstrated that the device frequency shifted incident radio waves, and that the frequency of the re-radiated waves varied linearly with temperature. Radio wave propagation tests in the convection pass of a small research boiler plant identified a spectral window between 10 and 13 GHz for low loss propagation of radio waves in the interior of the boiler.

  6. High temperature, minimally invasive optical sensing modules

    DOE Patents [OSTI]

    Riza, Nabeel Agha (Oviedo, FL); Perez, Frank (Tujunga, CA)

    2008-02-05T23:59:59.000Z

    A remote temperature sensing system includes a light source selectively producing light at two different wavelengths and a sensor device having an optical path length that varies as a function of temperature. The sensor receives light emitted by the light source and redirects the light along the optical path length. The system also includes a detector receiving redirected light from the sensor device and generating respective signals indicative of respective intensities of received redirected light corresponding to respective wavelengths of light emitted by the light source. The system also includes a processor processing the signals generated by the detector to calculate a temperature of the device.

  7. Design strategies for optimizing high burnup fuel in pressurized water reactors

    E-Print Network [OSTI]

    Xu, Zhiwen, 1975-

    2003-01-01T23:59:59.000Z

    This work is focused on the strategy for utilizing high-burnup fuel in pressurized water reactors (PWR) with special emphasis on the full array of neutronic considerations. The historical increase in batch-averaged discharge ...

  8. A summary of high-temperature electronics research and development

    SciTech Connect (OSTI)

    Thome, F.V.; King, D.B.

    1991-10-18T23:59:59.000Z

    Current and future needs in automative, aircraft, space, military, and well logging industries require operation of electronics at higher temperatures than today's accepted limit of 395 K. Without the availability of high-temperature electronics, many systems must operate under derated conditions or must accept severe mass penalties required by coolant systems to maintain electronic temperatures below critical levels. This paper presents ongoing research and development in the electronics community to bring high-temperature electronics to commercial realization. Much of this work was recently reviewed at the First International High-Temperature Electronics Conference held 16--20 June 1991 in Albuquerque, New Mexico. 4 refs., 1 tab.

  9. Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

    SciTech Connect (OSTI)

    Heeger, Karsten M [Yale University

    2014-09-13T23:59:59.000Z

    This reports presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zero$\\theta_{13}$. Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.

  10. High Temperature, High Pressure Devices for Zonal Isolation in Geothermal Wells

    Broader source: Energy.gov [DOE]

    DOE Geothermal Peer Review 2010 - Presentation. Project objectives: Design, demonstrate, and qualify high-temperature high pressure zonal isolation devices compatible with the high temperature downhole Enhanced Geothermal Systems (EGS) environment.

  11. High Flux Isotopes Reactor (HFIR) Cooling Towers Demolition Waste Management

    SciTech Connect (OSTI)

    Pudelek, R. E.; Gilbert, W. C.

    2002-02-26T23:59:59.000Z

    This paper describes the results of a joint initiative between Oak Ridge National Laboratory, operated by UT-Battelle, and Bechtel Jacobs Company, LLC (BJC) to characterize, package, transport, treat, and dispose of demolition waste from the High Flux Isotope Reactor (HFIR), Cooling Tower. The demolition and removal of waste from the site was the first critical step in the planned HFIR beryllium reflector replacement outage scheduled. The outage was scheduled to last a maximum of six months. Demolition and removal of the waste was critical because a new tower was to be constructed over the old concrete water basin. A detailed sampling and analysis plan was developed to characterize the hazardous and radiological constituents of the components of the Cooling Tower. Analyses were performed for Resource Conservation and Recovery Act (RCRA) heavy metals and semi-volatile constituents as defined by 40 CFR 261 and radiological parameters including gross alpha, gross beta, gross gamma, alpha-emitting isotopes and beta-emitting isotopes. Analysis of metals and semi-volatile constituents indicated no exceedances of regulatory limits. Analysis of radionuclides identified uranium and thorium and associated daughters. In addition 60Co, 99Tc, 226Rm, and 228Rm were identified. Most of the tower materials were determined to be low level radioactive waste. A small quantity was determined not to be radioactive, or could be decontaminated. The tower was dismantled October 2000 to January 2001 using a detailed step-by-step process to aid waste segregation and container loading. The volume of waste as packaged for treatment was approximately 1982 cubic meters (70,000 cubic feet). This volume was comprised of plastic ({approx}47%), wood ({approx}38%) and asbestos transite ({approx}14%). The remaining {approx}1% consisted of the fire protection piping (contaminated with lead-based paint) and incidental metal from conduit, nails and braces/supports, and sludge from the basin. The waste, except for the asbestos, was volume reduced via a private contract mechanism established by BJC. After volume reduction, the waste was packaged for rail shipment. This large waste management project successfully met cost and schedule goals.

  12. First high-temperature electronics products survey 2005.

    SciTech Connect (OSTI)

    Normann, Randy Allen

    2006-04-01T23:59:59.000Z

    On April 4-5, 2005, a High-Temperature Electronics Products Workshop was held. This workshop engaged a number of governmental and private industry organizations sharing a common interest in the development of commercially available, high-temperature electronics. One of the outcomes of this meeting was an agreement to conduct an industry survey of high-temperature applications. This report covers the basic results of this survey.

  13. Overview of Fraunhofer IPM Activities in High Temperature Bulk...

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

    Workshop including an overview about Fraunhofer IPM, new funding situation in Germany, high temperature material and modules, energy-autarkic sensors, and thermoelectric...

  14. Detecting Fractures Using Technology at High Temperatures and...

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

    Detecting Fractures Using Technology at High Temperatures and Depths - Geothermal Ultrasonic Fracture Imager (GUFI); 2010 Geothermal Technology Program Peer Review Report Detecting...

  15. Microchannel High-Temperature Recuperator for Fuel Cell Systems...

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

    developed an efficient, microchannel-based waste heat recuperator for a high-temperature fuel cell system. This technology increases the efficiency of fuel cells and improves...

  16. Detecting Fractures Using Technology at High Temperatures and...

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

    7 4.4.1 Detecting Fractures Using Technology at High Temperatures and Depths - Geothermal Ultrasonic Fracture Imager (GUFI) Presentation Number: 015 Investigator: Patterson, Doug...

  17. advanced high temperature: Topics by E-print Network

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

    John 3 Lake surface water temperature retrieval using advanced very high resolution radiometer and Geosciences Websites Summary: and Moderate Resolution Imaging Spectroradiometer...

  18. Project Profile: Fundamental Corrosion Studies in High-Temperature...

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

    Studies in High-Temperature Molten Salt Systems for Next-Generation CSP Systems Savannah River National Laboratory logo The Savannah River National Laboratory (SRNL), under the...

  19. Enhanced High and Low Temperature Performance of NOx Reduction...

    Energy Savers [EERE]

    High and Low Temperature Performance of NOx Reduction Materials 2013 DOE Hydrogen and Fuel Cells Program and Vehicle Technologies Program Annual Merit Review and Peer...

  20. Nanostructured High-Temperature Bulk Thermoelectric Energy Conversion...

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

    Nanostructured High-Temperature Bulk Thermoelectric Energy Conversion for Efficient Automotive Waste Heat Recovery Vehicle Technologies Office Merit Review 2014: Nanostructured...

  1. Feasibility and Design Studies for a High Temperature Downhole Tool

    Broader source: Energy.gov [DOE]

    Project objective: Perform feasibility and design studies for a high temperature downhole tool; which uses nuclear techniques for characterization purposes; using measurements and modeling/simulation.

  2. Enhanced High Temperature Performance of NOx Storage/Reduction...

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

    Program Annual Merit Review and Peer Evaluation ace026peden2011o.pdf More Documents & Publications Enhanced High Temperature Performance of NOx StorageReduction (NSR) Materials...

