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We encourage you to perform a real-time search of NLEBeta
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1

High temperature gas-cooled reactor: gas turbine application study  

SciTech Connect (OSTI)

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.

Not Available

1980-12-01T23:59:59.000Z

2

Development of a High Temperature Gas-Cooled Reactor TRISO-coated particle fuel chemistry model  

E-Print Network [OSTI]

The first portion of this work is a comprehensive analysis of the chemical environment in a High Temperature Gas-Cooled Reactor TRISO fuel particle. Fission product inventory versus burnup is calculated. Based on those ...

Diecker, Jane T

2005-01-01T23:59:59.000Z

3

Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor  

SciTech Connect (OSTI)

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.

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

2009-10-01T23:59:59.000Z

4

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

SciTech Connect (OSTI)

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 322°C and 750°C, 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.

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

2010-11-01T23:59:59.000Z

5

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

SciTech Connect (OSTI)

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.

Not Available

1980-01-01T23:59:59.000Z

6

Development of MELCOR Input Techniques for High Temperature Gas-Cooled Reactor Analysis  

E-Print Network [OSTI]

High Temperature Gas-cooled Reactors (HTGRs) can provide clean electricity,as well as process heat that can be used to produce hydrogen for transportation and other sectors. A prototypic HTGR, the Next Generation Nuclear Plant (NGNP),will be built...

Corson, James

2011-08-08T23:59:59.000Z

7

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

SciTech Connect (OSTI)

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.

Lee Nelson

2011-09-01T23:59:59.000Z

8

HIGH TEMPERATURE GAS-COOLED REACTOR KNOWLEDGE MANAGEMENT  

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

to assure the safety of the public. Such a request would mean that the reactor would go subcritical upon scram of the outer reflector control rods, and then, after 36 hours for...

9

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

SciTech Connect (OSTI)

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 540°C and the helium coolant was delivered at 7 MPa at 625–925°C. 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.

Lee O. Nelson

2011-04-01T23:59:59.000Z

10

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

SciTech Connect (OSTI)

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.

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

11

High Temperature Gas-cooled Reactor Projected Markets and Scoping Economics  

SciTech Connect (OSTI)

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.

Larry Demick

2010-08-01T23:59:59.000Z

12

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

SciTech Connect (OSTI)

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.

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

1989-04-01T23:59:59.000Z

13

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

SciTech Connect (OSTI)

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).

Wayne Moe

2013-05-01T23:59:59.000Z

14

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

SciTech Connect (OSTI)

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

Not Available

1986-10-01T23:59:59.000Z

15

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

DOE Patents [OSTI]

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.

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

1985-01-01T23:59:59.000Z

16

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

SciTech Connect (OSTI)

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.

Barsell, A.W.

1980-05-01T23:59:59.000Z

17

Modular high temperature gas-cooled reactor plant design duty cycle. Revision 3  

SciTech Connect (OSTI)

This document defines the Plant Design Duty Cycle (PCDC) for the Modular High Temperature Gas-cooled Reactor (MHTGR). The duty cycle is a set of events and their design number of occurrences over the life of the plant for which the MHTGR plant shall be designed to ensure that the plant meets all the top-level requirements. The duty cycle is representative of the types of events to be expected in multiple reactor module-turbine plant configurations of the MHTGR. A synopsis of each PDDC event is presented to provide an overview of the plant response and consequence. 8 refs., 1 fig., 4 tabs.

Chan, T.

1989-12-31T23:59:59.000Z

18

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

SciTech Connect (OSTI)

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

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

2012-06-06T23:59:59.000Z

19

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

SciTech Connect (OSTI)

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 VHTR’s 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.

Chang Oh

2008-02-01T23:59:59.000Z

20

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

SciTech Connect (OSTI)

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.

Rodriguez, C.; Swart, F.

1988-06-01T23:59:59.000Z

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


21

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

SciTech Connect (OSTI)

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.

L.E. Demick

2010-09-01T23:59:59.000Z

22

Mechanical properties of welds in commercial alloys for high-temperature gas-cooled reactor components  

SciTech Connect (OSTI)

Weld properties of Hastelloy-X, Incoloy alloy 800H (with and without Inconel-82 cladding), and 2 1/4 Cr-1 Mo are being studied to provide design data to support the development of steam generator, core auxiliary heat exchanger, and metallic thermal barrier components of the high-temperature gas-cooled reactor (HTGR) steam cycle/cogeneration plant. Tests performed include elevated-temperature creep rupture tests and tensile tests. So far, data from the literature and from relatively short-term tests at GA Technologies Inc. indicate that the weldments are satisfactory for HTGR application.

Lindgren, J.R.; Li, C.C.; Ryder, R.H.; Thurgood, B.E.

1984-07-01T23:59:59.000Z

23

Methods for nondestructive testing of austenitic high-temperature gas-cooled reactor components  

SciTech Connect (OSTI)

Safety-relevant components of high-temperature gas-cooled reactor components are mostly fabricated in nickel-based alloys and austenitic materials like Inconel-617, Hastelloy-X, Nimonic-86, or Incoloy-800H. Compared to ferritic steels, these austenitic materials can have a coarse-grained microstructure, especially in weldments and castings. Coarse-grained or elastic anisotropic materials are difficult to inspect with ultrasonics due to strong attenuation, high noise level (scattering, ''grass'' indications), and sound beam distortions (skewing, splitting, and mode conversion). Only few results dealing with the nondestructive testing of nickel-based alloys are known. The problem area, solutions, and first experiences are reported.

Gobbels, K.; Kapitza, H.

1984-09-01T23:59:59.000Z

24

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

SciTech Connect (OSTI)

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 750°C and provides electricity and/or process heat at 700°C to conventional process applications, including the production of hydrogen.

Ian Mckirdy

2010-12-01T23:59:59.000Z

25

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

SciTech Connect (OSTI)

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).

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

2012-02-01T23:59:59.000Z

26

Helium circulator design considerations for modular high temperature gas-cooled reactor plant  

SciTech Connect (OSTI)

Efforts are in progress to develop a standard modular high temperature gas-cooled reactor (MHTGR) plant that is amenable to design certification and serial production. The MHTGR reference design, based on a steam cycle power conversion system, utilizes a 350 MW(t) annular reactor core with prismatic fuel elements. Flexibility in power rating is afforded by utilizing a multiplicity of the standard module. The circulator, which is an electric motor-driven helium compressor, is a key component in the primary system of the nuclear plant, since it facilitates thermal energy transfer from the reactor core to the steam generator; and, hence, to the external turbo-generator set. This paper highlights the helium circulator design considerations for the reference MHTGR plant and includes a discussion on the major features of the turbomachine concept, operational characteristics, and the technology base that exists in the U.S.

McDonald, C.F.; Nichols, M.K.