  3. Combining Raman Microprobe and XPS to Study High Temperature...

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

    spectroscopy. Citation: Windisch CF, Jr, CH Henager, MH Engelhard, and WD Bennett.2011."Combining Raman Microprobe and XPS to Study High Temperature Oxidation of...

  4. Development of a 500 Watt High Temperature Thermoelectric Generator...

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

    Publications Development of a 100-Watt High Temperature Thermoelectric Generator Automotive Waste Heat Conversion to Power Program Automotive Waste Heat Conversion to Power Program...

  5. Metallic substrates for high temperature superconductors

    DOE Patents [OSTI]

    Truchan, Thomas G. (Chicago, IL); Miller, Dean J. (Darien, IL); Goretta, Kenneth C. (Downers Grove, IL); Balachandran, Uthamalingam (Hinsdale, IL); Foley, Robert (Chicago, IL)

    2002-01-01T23:59:59.000Z

    A biaxially textured face-centered cubic metal article having grain boundaries with misorientation angles greater than about 8.degree. limited to less than about 1%. A laminate article is also disclosed having a metal substrate first rolled to at least about 95% thickness reduction followed by a first annealing at a temperature less than about 375.degree. C. Then a second rolling operation of not greater than about 6% thickness reduction is provided, followed by a second annealing at a temperature greater than about 400.degree. C. A method of forming the metal and laminate articles is also disclosed.

  6. Effective theory of high-temperature superconductors

    E-Print Network [OSTI]

    Igor F. Herbut

    2005-06-16T23:59:59.000Z

    General field theory of a fluctuating d-wave superconductor is constructed and proposed as an effective description of superconducting cuprates at low energies. The theory is used to resolve a puzzle posed by recent experiments on superfluid density in severely underdoped YBCO. In particular, the overall temperature dependence of the superfluid density at low dopings is argued to be described well by the strongly anisotropic weakly interacting three-dimensional Bose gas, and thus approximately linear in temperature with an almost doping-independent slope.

  7. Quantum gravitational proton decay at high temperature

    E-Print Network [OSTI]

    Ulf H. Danielsson

    2005-12-29T23:59:59.000Z

    One of the most important challenges of contemporary physics is to find experimental signatures of quantum gravity. It is expected that quantum gravitational effects lead to proton decay but on time scales way beyond what is of any relevance to experiments. At non-zero temperatures there are reasons to believe that the situation is much more favourable. We will argue that at the temperatures and densities reached at present and future fusion facilities there is a realistic possibility that proton decay could be detectable.

  8. Tokamaks with high-performance resistive magnets: advanced test reactors and prospects for commercial applications

    SciTech Connect (OSTI)

    Bromberg, L.; Cohn, D.R.; Williams, J.E.C.; Becker, H.; Leclaire, R.; Yang, T.

    1981-10-01T23:59:59.000Z

    Scoping studies have been made of tokamak reactors with high performance resistive magnets which maximize advantages gained from high field operation and reduced shielding requirements, and minimize resistive power requirements. High field operation can provide very high values of fusion power density and n tau/sub e/ while the resistive power losses can be kept relatively small. Relatively high values of Q' = Fusion Power/Magnet Resistive Power can be obtained. The use of high field also facilitates operation in the DD-DT advanced fuel mode. The general engineering and operational features of machines with high performance magnets are discussed. Illustrative parameters are given for advanced test reactors and for possible commercial reactors. Commercial applications that are discussed are the production of fissile fuel, electricity generation with and without fissioning blankets and synthetic fuel production.

  9. Hydrogen production by high-temperature steam gasification of biomass and coal

    SciTech Connect (OSTI)

    Kriengsak, S.N.; Buczynski, R.; Gmurczyk, J.; Gupta, A.K. [University of Maryland, College Park, MD (United States). Dept. of Mechanical Engineering

    2009-04-15T23:59:59.000Z

    High-temperature steam gasification of paper, yellow pine woodchips, and Pittsburgh bituminous coal was investigated in a batch-type flow reactor at temperatures in the range of 700 to 1,200{sup o}C at two different ratios of steam to feedstock molar ratios. Hydrogen yield of 54.7% for paper, 60.2% for woodchips, and 57.8% for coal was achieved on a dry basis, with a steam flow rate of 6.3 g/min at steam temperature of 1,200{sup o}C. Yield of both the hydrogen and carbon monoxide increased while carbon dioxide and methane decreased with the increase in gasification temperature. A 10-fold reduction in tar residue was obtained at high-temperature steam gasification, compared to low temperatures. Steam and gasification temperature affects the composition of the syngas produced. Higher steam-to-feedstock molar ratio had negligible effect on the amount of hydrogen produced in the syngas in the fixed-batch type of reactor. Gasification temperature can be used to control the amounts of hydrogen or methane produced from the gasification process. This also provides mean to control the ratio of hydrogen to CO in the syngas, which can then be processed to produce liquid hydrocarbon fuel since the liquid fuel production requires an optimum ratio between hydrogen and CO. The syngas produced can be further processed to produce pure hydrogen. Biomass fuels are good source of renewable fuels to produce hydrogen or liquid fuels using controlled steam gasification.

  10. High-Temperature Viscosity of Commercial Glasses

    SciTech Connect (OSTI)

    Hrma, Pavel R.

    2006-08-31T23:59:59.000Z

    Arrhenius models were developed for glass viscosity within the processing temperature of six types of commercial glasses: low-expansion-borosilicate glasses, E glasses, fiberglass wool glasses, TV panel glasses, container glasses, and float glasses. Both local models (for each of the six glass types) and a global model (for the composition region of commercial glasses, i.e., the six glass types taken together) are presented. The models are based on viscosity data previously obtained with rotating spindle viscometers within the temperature range between 900 C and 1550 C; the viscosity varied from 1 Pa?s to 750 Pa?s. First-order models were applied to relate Arrhenius coefficients to the mass fractions of 15 components: SiO2, TiO2, ZrO2, Al2O3, Fe2O3, B2O3, MgO, CaO, SrO, BaO, PbO, ZnO, Li2O, Na2O, K2O. The R2 is 0.98 for the global model and ranges from .097 to 0.99 for the six local models. The models are recommended for glasses containing 42 to 84 mass% SiO2 to estimate viscosities or temperatures at a constant viscosity for melts within both the temperature range from 1100 C to 1550 C and viscosity range from 5 to 400 Pa?s.

  11. Rotational viscometer for high-pressure high-temperature fluids

    DOE Patents [OSTI]

    Carr, Kenneth R. (Knoxville, TN)

    1985-01-01T23:59:59.000Z

    The invention is a novel rotational viscometer which is well adapted for use with fluids at high temperatures and/or pressures. In one embodiment, the viscometer includes a substantially non-magnetic tube having a closed end and having an open end in communication with a fluid whose viscosity is to be determined. An annular drive magnet is mounted for rotation about the tube. The tube encompasses and supports a rotatable shaft assembly which carries a rotor, or bob, for insertion in the fluid. Affixed to the shaft are (a) a second magnet which is magnetically coupled to the drive magnet and (b) a third magnet. In a typical operation, the drive magnet is rotated to turn the shaft assembly while the shaft rotor is immersed in the fluid. The viscous drag on the rotor causes the shaft assembly to lag the rotation of the drive magnet by an amount which is a function of the amount of viscous drag. A first magnetic pickup generates a waveform whose phase is a function of the angular position of the drive magnet. A second magnetic pickup generates a waveform whose phase is a function of the angular position of the third magnet. An output is generated indicative of the phase difference between the two waveforms.

  12. System Evaluation and Economic Analysis of a HTGR Powered High-Temperature Electrolysis Hydrogen Production Plant

    SciTech Connect (OSTI)

    Michael G. McKellar; Edwin A. Harvego; Anastasia A. Gandrik

    2010-10-01T23:59:59.000Z

    A design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322C and 750C, respectively. The power conversion unit will be a Rankine steam cycle with a power conversion efficiency of 40%. The reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 40.4% at a hydrogen production rate of 1.75 kg/s and an oxygen production rate of 13.8 kg/s. An economic analysis of this plant was performed with realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a cost of $3.67/kg of hydrogen assuming an internal rate of return, IRR, of 12% and a debt to equity ratio of 80%/20%. A second analysis shows that if the power cycle efficiency increases to 44.4%, the hydrogen production efficiency increases to 42.8% and the hydrogen and oxygen production rates are 1.85 kg/s and 14.6 kg/s respectively. At the higher power cycle efficiency and an IRR of 12% the cost of hydrogen production is $3.50/kg.

  13. 3D CFD Model of High Temperature H2O/CO2 Co-electrolysis

    SciTech Connect (OSTI)

    Grant Hawkes; James O'Brien; Carl Stoots; Stephen Herring; Joe Hartvigsen

    2007-06-01T23:59:59.000Z

    3D CFD Model of High Temperature H2O/CO2 Co-Electrolysis Grant Hawkes1, James OBrien1, Carl Stoots1, Stephen Herring1 Joe Hartvigsen2 1 Idaho National Laboratory, Idaho Falls, Idaho, grant.hawkes@inl.gov 2 Ceramatec Inc, Salt Lake City, Utah INTRODUCTION A three-dimensional computational fluid dynamics (CFD) model has been created to model high temperature co-electrolysis of steam and carbon dioxide in a planar solid oxide electrolyzer (SOE) using solid oxide fuel cell technology. A research program is under way at the Idaho National Laboratory (INL) to simultaneously address the research and scale-up issues associated with the implementation of planar solid-oxide electrolysis cell technology for syn-gas production from CO2 and steam. Various runs have been performed under different run conditions to help assess the performance of the SOE. This paper presents CFD results of this model compared with experimental results. The Idaho National Laboratory (INL), in conjunction with Ceramatec Inc. (Salt Lake City, USA) has been researching for several years the use of solid-oxide fuel cell technology to electrolyze steam for large-scale nuclear-powered hydrogen production. Now, an experimental research project is underway at the INL to produce syngas by simultaneously electrolyzing at high-temperature steam and carbon dioxide (CO2) using solid oxide fuel cell technology. A strong interest exists in the large-scale production of syn-gas from CO2 and steam to be reformed into a usable transportation fuel. If biomass is used as the carbon source, the overall process is climate neutral. Consequently, there is a high level of interest in production of syn-gas from CO2 and steam electrolysis. With the price of oil currently around $60 / barrel, synthetically-derived hydrocarbon fuels (synfuels) have become economical. Synfuels are typically produced from syngas hydrogen (H2) and carbon monoxide (CO) -- using the Fischer-Tropsch process, discovered by Germany before World War II. High-temperature nuclear reactors have the potential for substantially increasing the efficiency of syn-gas production from CO2 and water, with no consumption of fossil fuels, and no production of greenhouse gases. Thermal CO2-splitting and water splitting for syn-gas production can be accomplished via high-temperature electrolysis, using high-temperature nuclear process heat and electricity. A high-temperature advanced nuclear reactor coupled with a high-efficiency high-temperature electrolyzer could achieve a competitive thermal-to-syn-gas conversion efficiency of 45 to 55%.

  14. High conduction neutron absorber to simulate fast reactor environment in an existing test reactor

    SciTech Connect (OSTI)

    Donna Post Guillen; Larry R. Greenwood; James R. Parry

    2014-06-22T23:59:59.000Z

    A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three binsthermal, epithermal, and fastto evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluence monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.

  15. High-Temperature-High-Volume Lifting for Enhanced Geothermal Systems

    Broader source: Energy.gov [DOE]

    Project objective: Advance the technology for well fluids lifting systems to meet the foreseeable pressure; temperature; and longevity needs of the Enhanced Geothermal Systems (EGS) industry.

  16. HIGH-TEMPERATURE ELECTROLYSIS FOR LARGE-SCALE HYDROGEN AND SYNGAS PRODUCTION FROM NUCLEAR ENERGY SYSTEM SIMULATION AND ECONOMICS

    SciTech Connect (OSTI)

    J. E. O'Brien; M. G. McKellar; E. A. Harvego; C. M. Stoots

    2009-05-01T23:59:59.000Z

    A research and development program is under way at the Idaho National Laboratory (INL) to assess the technological and scale-up issues associated with the implementation of solid-oxide electrolysis cell technology for efficient high-temperature hydrogen production from steam. This work is supported by the US Department of Energy, Office of Nuclear Energy, under the Nuclear Hydrogen Initiative. This paper will provide an overview of large-scale system modeling results and economic analyses that have been completed to date. System analysis results have been obtained using the commercial code UniSim, augmented with a custom high-temperature electrolyzer module. Economic analysis results were based on the DOE H2A analysis methodology. The process flow diagrams for the system simulations include an advanced nuclear reactor as a source of high-temperature process heat, a power cycle and a coupled steam electrolysis loop. Several reactor types and power cycles have been considered, over a range of reactor outlet temperatures. Pure steam electrolysis for hydrogen production as well as coelectrolysis for syngas production from steam/carbon dioxide mixtures have both been considered. In addition, the feasibility of coupling the high-temperature electrolysis process to biomass and coal-based synthetic fuels production has been considered. These simulations demonstrate that the addition of supplementary nuclear hydrogen to synthetic fuels production from any carbon source minimizes emissions of carbon dioxide during the production process.

  17. High temperature expanding cement composition and use

    DOE Patents [OSTI]

    Nelson, Erik B. (Tulsa County, OK); Eilers, Louis H. (Rogers County, OK)

    1982-01-01T23:59:59.000Z

    A hydratable cement composition useful for preparing a pectolite-containing expanding cement at temperatures above about 150.degree. C. comprising a water soluble sodium salt of a weak acid, a 0.1 molar aqueous solution of which salt has a pH of between about 7.5 and about 11.5, a calcium source, and a silicon source, where the atomic ratio of sodium to calcium to silicon ranges from about 0.3:0.6:1 to about 0.03:1:1; aqueous slurries prepared therefrom and the use of such slurries for plugging subterranean cavities at a temperature of at least about 150.degree. C. The invention composition is useful for preparing a pectolite-containing expansive cement having about 0.2 to about 2 percent expansion, by volume, when cured at at least 150.degree. C.

  18. High Temperature Materials for Aerospace Applications

    E-Print Network [OSTI]

    Adamczak, Andrea Diane

    2011-08-08T23:59:59.000Z

    Chair of Advisory Committee: Dr. Jaime C. Grunlan Further crosslinking of the fluorinated polyimide was examined to separate the cure reactions from degradation and to determine the optimum post curing conditions. Glass transition... ranging from 225 ? 362 ?C, with 1.7 - 3.0 wt% absorbed moisture, and the polyimide composite had blister temperatures from 246 ? 294 ?C with 0.5 - 1.5 wt% moisture. iv Weight loss of the fluorinated polyimide and its corresponding polyimide carbon...

  19. High-Temperature Viscosity Of Commercial Glasses

    SciTech Connect (OSTI)

    Hrma, Pavel R.; See, Clem A.; Lam, Oanh P.; Minister, Kevin B.