1987-01-01T23:59:59.000Z

27

Helium circulator design considerations for modular high temperature gas-cooled reactor plant  

SciTech Connect (OSTI)

Efforts are in progress to develop a standard modular high temperature gas-cooled reactor (MHTGR) plant that is amenable to design certification and serial production. The MHTGR reference design, based on a steam cycle power conversion system, utilizes a 350 MW(t) annular reactor core with prismatic fuel elements. Flexibility in power rating is afforded by utilizing a multiplicity of the standard module. The circulator, which is an electric motor-driven helium compressor, is a key component in the primary system of the nuclear plant, since it facilitates thermal energy transfer from the reactor core to the steam generator; and, hence, to the external turbo-generator set. This paper highlights the helium circulator design considerations for the reference MHTGR plant and includes a discussion on the major features of the turbomachine concept, operational characteristics, and the technology base that exists in the US.

McDonald, C.F.; Nichols, M.K.

1986-12-01T23:59:59.000Z

28

High Temperature Gas-Cooled Reactor Projected Markets and Preliminary Economics  

SciTech Connect (OSTI)

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.

Larry Demick

2011-08-01T23:59:59.000Z

29

Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach  

SciTech Connect (OSTI)

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.

David Petti; Jim Kinsey; Dave Alberstein

2014-01-01T23:59:59.000Z

30

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

SciTech Connect (OSTI)

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

Sonat Sen; Gilles Youinou

2013-02-01T23:59:59.000Z

31

Safety aspects of the Modular High-Temperature Gas-Cooled Reactor (MHTGR)  

SciTech Connect (OSTI)

The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept under development through a cooperative program involving the US Government, the nuclear industry and the utilities. The design utilizes the basic high-temperature gas-cooled reactor (HTGR) features of ceramic fuel, helium coolant, and a graphite moderator. The qualitative top-level safety requirement is that the plant's operation not disturb the normal day-to-day activities of the public. The MHTGR safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles has been evaluated. A broad range of challenges to core heat removal have been examined which include a loss of helium pressure and a simultaneous loss of forced cooling of the core. The challenges to control of heat generation have considered not only the failure to insert the reactivity control systems, but the withdrawal of control rods. Finally, challenges to control chemical attack of the ceramic coated fuel have been considered, including catastrophic failure of the steam generator allowing water ingress or of the pressure vessels allowing air ingress. The plant's response to these extreme challenges is not dependent on operator action and the events considered encompass conceivable operator errors. In the same vein, reliance on radionuclide retention within the full particle and on passive features to perform a few key functions to maintain the fuel within acceptable conditions also reduced susceptibility to external events, site-specific events, and to acts of sabotage and terrorism. 4 refs., 14 figs., 1 tab.

Silady, F.A.; Millunzi, A.C.

1989-08-01T23:59:59.000Z

32

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

SciTech Connect (OSTI)

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.

Gorensek, M.

2011-07-06T23:59:59.000Z

33

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

DOE Patents [OSTI]

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

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

2010-02-23T23:59:59.000Z

34

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

DOE Patents [OSTI]

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

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

2011-03-01T23:59:59.000Z

35

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

DOE Patents [OSTI]

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

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

2013-09-03T23:59:59.000Z

36

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

SciTech Connect (OSTI)

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.

Phillip Mills

2012-02-01T23:59:59.000Z

37

Hastelloy-X for high-temperature gas-cooled reactor applications  

SciTech Connect (OSTI)

Hastelloy-X is a potential structural material for use in gas-cooled reactor systems. In this application, data are necessary on the mechanical properties of base metal and weldments under realistic service conditions. The test environment studied was helium that contained small amounts of H/sub 2/, CH/sub 4/, and CO. This environment was found to be carburizing, with the kinetics of this process becoming rapid above 800/sup 0/C. Suitable weldments of Hastelloy-X were prepared by several processes; those weldments generally had the same properties as base metal except for lower fracture strains under some conditions. Some samples were aged for up to 20 000 h in the test gas and tested, and some creep tests on as-received material exceeded 40 000 h. The predominant effects of aging were the significant reduction in the fracture strains at ambient temperature and the lower strains for samples aged in high-temperature gas-cooled reactor (HTGR) helium than for those aged in inert gas. Under some conditions, aging also resulted in increased yield and ultimate tensile strength. Creep tests failed to show the effects of environment, aging, or welding on the creep strength of Hastelloy-X; however, the fracture strains for weldments were generally lower than they were for base metal. Prior aging in inert gas for 20 000 h at 538 and 871/sup 0/C reduced the fatigue life slightly, but no difference was observed in the fatigue properties of samples aged in air and HTGR helium environments.

McCoy, H.E.; King, J.F.; Strizak, J.P.

1984-07-01T23:59:59.000Z

38

Hastelloy-X for high-temperature gas-cooled reactor applications  

SciTech Connect (OSTI)

Hastelloy-X is a potential structural material for use in gas-cooled reactor systems. In this application, data are necessary on the mechanical properties of base metal and weldments under realistic service conditions. The test environment studied was helium that contained small amounts of H/sub 2/, CH/sub 4/, and CO. This environment was found to be carburizing, with the kinetics of this process becoming rapid above 800/sup 0/C. Suitable weldments of Hastelloy-X were prepared by several processes; those weldments generally had the same properties as base metal except for lower fracture strains under some conditions. Some samples were aged for up to 20000 h in the test gas and tested, and some creep tests on as-received material exceeded 40000 h. The predominant effects of aging were the significant reduction in the fracture strains at ambient temperature and the lower strains for samples aged in high-temperature gas-cooled reactor (HTGR) helium than for those aged in inert gas. Under some conditions, aging also resulted in increased yield and ultimate tensile strength. Creep tests failed to show the effects of environment, aging, or welding on the creep strength of Hastelloy-X; however, the fracture strains for weldments were generally lower than they were for base metal. Prior aging in inert gas for 20000 h at 538 and 871/sup 0/C reduced the fatigue life slightly, but no difference was observed in the fatigue properties of samples aged in air and HTGR helium environments.

McCoy, H.E.; King, J.F.; Strizak, J.P.

1984-07-01T23:59:59.000Z

39

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

SciTech Connect (OSTI)

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.

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

40

Fracture mechanics investigations on high-temperature gas-cooled reactor materials  

SciTech Connect (OSTI)

The prototype nuclear process heat plant and the high-temperature gas-cooled reactor need materials that can withstand temperatures up to 1223 K (950/sup 0/C). An elaboration of fracture mechanics concepts that holds for the complete temperature regime must consider all possible phenomena like creep damage and precipitation during exposure, etc. In tests on the Inconel-617, Hastelloy-X, and Nimonic-86 alloys with respect to fatigue crack growth, creep crack growth, and toughness (J integral R curves) up to 1273 K (1000/sup 0/C), the first creep crack growth results were obtained in helium to compare with the air results. It was shown that pure fatigue crack growth behavior can be described by linear elastic fracture mechanics up to 1273 K. An example of Hastelloy-X at 1223 K proves that evaluating fatigue crack growth according to the J intergral concept gives, within a small scatterband, the same results as by following the linear elastic concept. Hastelloy-X shows a decreasing fracture toughness with increasing temperatures. It is emphasized that the J integral concept holds only if creep deformation can be neglected. The experimental evidence at highest temperatures shows that the J integral R curve is not at all similar to that found at lower temperatures under ideal conditions. Creep crack growth for Nimonic-86 at 1073 less than or equal to T/K less than or equal to 1273 shows that crack growth at 1223 K in helium is found to be larger than in air. Problems arise when correlating the creep crack growth results. The application of the energy rate integral C* seems promising, but this has yet to be proven. A combination of long-term creep with fatigue crack growth is presently impossible.

Krompholz, K.; Bodmann, E.; Gnirss, G.K.; Huthmann, H.

1984-08-01T23:59:59.000Z

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


41

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

SciTech Connect (OSTI)

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.

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

2010-11-02T23:59:59.000Z

42

An integrated performance model for high temperature gas cooled reactor coated particle fuel  

E-Print Network [OSTI]

The performance of coated fuel particles is essential for the development and deployment of High Temperature Gas Reactor (HTGR) systems for future power generation. Fuel performance modeling is indispensable for understanding ...

Wang, Jing, 1976-

2004-01-01T23:59:59.000Z

43

Modular High-Temperature Gas-Cooled Reactor short term thermal response to flow and reactivity transients  

SciTech Connect (OSTI)

The analyses reported here have been conducted at the Oak Ridge National Laboratory (ORNL) for the US Nuclear Regulatory Commission's (NRC's) Division of Regulatory Applications of the Office of Nuclear Regulatory Research. The short-term thermal response of the Modular High-Temperature Gas-Cooled Reactor (MHTGR) is analyzed for a range of flow and reactivity transients. These include loss of forced circulation (LOFC) without scram, moisture ingress, spurious withdrawal of a control rod group, hypothetical large and rapid positive reactivity insertion, and a rapid core cooling event. The coupled heat transfer-neutron kinetics model is also described.

Cleveland, J.C.

1988-01-01T23:59:59.000Z

44

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

SciTech Connect (OSTI)

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.

Chang H. Oh; Eung Soo Kim; Steven Sherman

2008-04-01T23:59:59.000Z

45

THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code  

SciTech Connect (OSTI)

The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

Vondy, D.R.

1984-07-01T23:59:59.000Z

46

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

SciTech Connect (OSTI)

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.

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

2006-08-01T23:59:59.000Z

47

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

SciTech Connect (OSTI)

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.

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

1992-12-01T23:59:59.000Z

48

Mechanical characterization of metallic materials for high-temperature gas-cooled reactors in air and in helium environments  

SciTech Connect (OSTI)

In the French R and D program for high-temperature gas-cooled reactors (HTGRs), three metallic alloys were studied: steel Chromesco-3 with 2.25% chromium, alloy 800H, and Hastelloy-X. The Chromesco-3 and alloy 800H creep behavior is the same in air and in HTGR atmosphere (helium). The tensile tests of Hastelloy-X specimens reveal that aging has embrittlement and hardening effects up to 700/sup 0/C, but the creep tests at 800/sup 0/C show opposite effects. This particular behavior could be due to induced precipitation by aging and the depletion of hardening elements from the matrix. Tests show a low influence of cobalt content on mechanical properties of Hastelloy-X.

Sainfort, G.; Cappelaere, M.; Gregoire, J.; Sannier, J.

1984-07-01T23:59:59.000Z

49

MORECA: A computer code for simulating modular high-temperature gas-cooled reactor core heatup accidents  

SciTech Connect (OSTI)

The design features of the modular high-temperature gas-cooled reactor (MHTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. This report describes the ORNL MORECA code, which was developed for analyzing postulated long-term core heatup scenarios for which active cooling systems used to remove afterheat following the accidents can be assumed to the unavailable. Simulations of long-term loss-of-forced-convection accidents, both with and without depressurization of the primary coolant, have shown that maximum core temperatures stay below the point at which any significant fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. MORECA models the US Department of Energy reference design of a standard MHTGR.

Ball, S.J. (Oak Ridge National Lab., TN (United States))

1991-10-01T23:59:59.000Z

50

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

SciTech Connect (OSTI)

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)

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

51

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

SciTech Connect (OSTI)

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.

Not Available

1982-06-01T23:59:59.000Z

52

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

SciTech Connect (OSTI)

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.

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

1985-10-01T23:59:59.000Z

53

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

SciTech Connect (OSTI)

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.

Not Available

1980-10-09T23:59:59.000Z

54

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

SciTech Connect (OSTI)

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.

L.E. Demick

2011-10-01T23:59:59.000Z

55

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

SciTech Connect (OSTI)

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.

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

2014-07-14T23:59:59.000Z

56

Intermediate Heat Transfer Loop Study for High Temperature Gas-Cooled Reactor  

SciTech Connect (OSTI)

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 cycleefficiency 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. This paper also includes a portion of stress analyses performed on pipe configurations.

C. H. Oh; C. Davis; S. Sherman

2008-08-01T23:59:59.000Z

57

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

SciTech Connect (OSTI)

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.

Karel I. Kingrey

2003-04-01T23:59:59.000Z

58

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

SciTech Connect (OSTI)

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.

J. M. Beck; L. F. Pincock

2011-04-01T23:59:59.000Z

59

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

SciTech Connect (OSTI)

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.

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

60

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

SciTech Connect (OSTI)

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.

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

2013-11-01T23:59:59.000Z

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


61

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

SciTech Connect (OSTI)

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.

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

1984-06-01T23:59:59.000Z

62

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

SciTech Connect (OSTI)

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.

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

63

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

E-Print Network [OSTI]

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

Gitau, Ernest Travis Ngure

2012-10-19T23:59:59.000Z

64

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

SciTech Connect (OSTI)

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.

Not Available

1981-08-01T23:59:59.000Z

65

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)

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.