    2005-01-01T23:59:59.000Z

    Viscosity was measured for six types of commercial glasses: low-expansion-borosilicate glasses, E glasses, fiberglass wool glasses, TV panel glasses, container glasses, and float glasses. Viscosity data were obtained with rotating spindle viscometers within the temperature range between 900C and 1550C; the viscosity varied from 1 Pa?s to 750 Pa?s. Arrhenius coefficients were calculated for individual glasses and linear models were applied to relate them to the mass fractions of 11 major components (SiO2, CaO, Na2O, Al2O3, B2O3, BaO, SrO, K2O, MgO, PbO, and ZrO2) and 12 minor components (Fe2O3, ZnO, Li2O, TiO2, CeO2, F, Sb2O3, Cr2O3, As2O3, MnO2, SO3, and Co3O4). The models are recommended for glasses containing 42 to 84 mass% SiO2 to estimate viscosities or temperatures at a constant viscosity for melts within both the temperature range from 1100C to 1550C and viscosity range from 10 to 400 Pa?s.

  20. To Crack or Not to Crack: Strain in High Temperature Superconductors

    E-Print Network [OSTI]

    Godeke, Arno

    2008-01-01T23:59:59.000Z

    in High Temperature Superconductors Arno Godeke August 22,in High Temperature Superconductors Motivation Magneticin High Temperature Superconductors How do Nb 3 Sn magnets

  1. Tar-free fuel gas production from high temperature pyrolysis of sewage sludge

    SciTech Connect (OSTI)

    Zhang, Leguan; Xiao, Bo; Hu, Zhiquan; Liu, Shiming, E-mail: Zhangping101@yeah.net; Cheng, Gong; He, Piwen; Sun, Lei

    2014-01-15T23:59:59.000Z

    Highlights: High temperature pyrolysis of sewage sludge was efficient for producing tar-free fuel gas. Complete tar removal and volatile matter release were at elevated temperature of 1300 C. Sewage sludge was converted to residual solid with high ash content. 72.60% of energy conversion efficiency for gas production in high temperature pyrolysis. Investment and costing for tar cleaning were reduced. - Abstract: Pyrolysis of sewage sludge was studied in a free-fall reactor at 10001400 C. The results showed that the volatile matter in the sludge could be completely released to gaseous product at 1300 C. The high temperature was in favor of H{sub 2} and CO in the produced gas. However, the low heating value (LHV) of the gas decreased from 15.68 MJ/N m{sup 3} to 9.10 MJ/N m{sup 3} with temperature increasing from 1000 C to 1400 C. The obtained residual solid was characterized by high ash content. The energy balance indicated that the most heating value in the sludge was in the gaseous product.

  2. LARGE-SCALE HYDROGEN PRODUCTION FROM NUCLEAR ENERGY USING HIGH TEMPERATURE ELECTROLYSIS

    SciTech Connect (OSTI)

    James E. O'Brien

    2010-08-01T23:59:59.000Z

    Hydrogen can be produced from water splitting with relatively high efficiency using high-temperature electrolysis. This technology makes use of solid-oxide cells, running in the electrolysis mode to produce hydrogen from steam, while consuming electricity and high-temperature process heat. When coupled to an advanced high temperature nuclear reactor, the overall thermal-to-hydrogen efficiency for high-temperature electrolysis can be as high as 50%, which is about double the overall efficiency of conventional low-temperature electrolysis. Current large-scale hydrogen production is based almost exclusively on steam reforming of methane, a method that consumes a precious fossil fuel while emitting carbon dioxide to the atmosphere. Demand for hydrogen is increasing rapidly for refining of increasingly low-grade petroleum resources, such as the Athabasca oil sands and for ammonia-based fertilizer production. Large quantities of hydrogen are also required for carbon-efficient conversion of biomass to liquid fuels. With supplemental nuclear hydrogen, almost all of the carbon in the biomass can be converted to liquid fuels in a nearly carbon-neutral fashion. Ultimately, hydrogen may be employed as a direct transportation fuel in a hydrogen economy. The large quantity of hydrogen that would be required for this concept should be produced without consuming fossil fuels or emitting greenhouse gases. An overview of the high-temperature electrolysis technology will be presented, including basic theory, modeling, and experimental activities. Modeling activities include both computational fluid dynamics and large-scale systems analysis. We have also demonstrated high-temperature electrolysis in our laboratory at the 15 kW scale, achieving a hydrogen production rate in excess of 5500 L/hr.

  3. High Temperature Evaluation of Tantalum Capacitors - Test 1

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

    Cieslewski, Grzegorz

    Tantalum capacitors can provide much higher capacitance at high-temperatures than the ceramic capacitors. This study evaluates selected tantalum capacitors at high temperatures to determine their suitability for you in geothermal field. This data set contains results of the first test where three different types of capacitors were evaluated at 260C.

  4. High Temperature Evaluation of Tantalum Capacitors - Test 1

    SciTech Connect (OSTI)

    Cieslewski, Grzegorz

    2014-09-28T23:59:59.000Z

    Tantalum capacitors can provide much higher capacitance at high-temperatures than the ceramic capacitors. This study evaluates selected tantalum capacitors at high temperatures to determine their suitability for you in geothermal field. This data set contains results of the first test where three different types of capacitors were evaluated at 260C.

  5. ANALYSIS OF FUTURE PRICES AND MARKETS FOR HIGH TEMPERATURE SUPERCONDUCTORS

    E-Print Network [OSTI]

    1 ANALYSIS OF FUTURE PRICES AND MARKETS FOR HIGH TEMPERATURE SUPERCONDUCTORS BY JOSEPH MULHOLLAND temperature superconductors (HTS) may impact the national electrical system over the next 25 years dollars. However, the savings from superconductivity are offset somewhat by the high cost of manufacturing

  6. Calculated Phonon Spectra of Plutonium at High Temperatures

    E-Print Network [OSTI]

    Savrasov, Sergej Y.

    Calculated Phonon Spectra of Plutonium at High Temperatures X. Dai,1 S. Y. Savrasov,2 * G. Kotliar dynamical proper- ties of plutonium using an electronic structure method, which incorporates correlation anharmonic and can be stabilized at high temperatures by its phonon entropy. Plutonium (Pu) is a material

  7. High Temperature Electrolysis of Steam and Carbon Dioxide

    E-Print Network [OSTI]

    High Temperature Electrolysis of Steam and Carbon Dioxide Søren Højgaard Jensen+,#, Jens V. T. Høgh + O2 #12;Electrolysis of steam at high temperature Interesting because · Improved thermodynamic of electrolysis of steam Picture taken from E. Erdle, J. Gross, V. Meyringer, "Solar thermal central receiver

  8. Pyrolysis and gasification of coal at high temperatures

    SciTech Connect (OSTI)

    Zygourakis, K.

    1988-01-01T23:59:59.000Z

    Coals of different ranks will be pyrolyzed in a microscope hot-stage reactor using inert and reacting atmospheres. The macropore structure of the produced chars will be characterized using video microscopy and digital image processing techniques to obtain pore size distributions. Comparative studies will quantify the effect of pyrolysis conditions (heating rates, final heat treatment temperatures, particle size and inert or reacting atmosphere) on the pore structure of the devolatilized chars. The devolatilized chars will be gasified in the regime of strong intraparticle diffusional limitations using O{sub 2}/N{sub 2} and O{sub 2}/H{sub 2}O/N{sub 2}2 mixtures. Constant temperature and programmed-temperature experiments in a TGA will be used for these studies. Additional gasification experiments performed in the hot-stage reactor will be videotaped and selected images will be analyzed to obtain quantitative data on particle shrinkage and fragmentation. Discrete mathematical models will be developed and validated using the experimental gasification data.