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

2007-05-02T23:59:59.000Z

66

Gas-cooled nuclear reactor  

DOE Patents [OSTI]

A gas-cooled nuclear reactor includes a central core located in the lower portion of a prestressed concrete reactor vessel. Primary coolant gas flows upward through the core and into four overlying heat-exchangers wherein stream is generated. During normal operation, the return flow of coolant is between the core and the vessel sidewall to a pair of motor-driven circulators located at about the bottom of the concrete pressure vessel. The circulators repressurize the gas coolant and return it back to the core through passageways in the underlying core structure. If during emergency conditions the primary circulators are no longer functioning, the decay heat is effectively removed from the core by means of natural convection circulation. The hot gas rising through the core exits the top of the shroud of the heat-exchangers and flows radially outward to the sidewall of the concrete pressure vessel. A metal liner covers the entire inside concrete surfaces of the concrete pressure vessel, and cooling tubes are welded to the exterior or concrete side of the metal liner. The gas coolant is in direct contact with the interior surface of the metal liner and transfers its heat through the metal liner to the liquid coolant flowing through the cooling tubes. The cooler gas is more dense and creates a downward convection flow in the region between the core and the sidewall until it reaches the bottom of the concrete pressure vessel when it flows radially inward and up into the core for another pass. Water is forced to flow through the cooling tubes to absorb heat from the core at a sufficient rate to remove enough of the decay heat created in the core to prevent overheating of the core or the vessel.

Peinado, Charles O. (La Jolla, CA); Koutz, Stanley L. (San Diego, CA)

1985-01-01T23:59:59.000Z

67

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

E-Print Network [OSTI]

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

Moore, Eugene James Thomas

2006-08-16T23:59:59.000Z

68

Overview of environmental control aspects for the gas-cooled fast reactor  

SciTech Connect (OSTI)

Environmental control aspects relating to release of radionuclides have been analyzed for the Gas-Cooled Fast Reactor (GCFR). Information on environmental control systems was obtained for the most recent GCFR designs, and was used to evaluate the adequacy of these systems. The GCFR has been designed by the General Atomic Company as an alternative to other fast breeder reactor designs, such as the Liquid Metal Fast Breeder Reactor (LMFBR). The GCFR design includes mixed oxide fuel and helium coolant. The environmental impact of expected radionuclide releases from normal operation of the GCFR was evaluated using estimated collective dose equivalent commitments resulting from 1 year of plant operation. The results were compared to equivalent estimates for the Light Water Reactor (LWR) and High-Temperature Gas-Cooled Reactor (HTGR). A discussion of uncertainties in system performances, tritium production rates, and radiation quality factors for tritium is included.

Nolan, A.M.

1981-05-01T23:59:59.000Z

69

A review of existing gas-cooled reactor circulators with application of the lessons learned to the new production reactor circulators  

SciTech Connect (OSTI)

This report presents the results of a study of the lessons learned during the design, testing, and operation of gas-cooled reactor coolant circulators. The intent of this study is to identify failure modes and problem areas of the existing circulators so this information can be incorporated into the design of the circulators for the New Production Reactor (NPR)-Modular High-Temperature Gas Cooled Reactor (MHTGR). The information for this study was obtained primarily from open literature and includes data on high-pressure, high-temperature helium test loop circulators as well as the existing gas cooled reactors worldwide. This investigation indicates that trouble free circulator performance can only be expected when the design program includes a comprehensive prototypical test program, with the results of this test program factored into the final circulator design. 43 refs., 7 tabs.

White, L.S.

1990-07-01T23:59:59.000Z

70

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

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

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

71

Key Thermal Fluid Phenomena In Prismatic Gas-Cooled Reactors  

SciTech Connect (OSTI)

Several types of gas-cooled nuclear reactors have been suggested as part of the international Generation IV initiative with the proposed NGNP (Next Generation Nuclear Plant) as one of the main concepts [MacDonald et al., 2003]. Meaningful studies for these designs will require accurate, reliable predictions of material temperatures to evaluate the material capabilities; these temperatures depend on the thermal convection in the core and in other important components. Some of these reactors feature complex geometries and wide ranges of temperatures, leading to significant variations of the gas thermodynamic and transport properties plus possible effects of buoyancy during normal and reduced power operations and loss-of-flow (LOFA) and loss-of-coolant scenarios. Potential issues identified to date include ''hot streaking'' in the lower plenum evolving from ''hot channels'' in prismatic cores. In order to predict thermal hydraulic behavior of proposed designs effectively and efficiently, it is useful to identify the dominant phenomena occurring.

D. M. McEligot; G. E. McCreery; P. D. Bayless; T. D. Marshall

2005-06-01T23:59:59.000Z

72

Gas-Cooled Fast Reactor (GFR) FY04 Annual Report  

SciTech Connect (OSTI)

The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

K. D. Weaver; T. C. Totemeier; D. E. Clark; E. E. Feldman; E. A. Hoffman; R. B. Vilim; T. Y. C. Wei; J. Gan; M. K. Meyer; W. F. Gale; M. J. Driscoll; M. Golay; G. Apostolakis; K. Czerwinski

2004-09-01T23:59:59.000Z

73

Evaluation of Hastelloy X for gas-cooled-reactor applications  

SciTech Connect (OSTI)

Hastelloy X is a potential structural material for use in gas-cooled reactor systems. In this application data are necessary on the mechanical properties of base metals and weldments under realistic service conditions. The test environment studied was helium that contained small amounts of H/sub 2/, CH/sub 4/, and CO. It is shown that this environment is carburizing with the kinetics of this process, becoming rapid above 800/sup 0/C. Suitable weldments of Hastelloy X were prepared by several processes; those weldments generally had properties similar to the base metal except for lower fracture strains under some conditions. Some samples were aged up to 20,000 h in the test gas and tested, and some creep tests on as-received material exceeded 40,000 h. The predominant effect of aging was the significant reduction of the fracture strains at ambient temperature; the strains were lower when the samples were aged in HTGR helium than when aged in inert gas. Under some conditions aging also increased the yield and ultimate tensile strength. Limited impact testing showed that the impact energy at 25/sup 0/C was reduced drastically by aging at 871 and 704/sup 0/C.

McCoy, H.E.; King, J.F.

1982-11-01T23:59:59.000Z

74

Gas-Cooled Fast Reactor (GFR) FY05 Annual Report  

SciTech Connect (OSTI)

The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom and Switzerland), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report outlines the current design status of the GFR, and includes work done in the areas mentioned above for this fiscal year. In addition, this report fulfills the Level 2 milestones, ''Complete annual status report on GFR reactor design'', and ''Complete annual status report on pre-conceptual GFR reactor designs'' in work package GI0401K01. GFR funding for FY05 included FY04 carryover funds, and was comprised of multiple tasks. These tasks involved a consortium of national laboratories and universities, including the Idaho National Laboratory (INL), Argonne National Laboratory (ANL), Brookhaven National Laboratory (BNL), Oak Ridge National Laboratory (ORNL), Auburn University (AU), Idaho State University (ISU), and the University of Wisconsin-Madison (UW-M). The total funding for FY05 was $1000K, with FY04 carryover of $174K. The cost breakdown can be seen in Table 1.

K. D. Weaver; T. Marshall; T. Totemeier; J. Gan; E.E. Feldman; E.A Hoffman; R.F. Kulak; I.U. Therios; C. P. Tzanos; T.Y.C. Wei; L-Y. Cheng; H. Ludewig; J. Jo; R. Nanstad; W. Corwin; V. G. Krishnardula; W. F. Gale; J. W. Fergus; P. Sabharwall; T. Allen

2005-09-01T23:59:59.000Z

75

Space power reactor ground test in the Experimental Gas Cooled Reactor (EGCR) at Oak Ridge  

SciTech Connect (OSTI)

The Experimental Gas Cooled Reactor (EGCR) facility and the supporting technical infrastructure at the Oak Ridge National Laboratory have the capabilities of performing ground tests of space nuclear power reactor systems. A candidate test would be a 10 MWt lithium cooled reactor, generating potassium vapor that would drive a power turbine. The facility is a large containment vessel originally intended to test the EGCR. Large, contained, and shielded spaces are available for testing, assembly, disassembly, and post-test examination.

Fontana, M.H.; Holcomb, R.S.; Cooper, R.H.

1992-08-01T23:59:59.000Z

76

Core design and reactor physics of a breed and burn gas-cooled fast reactor  

E-Print Network [OSTI]

In order to fulfill the goals set forth by the Generation IV International Forum, the current NERI funded research has focused on the design of a Gas-cooled Fast Reactor (GFR) operating in a Breed and Burnm (B&B) fuel cycle ...

Yarsky, Peter

2005-01-01T23:59:59.000Z

77

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

SciTech Connect (OSTI)

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

Not Available

1980-06-25T23:59:59.000Z

78

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

79

A review of gas-cooled reactor concepts for SDI (Strategic Defense Initiative) applications  

SciTech Connect (OSTI)

We have completed a review of multimegawatt gas-cooled reactor concepts proposed for SDI applications. Our study concluded that the principal reason for considering gas-cooled reactors for burst-mode operation was the potential for significant system mass savings over closed-cycle systems if open-cycle gas-cooled operation (effluent exhausted to space) is acceptable. The principal reason for considering gas-cooled reactors for steady-state operation is that they may represent a lower technology risk than other approaches. In the review, nine gas-cooled reactor concepts were compared to identify the most promising. For burst-mode operation, the NERVA (Nuclear Engine for Rocket Vehicle Application) derivative reactor concept emerged as a strong first choice since its performance exceeds the anticipated operational requirements and the technology has been demonstrated and is retrievable. Although the NERVA derivative concepts were determined to be the lead candidates for the Multimegawatt Steady-State (MMWSS) mode as well, their lead over the other candidates is not as great as for the burst mode. 90 refs., 2 figs., 10 tabs.

Marshall, A.C.

1989-08-01T23:59:59.000Z

80

Experimental Design and Flow Visualization for the Upper Plenum of a Very High Temperature Gas Cooled for Computer Fluid Dynamics Validation  

E-Print Network [OSTI]

The Very High Temperature Reactor (VHTR) is a Generation IV nuclear reactor that is currently under design. It modifies the current high temperature gas reactor (HTGR) design to have a 1000 ^(0)C coolant outlet. This increases fuel efficiency...

Mcvay, Kyle

2014-08-08T23:59:59.000Z

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


81

Baseline Concept Description of a Small Modular High Temperature Reactor  

SciTech Connect (OSTI)

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

Hans Gougar

2014-05-01T23:59:59.000Z

82

Validation of SCALE and the TRITON Depletion Sequence for Gas-Cooled Reactor Analysis  

SciTech Connect (OSTI)

The very-high-temperature reactor (VHTR) is an advanced reactor concept that uses graphite-moderated fuel and helium gas as a coolant. At present there are two primary VHTR reactor designs under consideration for development: in the pebble-bed reactor, a core is loaded with 'pebbles' consisting of 6 cm diameter spheres, while in a high-temperature gas-cooled reactor, fuel rods are placed within prismatic graphite blocks. In both systems, fuel elements (spheres or rods) are comprised of tristructural-isotropic (TRISO) fuel particles. The TRISO particles are either dispersed in the matrix of a graphite pebble for the pebble-bed design or molded into compacts/rods that are then inserted into the hexagonal graphite blocks for the prismatic concept. Two levels of heterogeneity exist in such fuel designs: (1) microspheres of TRISO particles dispersed in a graphite matrix of a cylindrical or spherical shape, and (2) neutron interactions at the rod-to-rod or sphere-to-sphere level. Such double heterogeneity (DH) provides a challenge to multigroup cross-section processing methods, which must treat each level of heterogeneity separately. A new capability to model doubly heterogeneous systems was added to the SCALE system in the release of Version 5.1. It was included in the control sequences CSAS and CSAS6, which use the Monte Carlo codes KENO V.a and KENO-VI, respectively, for three-dimensional neutron transport analyses and in the TRITON sequence, which uses the two-dimensional lattice physics code NEWT along with both versions of KENO for transport and depletion analyses. However, the SCALE 5.1 version of TRITON did not support the use of the DH approach for depletion. This deficiency has been addressed, and DH depletion will be available as an option in the upcoming release of SCALE 6. At present Oak Ridge National Laboratory (ORNL) staff are developing a set of calculations that may be used to validate SCALE for DH calculations. This paper discusses the results of calculations completed to date and the direction of future validation work.

DeHart, Mark D [ORNL; Pritchard, Megan L [ORNL

2008-01-01T23:59:59.000Z

83

Analysis and Development of A Robust Fuel for Gas-Cooled Fast Reactors  

SciTech Connect (OSTI)

The focus of this effort was on the development of an advanced fuel for gas-cooled fast reactor (GFR) applications. This composite design is based on carbide fuel kernels dispersed in a ZrC matrix. The choice of ZrC is based on its high temperature properties and good thermal conductivity and improved retention of fission products to temperatures beyond that of traditional SiC based coated particle fuels. A key component of this study was the development and understanding of advanced fabrication techniques for GFR fuels that have potential to reduce minor actinide (MA) losses during fabrication owing to their higher vapor pressures and greater volatility. The major accomplishments of this work were the study of combustion synthesis methods for fabrication of the ZrC matrix, fabrication of high density UC electrodes for use in the rotating electrode process, production of UC particles by rotating electrode method, integration of UC kernels in the ZrC matrix, and the full characterization of each component. Major accomplishments in the near-term have been the greater characterization of the UC kernels produced by the rotating electrode method and their condition following the integration in the composite (ZrC matrix) following the short time but high temperature combustion synthesis process. This work has generated four journal publications, one conference proceeding paper, and one additional journal paper submitted for publication (under review). The greater significance of the work can be understood in that it achieved an objective of the DOE Generation IV (GenIV) roadmap for GFR Fuel—namely the demonstration of a composite carbide fuel with 30% volume fuel. This near-term accomplishment is even more significant given the expected or possible time frame for implementation of the GFR in the years 2030 -2050 or beyond.