  9. Apparatus for monitoring high temperature ultrasonic characterization

    DOE Patents [OSTI]

    Lanagan, M.T.; Kupperman, D.S.; Yaconi, G.A.

    1998-03-24T23:59:59.000Z

    A method and an apparatus for nondestructive detecting and evaluating changes in the microstructural properties of a material by employing one or more magnetostrictive transducers linked to the material by means of one or more sonic signal conductors. The magnetostrictive transducer or transducers are connected to a pulser/receiver which in turn is connected to an oscilloscope. The oscilloscope is connected to a computer which employs an algorithm to evaluate changes in the velocity of a signal transmitted to the material sample as function of time and temperature. 6 figs.

  10. Apparatus for monitoring high temperature ultrasonic characterization

    DOE Patents [OSTI]

    Lanagan, Michael T. (Woodridge, IL); Kupperman, David S. (Oak Park, IL); Yaconi, George A. (Berwyn, IL)

    1998-01-01T23:59:59.000Z

    A method and an apparatus for nondestructive detecting and evaluating chas in the microstructural properties of a material by employing one or more magnetostrictive transducers linked to the material by means of one or more sonic signal conductors. The magnetostrictive transducer or transducers are connected to a pulser/receiver which in turn is connected to an oscilloscope. The oscilloscope is connected to a computer which employs an algorithm to evaluate changes in the velocity of a signal transmitted to the material sample as function of time and temperature.

  11. High Temperature Oxidation Resistance and Surface Electrical...

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

    with Filtered Arc Cr-Al-N Abstract: The requirements for low cost and high-tempurater corrosion resistance for bipolar interconnect plates in solid oxide fuel cell (SOFC) stacks...

  12. An experimental investigation of high temperature, high pressure paper drying

    E-Print Network [OSTI]

    Patel, Kamal Raoji

    1994-01-01T23:59:59.000Z

    % moisture removed oven dried mass of handsheet, g mass of handsheet after drying test, g mass of handsheet before drying test, g relative moisture removed from handsheet moisture removed by drying, % initial moisture (im) initial handsheet sample mass..., and the effects on the paper sheet and drying felt can be detrimental. Elevated temperatures reduce water viscosity which permits reduced resistance to water flow in the sheet. Pressing with a drying temperature of 95 C gives increased drying capacity, reduced...

  13. Design of an Integrated Laboratory Scale Test for Hydrogen Production via High Temperature Electrolysis

    SciTech Connect (OSTI)

    G.K. Housley; K.G. Condie; J.E. O'Brien; C. M. Stoots

    2007-06-01T23:59:59.000Z

    The Idaho National Laboratory (INL) is researching the feasibility of high-temperature steam electrolysis for high-efficiency carbon-free hydrogen production using nuclear energy. Typical temperatures for high-temperature electrolysis (HTE) are between 800-900C, consistent with anticipated coolant outlet temperatures of advanced high-temperature nuclear reactors. An Integrated Laboratory Scale (ILS) test is underway to study issues such as thermal management, multiple-stack electrical configuration, pre-heating of process gases, and heat recuperation that will be crucial in any large-scale implementation of HTE. The current ILS design includes three electrolysis modules in a single hot zone. Of special design significance is preheating of the inlet streams by superheaters to 830C before entering the hot zone. The ILS system is assembled on a 10 x 16 skid that includes electronics, power supplies, air compressor, pumps, superheaters, , hot zone, condensers, and dew-point sensor vessels. The ILS support system consists of three independent, parallel supplies of electrical power, sweep gas streams, and feedstock gas mixtures of hydrogen and steam to the electrolysis modules. Each electrolysis module has its own support and instrumentation system, allowing for independent testing under different operating conditions. The hot zone is an insulated enclosure utilizing electrical heating panels to maintain operating conditions. The target hydrogen production rate for the ILS is 5000 Nl/hr.

  14. High Temperature Interfacial Superconductivity - Energy Innovation Portal

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville PowerCherries 82981-1cnHigh School football High School football Fancy footwork by C. Kim

  15. High Temperature PEM - Energy Innovation Portal

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville PowerCherries 82981-1cnHigh School football High School football Fancy footwork by C.

  16. HIGH TEMPERATURE ELECTROLYZER MATERIALS PROJECT GOAL

    E-Print Network [OSTI]

    Mease, Kenneth D.

    a fuel for the SOFC itself, as a fuel for other devices (e.g., fuel cell vehicles), or as a raw material with compatible electrodes to develop reversible solid oxide fuel cells for low-cost, high efficient power fuel cell concept has been proven, no complete reversible fuel cell materials set has yet been

  17. Method And Apparatus For Evaluatin Of High Temperature Superconductors

    DOE Patents [OSTI]

    Fishman, Ilya M. (Palo Alto, CA); Kino, Gordon S. (Stanford, CA)

    1996-11-12T23:59:59.000Z

    A technique for evaluation of high-T.sub.c superconducting films and single crystals is based on measurement of temperature dependence of differential optical reflectivity of high-T.sub.c materials. In the claimed method, specific parameters of the superconducting transition such as the critical temperature, anisotropy of the differential optical reflectivity response, and the part of the optical losses related to sample quality are measured. The apparatus for performing this technique includes pump and probe sources, cooling means for sweeping sample temperature across the critical temperature and polarization controller for controlling a state of polarization of a probe light beam.

  18. Fast quench reactor and method

    DOE Patents [OSTI]

    Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.

    1998-05-12T23:59:59.000Z

    A fast quench reactor includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This ``freezes`` the desired end product(s) in the heated equilibrium reaction stage. 7 figs.

  19. Partial Safety Analysis for a Reduced Uranium Enrichment Core for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Primm, Trent [ORNL; Gehin, Jess C [ORNL

    2009-04-01T23:59:59.000Z

    A computational model of the reactor core of the High Flux Isotope Rector (HFIR) was developed in order to analyze non-destructive accidents caused by transients during reactor operation. The reactor model was built for the latest version of the nuclear analysis software package called Program for the Analysis of Reactor Transients (PARET). Analyses performed with the model constructed were compared with previous data obtained with other tools in order to benchmark the code. Finally, the model was used to analyze the behavior of the reactor under transients using a different nuclear fuel with lower enrichment of uranium (LEU) than the fuel currently used, which has a high enrichment of uranium (HEU). The study shows that the presence of fertile isotopes in LEU fuel, which increases the neutron resonance absorption, reduces the impact of transients on the fuel and enhances the negative reactivity feedback, thus, within the limitations of this study, making LEU fuel appear to be a safe alternative fuel for the reactor core.

  20. High Temperature Superconductivity in Cuprates: a model

    E-Print Network [OSTI]

    P. R. Silva

    2010-07-16T23:59:59.000Z

    A model is proposed such that quasi-particles (electrons or holes) residing in the CuO2 planes of cuprates may interact leading to metallic or superconducting behaviors. The metallic phase is obtained when the quasi-particles are treated as having classical kinetic energies and the superconducting phase occurs when the quasi-particles are taken as extremely relativistic objects. The interaction between both kinds of particles is provided by a force dependent-on-velocity. In the case of the superconducting behavior, the motion of apical oxygen ions provides the glue to establish the Cooper pair. The model furnishes explicit relations for the Fermi velocity, the perpendicular and the in-plane coherence lengths, the zero-temperature energy gap, the critical current density, the critical parallel and perpendicular magnetic fields. All these mentioned quantities are expressed in terms of fundamental physical constants as: charge and mass of the electron, light velocity in vacuum, Planck constant, electric permittivity of the vacuum. Numerical evaluation of these quantities show that their values are close those found for the superconducting YBaCuO, leading to think the model as being a possible scenario to explain superconductivity in cuprates.