Knight, Travis W

2010-01-31T23:59:59.000Z

84

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

85

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

SciTech Connect (OSTI)

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.

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

2011-05-01T23:59:59.000Z

86

A solution to level 3 dismantling of gas-cooled reactors: Graphite incineration  

SciTech Connect (OSTI)

This paper presents an approach developed to solve the specific decommissioning problems of the G2 and G3 gas cooled reactors at Marcoule and the strategy applied with emphasis in incinerating the graphite core components, using a fluidized-bed incinerator developed jointly between the CEA and FRAMATOME. The incineration option was selected over subsurface storage for technical and economic reasons. Studies have shown that gaseous incineration releases are environmentally acceptable.

Dubourg, M. [FRAMATOME, Paris-La Defense (France)

1993-12-31T23:59:59.000Z

87

Modular Pebble Bed Reactor High Temperature Gas Reactor  

E-Print Network [OSTI]

Modular Pebble Bed Reactor High Temperature Gas Reactor Andrew C Kadak Massachusetts Institute For 1150 MW Combined Heat and Power Station Oil Refinery Hydrogen Production Desalinization Plant VHTR/Graphite Discrimination system Damaged Sphere ContainerGraphiteReturn FuelReturn Fresh Fuel Container Spent Fuel Tank #12

88

A 50-100 kWe gas-cooled reactor for use on Mars.  

SciTech Connect (OSTI)

In the space exploration field there is a general consensus that nuclear reactor powered systems will be extremely desirable for future missions to the outer solar system. Solar systems suffer from the decreasing intensity of solar radiation and relatively low power density. Radioisotope Thermoelectric Generators are limited to generating a few kilowatts electric (kWe). Chemical systems are short-lived due to prodigious fuel use. A well designed 50-100 kWe nuclear reactor power system would provide sufficient power for a variety of long term missions. This thesis will present basic work done on a 50-100 kWe reactor power system that has a reasonable lifespan and would function in an extraterrestrial environment. The system will use a Gas-Cooled Reactor that is directly coupled to a Closed Brayton Cycle (GCR-CBC) power system. Also included will be some variations on the primary design and their effects on the characteristics of the primary design. This thesis also presents a variety of neutronics related calculations, an examination of the reactor's thermal characteristics, feasibility for use in an extraterrestrial environment, and the reactor's safety characteristics in several accident scenarios. While there has been past work for space reactors, the challenges introduced by thin atmospheres like those on Mars have rarely been considered.

Peters, Curtis D. (.)

2006-04-01T23:59:59.000Z

89

Power flattening on modified CANDLE small long life gas-cooled fast reactor  

SciTech Connect (OSTI)

Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

Monado, Fiber [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su'ud, Zaki; Waris, Abdul; Basar, Khairul [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)

2014-09-30T23:59:59.000Z

90

Multi channel thermal hydraulic analysis of gas cooled fast reactor using genetic algorithm  

SciTech Connect (OSTI)

There are three analyzes to be done in the design process of nuclear reactor i.e. neutronic analysis, thermal hydraulic analysis and thermodynamic analysis. The focus in this article is the thermal hydraulic analysis, which has a very important role in terms of system efficiency and the selection of the optimal design. This analysis is performed in a type of Gas Cooled Fast Reactor (GFR) using cooling Helium (He). The heat from nuclear fission reactions in nuclear reactors will be distributed through the process of conduction in fuel elements. Furthermore, the heat is delivered through a process of heat convection in the fluid flow in cooling channel. Temperature changes that occur in the coolant channels cause a decrease in pressure at the top of the reactor core. The governing equations in each channel consist of mass balance, momentum balance, energy balance, mass conservation and ideal gas equation. The problem is reduced to finding flow rates in each channel such that the pressure drops at the top of the reactor core are all equal. The problem is solved numerically with the genetic algorithm method. Flow rates and temperature distribution in each channel are obtained here.

Drajat, R. Z.; Su'ud, Z.; Soewono, E.; Gunawan, A. Y. [Department of Mathematics, Institut Teknologi Bandung, Bandung 40132 (Indonesia); Department of Physics, Institut Teknologi Bandung, Bandung 40132 (Indonesia); Department of Mathematics, Institut Teknologi Bandung, Bandung 40132 (Indonesia)

2012-05-22T23:59:59.000Z

91

ATWS Transients for the 2400 MWt Gas-Cooled Fast Reactor  

SciTech Connect (OSTI)

Reactivity transients have been analyzed with an updated RELAPS-3D (ver. 2.4.2) system model of the pin core design for the 2400MWt gas-cooled fast reactor (GCFR). Additional reactivity parameters were incorporated in the RELAP5 point-kinetics model to account for reactivity feedbacks due to axial and radial expansion of the core, fuel temperature changes (Doppler effect), and pressure changes (helium density changes). Three reactivity transients without scram were analyzed and the incidents were initiated respectively by reactivity ramp, loss of load, and depressurization. During the course of the analysis the turbine bypass model for the power conversion unit (PCU) was revised to enable a better utilization of forced flow cooling after the PCU is tripped. The analysis of the reactivity transients demonstrates the significant impact of the PCU on system pressure and core flow. Results from the modified turbine bypass model suggest a success path for the GCFR to mitigate reactivity transients without scram.

Cheng,L.Y.; Ludewig, H.

2007-08-05T23:59:59.000Z

92

Technical basis for extending storage of the UK's advanced gas-cooled reactor fuel  

SciTech Connect (OSTI)

The UK Nuclear Decommissioning Agency has recently declared a date for cessation of reprocessing of oxide fuel from the UK's Advanced Gas-cooled Reactors (AGRs). This will fundamentally change the management of AGR fuel: from short term storage followed by reprocessing to long term fuel storage followed, in all likelihood, by geological disposal. In terms of infrastructure, the UK has an existing, modern wet storage asset that can be adapted for centralised long term storage of dismantled AGR fuel under the required pond water chemistry. No AGR dry stores exist, although small quantities of fuel have been stored dry as part of experimental programmes in the past. These experimental programmes have shown concerns about corrosion rates.

Hambley, D.I. [National Nuclear Laboratory, Sellafield, Seascale, Cumbria, CA20 1PG (United Kingdom)

2013-07-01T23:59:59.000Z

93

Evaluation of an Integrated Gas-Cooled Reactor Simulator and Brayton Turbine-Generator  

SciTech Connect (OSTI)

A closed-loop Brayton cycle, powered by a fission reactor, offers an attractive option for generating both planetary and in-space electric power. Non-nuclear testing of this type of system provides the opportunity to safely work out integration and system control challenges for a modest investment. Recognizing this potential, a team at Marshall Space Flight Center has evaluated the viability of integrating and testing an existing gas-cooled reactor simulator and a modified, commercially available, Brayton turbine-generator. Since these two systems were developed independently of one another, this evaluation sought to determine if they could be operated together at acceptable power levels, temperatures, and pressures. Thermal, fluid, and structural analyses show that this combined system can operate at acceptable power levels and temperatures. In addition, pressure drops across the reactor simulator, although higher than desired, are also viewed as acceptable. Three potential working fluids for the system were evaluated: N{sub 2}, He/Ar, and He/Xe. Other technical issues, such as electrical breakdown in the generator and the operation of the Brayton foil bearings using various gas mixtures, were also investigated. (authors)

Hissam, D. Andy; Stewart, Eric [National Aeronautics and Space Administration, Marshall Space Flight Center, ER34, Huntsville, AL 35812 (United States)

2006-07-01T23:59:59.000Z

94

Analysis of the conceptual shielding design for the upflow Gas-Cooled Fast Breeder Reactor  

SciTech Connect (OSTI)

Conceptual Shielding Configuration III for the Gas-Cooled Fast Breeder Reactor (GCFR) was analyzed by performing global calculations of neutron and gamma-ray fluences and correcting the results as appropriate with bias factors from localized calculations. Included among the localized calculations were the radial and axial cell streaming calculations, plus extensive preliminary calculations and three final confirmation calculations of the plenum flow-through shields. The global calculations were performed on the GCFR mid-level and the lower and upper plenum regions. Calculated activities were examined with respect to the design constraint, if any, imposed on the particular activity. The spatial distributions of several activities of interest were examined with the aid of isoplots (i.e., symbols are used to describe a surface on which the activity level is everywhere the same). In general the results showed that most activities were below the respective design constraints. Only the total neutron fluence in the core barrel appeared to be marginal with the present reactor design. Since similar results were obtained for an earlier design, it has been proposed that the core barrel be cooled with inlet plenum gas to maintain it at a temperature low enough that it can withstand a higher fluence limit. Radiation levels in the prestressed concrete reactor vessel (PCRV) and liner appeared to be sufficiently below the design constraint that expected results from the Radial Shield Heterogeneity Experiment should not force any levels above the design constraint. A list was also made of a number of issues which should be examined before completion of the final shielding design.

Slater, C.O.; Reed, D.A.; Cramer, S.N.; Emmett, M.B.; Tomlinson, E.T.

1981-01-01T23:59:59.000Z

95

Method to Assess the Radionuclide Inventory of Irradiated Graphite from Gas-Cooled Reactors - 13072  

SciTech Connect (OSTI)

About 17,000 t of irradiated graphite waste will be produced from the decommissioning of the six French gas-cooled nuclear reactors. Determining the radionuclide (RN) content of this waste is of relevant importance for safety reasons and in order to determine the best way to manage them. For many reasons the impurity content that gave rise to the RNs in irradiated graphite by neutron activation during operation is not always well known and sometimes actually unknown. So, assessing the RN content by the use of traditional calculation activation, starting from assumed impurity content, leads to a false assessment. Moreover, radiochemical measurements exhibit very wide discrepancies especially on RN corresponding to precursor at the trace level such as natural chlorine corresponding to chlorine 36. This wide discrepancy is unavoidable and is due to very simple reasons. The level of impurity is very low because the uranium fuel used at that very moment was not enriched, so it was a necessity to have very pure nuclear grade graphite and the very low size of radiochemical sample is a simple technical constraint because device size used to get mineralization product for measurement purpose is limited. The assessment of a radionuclide inventory only based on few number of radiochemical measurements lead in most cases, to a gross over or under-estimation that is detrimental for graphite waste management. A method using an identification calculation-measurement process is proposed in order to assess a radiological inventory for disposal sizing purpose as precise as possible while guaranteeing its upper character. This method present a closer approach to the reality of the main phenomenon at the origin of RNs in a reactor, while also incorporating the secondary effects that can alter this result such as RN (or its precursor) release during reactor operation. (authors)

Poncet, Bernard [EDF-CIDEN, 154 Avenue Thiers, CS 60018, F-69458 LYON cedex 06 (France)] [EDF-CIDEN, 154 Avenue Thiers, CS 60018, F-69458 LYON cedex 06 (France)

2013-07-01T23:59:59.000Z

96

Gas cooled fast reactor control rod drive mechanism deceleration unit. Test program  

SciTech Connect (OSTI)

This report presents the results of the airtesting portion of the proof-of-principle testing of a Control Rod Scram Deceleration Device developed for use in the Gas Cooled Fast Reactor (GCFR). The device utilizes a grooved flywheel to decelerate the translating assembly (T/A). Two cam followers on the translating assembly travel in the flywheel grooves and transfer the energy of the T/A to the flywheel. The grooves in the flywheel are straight for most of the flywheel length. Near the bottom of the T/A stroke the grooves are spiraled in a decreasing slope helix so that the cam followers accelerate the flywheel as they transfer the energy of the falling T/A. To expedite proof-of-principle testing, some of the materials used in the fabrication of certain test article components were not prototypic. With these exceptions the concept appears to be acceptable. The initial test of 300 scrams was completed with only one failure and the failure was that of a non-prototypic cam follower outer sleeve material.

Wagner, T.H.

1981-10-01T23:59:59.000Z

97

The Gas-Cooled Fast Reactor: Report on Safety System Design for Decay Heat Removal  

SciTech Connect (OSTI)

The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radiotoxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. This report addresses/discusses the decay heat removal options available to the GFR, and the current solutions. While it is possible to design a GFR with complete passive safety (i.e., reliance solely on conductive and radiative heat transfer for decay heat removal), it has been shown that the low power density results in unacceptable fuel cycle costs for the GFR. However, increasing power density results in higher decay heat rates, and the attendant temperature increase in the fuel and core. Use of active movers, or blowers/fans, is possible during accident conditions, which only requires 3% of nominal flow to remove the decay heat. Unfortunately, this requires reliance on active systems. In order to incorporate passive systems, innovative designs have been studied, and a mix of passive and active systems appears to meet the requirements for decay heat removal during accident conditions.

K. D. Weaver; T. Marshall; T. Y. C. Wei; E. E. Feldman; M. J. Driscoll; H. Ludewig

2003-09-01T23:59:59.000Z

98

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

SciTech Connect (OSTI)

This report is a preliminary comparison of conventional and potential HTGR-integrated processesa in several common industrial areas: ? Producing electricity via a traditional power cycle ? Producing hydrogen ? Producing ammonia and ammonia-derived products, such as fertilizer ? Producing gasoline and diesel from natural gas or coal ? Producing substitute natural gas from coal, and ? Steam-assisted gravity drainage (extracting oil from tar sands).