  1. Method for synthesizing extremely high-temperature melting materials

    DOE Patents [OSTI]

    Saboungi, Marie-Louise (Chicago, IL); Glorieux, Benoit (Perpignan, FR)

    2007-11-06T23:59:59.000Z

    The invention relates to a method of synthesizing high-temperature melting materials. More specifically the invention relates to a containerless method of synthesizing very high temperature melting materials such as carbides and transition-metal, lanthanide and actinide oxides, using an aerodynamic levitator and a laser. The object of the invention is to provide a method for synthesizing extremely high-temperature melting materials that are otherwise difficult to produce, without the use of containers, allowing the manipulation of the phase (amorphous/crystalline/metastable) and permitting changes of the environment such as different gaseous compositions.

  2. Method For Synthesizing Extremely High-Temperature Melting Materials

    DOE Patents [OSTI]

    Saboungi, Marie-Louise (Chicago, IL); Glorieux, Benoit (Perpignan, FR)

    2005-11-22T23:59:59.000Z

    The invention relates to a method of synthesizing high-temperature melting materials. More specifically the invention relates to a containerless method of synthesizing very high temperature melting materials such as borides, carbides and transition-metal, lanthanide and actinide oxides, using an Aerodynamic Levitator and a laser. The object of the invention is to provide a method for synthesizing extremely high-temperature melting materials that are otherwise difficult to produce, without the use of containers, allowing the manipulation of the phase (amorphous/crystalline/metastable) and permitting changes of the environment such as different gaseous compositions.

  3. Method for Synthesizing Extremeley High Temperature Melting Materials

    DOE Patents [OSTI]

    Saboungi, Marie-Louise and Glorieux, Benoit

    2005-11-22T23:59:59.000Z

    The invention relates to a method of synthesizing high-temperature melting materials. More specifically the invention relates to a containerless method of synthesizing very high temperature melting materials such as borides, carbides and transition-metal, lanthanide and actinide oxides, using an Aerodynamic Levitator and a laser. The object of the invention is to provide a method for synthesizing extremely high-temperature melting materials that are otherwise difficult to produce, without the use of containers, allowing the manipulation of the phase (amorphous/crystalline/metastable) and permitting changes of the environment such as different gaseous compositions.

  4. Recent Developments in High Temperature Superconductivity

    E-Print Network [OSTI]

    Hor, P. H.

    -Ca-Ba-Cu-O (TCBCO) [5] have been found to be superconducting at as high at 125K in TCBCO. Superconductivity up to - 30K has also been found in the Ba-K-Bi-O type perovskite system [6,7]. Without a copper-oxygen planar structure involved, this system offers a...Can-1 Cu n 04+2n where A =Bi or Tl and B =Ba or Sr and n is the number of CU-O layers stacked consecutively in the unit cell. For the BCSCO and TCBCO compound series, they all have layers of perovskite-like structures (with n =1, 2, or 3...

  5. High Temperature Cements | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are8COaBulkTransmissionSitingProcess.pdfGetecGtel Jump to: navigation, search Name:Hidralia EnergiaFalls,High

  6. High Temperature Materials Laboratory (HTML) - PSD Directorate

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOnItem NotEnergy, science,SpeedingWu,IntelligenceYou are here ‹FIRST CenterAboutHigh Flux

  7. Next Generation Nuclear Plant Project Technology Development Roadmaps: The Technical Path Forward for 750800C Reactor Outlet Temperature

    SciTech Connect (OSTI)

    John Collins

    2009-08-01T23:59:59.000Z

    This document presents the NGNP Critical PASSCs and defines their technical maturation path through Technology Development Roadmaps (TDRMs) and their associated Technology Readiness Levels (TRLs). As the critical PASSCs advance through increasing levels of technical maturity, project risk is reduced and the likelihood of within-budget and on-schedule completion is enhanced. The current supplier-generated TRLs and TDRMs for a 750800C reactor outlet temperature (ROT) specific to each supplier are collected in Appendix A.

  8. Analyzing the Radiation Properties of High-Z Impurities in High-Temperature Plasmas

    SciTech Connect (OSTI)

    Reinke, M. L.; Ince-Cushman, A.; Podpaly, Y.; Rice, J. E. [Plasma Science and Fusion Center, Cambridge, MA (United States); Bitter, M.; Hill, K. W. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Fournier, K. B. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Gu, M. F. [Space Science Laboratory, Berkeley, CA (United States)

    2009-09-10T23:59:59.000Z

    Most tokamak-based reactor concepts require the use of noble gases to form either a radiative mantle or divertor to reduce conductive heat exhaust to tolerable levels for plasma facing components. Predicting the power loss necessary from impurity radiation is done using electron temperature-dependent 'cooling-curves' derived from ab initio atomic physics models. We present here a technique to verify such modeling using highly radiative, argon infused discharges on Alcator C-Mod. A novel x-ray crystal imaging spectrometer is used to measure spatially resolved profiles of line-emissivity, constraining impurity transport simulations. Experimental data from soft x-ray diodes, bare AXUV diodes and foil bolometers are used to determine the local emissivity in three overlapping spectral bands, which are quantitatively compared to models. Comparison of broadband measurements show agreement between experiment and modeling in the core, but not over the entire profile, with the differences likely due to errors in the assumed radial impurity transport outside of the core. Comparison of Ar{sup 16+} x-ray line emission modeling to measurements suggests an additional problem with the collisional-radiative modeling of that charge state.

  9. Viability of Pushrod Dilatometry Techniques for High Temperature In-Pile Measurements

    SciTech Connect (OSTI)

    J. E. Daw; J. L. Rempe; D. L. Knudson; K. G. Condie; J. C. Crepeau

    2008-03-01T23:59:59.000Z

    To evaluate the performance of new fuel, cladding, and structural materials for use in advanced and existing nuclear reactors, robust instrumentation is needed. Changes in material deformation are typically evaluated out-of-pile, where properties of materials are measured after samples were irradiated for a specified length of time. To address this problem, a series of tests were performed to examine the viability of using pushrod dilatometer techniques for in-pile instrumentation to measure deformation. The tests were performed in three phases. First, familiarity was gained in the use and accuracy of this system by testing samples with well defined thermal elongation characteristics. Second, high temperature data for steels, specifically SA533 Grade B, Class 1 (SA533B1) Low Alloy Steel and Stainless Steel 304 (SS304), found in Light Water Reactor (LWR) vessels, were aquired. Finally, data were obtained from a short pushrod in a horizontal geometry to data obtained from a longer pushrod in a vertical geometry, the configuration likely to be used for in-situ measurements. Results of testing show that previously accepted data for the structural steels tested, SA533B1 and SS304, are inaccurate at high temperatures (above 500 oC) due to extrpolation of high temperature data. This is especially true for SA533B1, as previous data do not account for the phase transformation of the material between 730 oC and 830 oC. Also, comparison of results for horizontal and vertical configurations show a maximum percent difference of 2.02% for high temperature data.

  10. INDEPENDENT VERIFICATION SURVEY OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT OUTSIDE AREAS BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    SciTech Connect (OSTI)

    P.C. Weaver

    2010-12-15T23:59:59.000Z

    5098-SR-03-0 FINAL REPORT- INDEPENDENT VERIFICATION SURVEY OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT OUTSIDE AREAS, BROOKHAVEN NATIONAL LABORATORY

  11. LETTER REPORT INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT FAN HOUSE, BUILDING 704 BNL

    SciTech Connect (OSTI)

    P.C. Weaver

    2010-10-22T23:59:59.000Z

    5098-LR-01-0 -LETTER REPORT INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT FAN HOUSE, BUILDING 704 BROOKHAVEN NATIONAL LABORATORY

  12. EIS-0291: High Flux Beam Reactor (HFBR) Transition Project at the Brookhaven National Laboratory, Upton, New York

    Broader source: Energy.gov [DOE]

    The EIS evaluates the range of reasonable alternatives and their impacts regarding the future management of the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL).

  13. High Temperature Irradiation Resistant Thermocouple (HTIR-TC)

    ScienceCinema (OSTI)

    None

    2013-05-28T23:59:59.000Z

    INL researchers have created a new thermocouple that can resist high temperature and radiation. This device will improve safety and reduce costs associated with unit failures. Learn more about INL research at http://www.facebook.com/idahonationallaboratory

  14. Enabling high-temperature nanophotonics for energy applications

    E-Print Network [OSTI]

    Yeng, YiXiang

    The nascent field of high-temperature nanophotonics could potentially enable many important solid-state energy conversion applications, such as thermophotovoltaic energy generation, selective solar absorption, and selective ...

  15. Low Temperature Combustion Demonstrator for High Efficiency Clean...

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

    Merit Review Low Temperature Combustion Demonstrator for High Efficiency Clean Combustion DE-FC26-05NT42413 William de Ojeda International Truck and Engine Company 26 Feb 2008 This...

  16. Electronic properties of doped Mott insulators and high temperature superconductors

    E-Print Network [OSTI]

    Ribeiro, Tiago Castro

    2005-01-01T23:59:59.000Z

    High-temperature superconducting cuprates, which are the quintessential example of a strongly correlated system and the most extensively studied materials after semiconductors, spurred the development in the fields of ...

  17. Copper Aluminate as a potential material for high temperature...

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

    Copper Aluminate as a potential material for high temperature thermoelectric power generation Home Author: D. T. Morelli, E. D. Case, B. D. Hall, S. Wang Year: 2008 Abstract: URL:...

  18. High-Temperature Thermal Array for Next Generation Solar Thermal...

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

    3 Q1 High-Temperature Thermal Array for Next Generation Solar Thermal Power Production - FY13 Q1 This document summarizes the progress of this Los Alamos National Laboratory...

  19. Project Profile: High Operating Temperature Liquid Metal Heat...

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

    A freezing point below 100C Stable at temperatures greater than 800C Low corrosion of stainless steel and high-nickel content alloys A heat capacity greater than 2...

  20. Stability and quench protection of high-temperature superconductors

    E-Print Network [OSTI]

    Ang, Ing Chea

    2006-01-01T23:59:59.000Z

    In the design and operation of a superconducting magnet, stability and protection are two key issues that determine the magnet's reliability and safe operation. Although the high-temperature superconductor (HTS) is considered ...

  1. High Temperature Fuel Cell (Phosphoric Acid) Manufacturing R...

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

    Fuel Cell (Phosphoric Acid) Manufacturing R&D High Temperature Fuel Cell (Phosphoric Acid) Manufacturing R&D Presented at the NREL Hydrogen and Fuel Cell Manufacturing R&D Workshop...

  2. Assessment of Moderate- and High-Temperature Geothermal Resources...

    Open Energy Info (EERE)

    Moderate- and High-Temperature Geothermal Resources of the United States Jump to: navigation, search OpenEI Reference LibraryAdd to library Report: Assessment of Moderate- and...

  3. Low Temperature Combustion Demonstrator for High Efficiency Clean...

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

    Laboratory Department of Energy Project ID ace37deojeda 2 Low Temperature Combustion Demonstrator for High Efficiency Clean Combustion DE-FC26-05NT42413 Project Overview...

  4. Diamond anvil cell for spectroscopic investigation of materials at high temperature, high pressure and shear

    DOE Patents [OSTI]

    Westerfield, C.L.; Morris, J.S.; Agnew, S.F.

    1997-01-14T23:59:59.000Z

    Diamond anvil cell is described for spectroscopic investigation of materials at high temperature, high pressure and shear. A cell is described which, in combination with Fourier transform IR spectroscopy, permits the spectroscopic investigation of boundary layers under conditions of high temperature, high pressure and shear. 4 figs.

  5. Diamond anvil cell for spectroscopic investigation of materials at high temperature, high pressure and shear

    DOE Patents [OSTI]

    Westerfield, Curtis L. (Espanola, NM); Morris, John S. (Los Alamos, NM); Agnew, Stephen F. (Los Alamos, NM)

    1997-01-01T23:59:59.000Z

    Diamond anvil cell for spectroscopic investigation of materials at high temperature, high pressure and shear. A cell is described which, in combination with Fourier transform IR spectroscopy, permits the spectroscopic investigation of boundary layers under conditions of high temperature, high pressure and shear.

  6. Understanding Fundamental Material Degradation Processes in High Temperature Aggressive Chemomechanical Environments

    SciTech Connect (OSTI)

    Stubbins, James; Gewirth, Andrew; Sehitoglu, Huseyin; Sofronis, Petros; Robertson, Ian

    2014-01-16T23:59:59.000Z

    The objective of this project is to develop a fundamental understanding of the mechanisms that limit materials durability for very high-temperature applications. Current design limitations are based on material strength and corrosion resistance. This project will characterize the interactions of high-temperature creep, fatigue, and environmental attack in structural metallic alloys of interest for the very high-temperature gas-cooled reactor (VHTR) or NextGeneration Nuclear Plant (NGNP) and for the associated thermo-chemical processing systems for hydrogen generation. Each of these degradation processes presents a major materials design challenge on its own, but in combination, they can act synergistically to rapidly degrade materials and limit component lives. This research and development effort will provide experimental results to characterize creep-fatigue-environment interactions and develop predictive models to define operation limits for high-temperature structural material applications. Researchers will study individually and in combination creep-fatigue-environmental attack processes in Alloys 617, 230, and 800H, as well as in an advanced Ni-Cr oxide dispersion strengthened steel (ODS) system. For comparison, the study will also examine basic degradation processes in nichrome (Ni-20Cr), which is a basis for most high-temperature structural materials, as well as many of the superalloys. These materials are selected to represent primary candidate alloys, one advanced developmental alloy that may have superior high-temperature durability, and one model system on which basic performance and modeling efforts can be based. The research program is presented in four parts, which all complement each other. The first three are primarily experimental in nature, and the last will tie the work together in a coordinated modeling effort. The sections are 1) dynamic creep-fatigue-environment process, 2) subcritical crack processes, 3) dynamic corrosion crack initiation processes, and 4) modeling.

  7. Low GWP Working Fluid for High Temperature Heat Pumps

    E-Print Network [OSTI]

    Oak Ridge National Laboratory

    Low GWP Working Fluid for High Temperature Heat Pumps: DR-2 Chemical Stability at High Temperatures Temp Heat Pumps: DR-2 Very Low GWP AND Non-Flammable HFC-245fa DR-2 Chemical Formula CF3CH2CHF2 HFO 171.3 Pcr [MPa] 3.65 2.9 Kontomaris-DuPont; European Heat Pump Summit, Nuremberg, October 15th, 2013

  8. Microwave characterization of high-temperature superconductors

    SciTech Connect (OSTI)

    Cooke, D.W.; Gray, E.R.; Arendt, P.N.; Beery, J.G.; Bennett, B.L.; Brown, D.R.; Houlton, R.J.; Jahan, M.S.; Klapetzky, A.J.; Maez, M.A.; Raistrick, I.D.; Reeves, G.A.; Rusnak, B.

    1989-01-01T23:59:59.000Z

    Thick (10-15 {mu}m) Tl-Ba-Ca-Cu-O films have been deposited onto yttria-stabilized zirconia and Ag substrates by d.c. magnetron sputtering techniques. Direct deposition onto 1'' diameter yttria-stabilized zirconia yields films with typical 22 GHz surface resistance (R{sub s}) values of 5.2 {plus minus} 2 m{Omega} and 52 {plus minus} 2 m{Omega} at 10 K and 77 K, respectively. For comparison, R{sub s} of Cu at this same frequency is 10 m{Omega} at 4 K and 22 m{Omega} at 77 K. Tl-Ba-Ca-Cu-O films have also been deposited onto 1'' diameter Ag substrates using Au/Cu, Cu, and BaF{sub 2} buffer layers. The lowest R{sub s} values were obtained on films with a BaF{sub 2} buffer layer, typical values being 7.8 {plus minus} 2 m{Omega} and 30.6 {plus minus} 2 m{Omega} (measured at 22 GHz) at 10 K and 77 K, respectively. Larger films (1.5'' diameter) with similar R{sub s} values were prepared using this same technique, demonstrating that the fabrication process can be scaled to larger surface areas. These films are promising for radiofrequency cavity applications because they are thick (50-75 times the London penetration depth), have relatively large surface areas, are fabricated on metallic substrates, and have R{sub s} values that are competitive with Cu at 77 K and are lower than Cu at 4 K. Because they are polycrystalline and unoriented, it is anticipated that their R{sub s} values can be lowered by improving the processing technique. High-quality films of YBa{sub 2}Cu{sub 3}O{sub 7} have been electron-beam deposited onto 1'' LaGaO{sub 3} and 1.5'' LaAlO{sub 3} substrates. The 1'' sample is characterized by R{sub s} values of 0.2 {plus minus} 0.1 m{Omega} at 4 K and 18.6 {plus minus} 2 m{Omega} at 77 K. The 4-K value is only 2-4 times higher than Nb. The 1.5'' sample has R{sub s} values (measured at 18 GHz) of 0.93 {plus minus} 2 m{Omega} and 71 {plus minus} 3 m{Omega} at 10 K and 77 K, respectively. 18 refs., 8 figs.

  9. Lessons Learned in the Update of a Safety Limit for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Cook, David Howard [ORNL

    2009-01-01T23:59:59.000Z

    A recent unreviewed safety question (USQ) regarding a portion of the High Flux Isotope Reactor (HFIR) transient decay heat removal analysis focused on applicability of a heat transfer correlation at the low flow end of reactor operations. During resolution of this issue, review of the correlations used to establish the safety limit (SL) on reactor flux-to-flow ratio revealed the need to change the magnitude of the SL at the low flow end of reactor operations and the need to update the hot spot fuel damage criteria to incorporate current knowledge involving parallel channel flow stability. Because of the original safety design strategy for the reactor, resolution of the issues for the flux-to-flow ratio involved reevaluation of all key process variable SLs and limiting control settings (LCSs) using the current version of the heat transfer analysis code for the reactor. Goals of the work involved updating and upgrading the SL analysis where necessary, while preserving the safety design strategy for the reactor. Changes made include revisions to the safety design criteria at low flows to address the USQ, update of the process- and analysis input-variable uncertainty considerations, and upgrade of the safety design criteria at high flow. The challenges faced during update/upgrade of this SL and LCS are typical of the problems found in the integration of safety into the design process for a complex facility. In particular, the problems addressed in the area of instrument uncertainties provide valuable lessons learned for establishment and configuration control of SLs for large facilities.

  10. Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 4: High-Temperature Materials PIRTs

    SciTech Connect (OSTI)

    Corwin, William R [ORNL; Ballinger, R. [Massachusetts Institute of Technology (MIT); Majumdar, S. [Argonne National Laboratory (ANL); Weaver, K. D. [Idaho National Laboratory (INL)

    2008-03-01T23:59:59.000Z

    The Phenomena Identification and Ranking Table (PIRT) technique was used to identify safety-relevant/safety-significant phenomena and assess the importance and related knowledge base of high-temperature structural materials issues for the Next Generation Nuclear Plant (NGNP), a very high temperature gas-cooled reactor (VHTR). The major aspects of materials degradation phenomena that may give rise to regulatory safety concern for the NGNP were evaluated for major structural components and the materials comprising them, including metallic and nonmetallic materials for control rods, other reactor internals, and primary circuit components; metallic alloys for very high-temperature service for heat exchangers and turbomachinery, metallic alloys for high-temperature service for the reactor pressure vessel (RPV), other pressure vessels and components in the primary and secondary circuits; and metallic alloys for secondary heat transfer circuits and the balance of plant. These materials phenomena were primarily evaluated with regard to their potential for contributing to fission product release at the site boundary under a variety of event scenarios covering normal operation, anticipated transients, and accidents. Of all the high-temperature metallic components, the one most likely to be heavily challenged in the NGNP will be the intermediate heat exchanger (IHX). Its thin, internal sections must be able to withstand the stresses associated with thermal loading and pressure drops between the primary and secondary loops under the environments and temperatures of interest. Several important materials-related phenomena related to the IHX were identified, including crack initiation and propagation; the lack of experience of primary boundary design methodology limitations for new IHX structures; and manufacturing phenomena for new designs. Specific issues were also identified for RPVs that will likely be too large for shop fabrication and transportation. Validated procedures for on-site welding, post-weld heat treatment (PWHT), and inspections will be required for the materials of construction. High-importance phenomena related to the RPV include crack initiation and subcritical crack growth; field fabrication process control; property control in heavy sections; and the maintenance of high emissivity of the RPV materials over their service lifetime to enable passive heat rejection from the reactor core. All identified phenomena related to the materials of construction for the IHX, RPV, and other components were evaluated and ranked for their potential impact on reactor safety.

  11. Viscosities of natural gases at high pressures and high temperatures

    E-Print Network [OSTI]

    Viswanathan, Anup

    2007-09-17T23:59:59.000Z

    Estimation of viscosities of naturally occurring petroleum gases provides the information needed to accurately work out reservoir-engineering problems. Existing models for viscosity prediction are limited by data, especially at high pressures...

  12. Investigation of magnetic reconnection during a sawtooth crash in a high-temperature tokamak plasma

    E-Print Network [OSTI]

    -temperature Tokamak Fusion Test Reactor (TFTR) plasmas [Plasma Physics and Controlled Nuclear Fusion Research 2986 in a poloidal plane, an observation consistent with the Kadomtsev reconnection theory. On the other hand Test Reactor)17 plasmas (where the magnetic Reynolds number exceeds 107) by a set of nonper- turbative

  13. High poloidal beta long-pulse experiments in the Tokamak Fusion Test Reactor*

    E-Print Network [OSTI]

    Mauel, Michael E.

    High poloidal beta long-pulse experiments in the Tokamak Fusion Test Reactor* J. Kesner+ Plasma stability and confinement. As the current profile evolved, a significantly reduced beta limit was observed after the current ramp-down carried negative current. At later times in lower flN discharges, beta

  14. CRAD, Radiological Controls- Oak Ridge National Laboratory High Flux Isotope Reactor

    Broader source: Energy.gov [DOE]

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

  15. CRAD, Training- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR

    Broader source: Energy.gov [DOE]

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

  16. CRAD, Management- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR

    Broader source: Energy.gov [DOE]

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

  17. CRAD, Conduct of Operations- Oak Ridge National Laboratory High Flux Isotope Reactor

    Broader source: Energy.gov [DOE]

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

  18. CRAD, Engineering- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR

    Broader source: Energy.gov [DOE]

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

  19. CRAD, Environmental Protection- Oak Ridge National Laboratory High Flux Isotope Reactor

    Broader source: Energy.gov [DOE]

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

  20. CRAD, Configuration Management- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR

    Broader source: Energy.gov [DOE]

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