Lee Nelson

2009-10-01T23:59:59.000Z

99

Design data needs modular high-temperature gas-cooled reactor. Revision 2  

SciTech Connect (OSTI)

The Design Data Needs (DDNs) provide summary statements for program management, of the designer`s need for experimental data to confirm or validate assumptions made in the design. These assumptions were developed using the Integrated Approach and are tabulated in the Functional Analysis Report. These assumptions were also necessary in the analyses or trade studies (A/TS) to develop selections of hardware design or design requirements. Each DDN includes statements providing traceability to the function and the associated assumption that requires the need.

NONE

1987-03-01T23:59:59.000Z

100

Numerical simulation of flow distribution for pebble bed high temperature gas cooled reactors  

E-Print Network [OSTI]

?....................................................................................26 V CFD MODELLING ...................................................................................28 V.1 Computational Fluid Dynamics (CFD) ..........................................28 V.2 The History.... Hassan The premise of the work presented here is to use a common analytical tool, Computational Fluid Dynamics (CFD), along with different turbulence models. Eddy viscosity models as well as state-of-the-art Large Eddy Simulation (LES) were...

Yesilyurt, Gokhan

2004-09-30T23:59:59.000Z

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


101

Gas-Cooled Thermal Reactor Program. Semiannual technical progress report, April 1, 1983-September 30, 1983  

SciTech Connect (OSTI)

An assessment of the HTGR opportunities from the year 2000 through 2045 was the principal activity on the Market Definition Task (WBS 03). Within the Plant Technology (WBS 13) task, there were activities to develop analytical methods for investigation of Coolant Transport Behavior and to define methods and criteria for High Temperature Structural Engineering design. The activities in support of the HTGR-SC/C Lead Plant (WBS 30 and 31) were the participation in the Lead Plant System Engineering (LPSE) effort and the plant simulation task. The efforts on the Advanced HTGR systems was performed under the Modular Reactor Systems (MRS) (WBS 41) to study the potential for multiple small reactors to provide lower costs, improved safety, and higher availability than the large monolithic core reactors.

Not Available

1983-12-01T23:59:59.000Z

102

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

SciTech Connect (OSTI)

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

Murphy, BD

2004-03-10T23:59:59.000Z

103

FISSION REACTORS KEYWORDS: high-temperature  

E-Print Network [OSTI]

pipeline delivery pressure is assumed to be 7 MPa for the evaluation of the plant performance. The reactor conversion system, and the progress in the electrolysis cell materials field can help the econom- ical on petroleum, and to prepare for the time at which oil reserves become de- pleted. Nuclear energy

Yildiz, Bilge

104

Thermally Simulated 32kW Direct-Drive Gas-Cooled Reactor: Design, Assembly, and Test  

SciTech Connect (OSTI)

One of the power systems under consideration for nuclear electric propulsion is a direct-drive gas-cooled reactor coupled to a Brayton cycle. In this system, power is transferred from the reactor to the Brayton system via a circulated closed loop gas. To allow early utilization, system designs must be relatively simple, easy to fabricate, and easy to test using non-nuclear heaters to closely mimic heat from fission. This combination of attributes will allow pre-prototypic systems to be designed, fabricated, and tested quickly and affordably. The ability to build and test units is key to the success of a nuclear program, especially if an early flight is desired. The ability to perform very realistic non-nuclear testing increases the success probability of the system. In addition, the technologies required by a concept will substantially impact the cost, time, and resources required to develop a successful space reactor power system. This paper describes design features, assembly, and test matrix for the testing of a thermally simulated 32kW direct-drive gas-cooled reactor in the Early Flight Fission - Test Facility (EFF-TF) at Marshall Space Flight Center. The reactor design and test matrix are provided by Los Alamos National Laboratories.

Godfroy, Thomas J.; Bragg-Sitton, Shannon M. [NASA Marshall Space Flight Center, TD40, Huntsville, Alabama, 35812 (United States); University of Michgan, Dept. of Nuclear Engineering and Radiological Sciences, Ann Arbor MI 48109 (United States); Kapernick, Richard J. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

2004-02-04T23:59:59.000Z

105

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

106

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

SciTech Connect (OSTI)

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.

Ingersoll, D.T.

2004-07-29T23:59:59.000Z

107

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

SciTech Connect (OSTI)

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

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

108

Advanced High Temperature Reactor Neutronic Core Design  

SciTech Connect (OSTI)

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.

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

2012-01-01T23:59:59.000Z

109

Refueling Liquid-Salt-Cooled Very High-Temperature Reactors  

SciTech Connect (OSTI)

The liquid-salt-cooled very high-temperature reactor (LS-VHTR), also called the Advanced High-Temperature Reactor (AHTR), is a new reactor concept that combines in a novel way four established technologies: (1) coated-particle graphite-matrix nuclear fuels, (2) Brayton power cycles, (3) passive safety systems and plant designs previously developed for liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants. Depending upon goals, the peak coolant operating temperatures are between 700 and 1000 deg. C, with reactor outputs between 2400 and 4000 MW(t). Several fluoride salt coolants that are being evaluated have melting points between 350 and 500 deg. C, values that imply minimum refueling temperatures between 400 and 550 deg. C. At operating conditions, the liquid salts are transparent and have physical properties similar to those of water. A series of refueling studies have been initiated to (1) confirm the viability of refueling, (2) define methods for safe rapid refueling, and (3) aid the selection of the preferred AHTR design. Three reactor cores with different fuel element designs (prismatic, pebble bed, and pin-type fuel assembly) are being evaluated. Each is a liquid-salt-cooled variant of a graphite-moderated high-temperature reactor. The refueling studies examined applicable refueling experience from high-temperature reactors (similar fuel element designs) and sodium-cooled fast reactors (similar plant design with liquid coolant, high temperatures, and low pressures). The findings indicate that refueling is viable, and several approaches have been identified. The study results are described in this paper. (authors)

Forsberg, Charles W. [Oak Ridge National Laboratory, P.O. Box 2008 Oak Ridge, TN 37831 (United States); Peterson, Per F. [Nuclear Engineering Department, University of California at Berkeley, 6124a Etcheverry Hall, Berkeley, CA 94720 (United States); Cahalan, James E. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States); Enneking, Jeffrey A. [Areva NP (United States); Phil MacDonald [Consultant, Cedar Hill, TX (United States)

2006-07-01T23:59:59.000Z

110

Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices  

SciTech Connect (OSTI)

The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)

Perret, G.; Pattupara, R. M. [Paul Scherrer Inst., 5232 Villigen (Switzerland); Girardin, G. [Ecole Polytechnique Federale de Lausanne, 1015 Lausanne (Switzerland); Chawla, R. [Paul Scherrer Inst., 5232 Villigen (Switzerland); Ecole Polytechnique Federale de Lausanne, 1015 Lausanne (Switzerland)

2012-07-01T23:59:59.000Z

111

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

SciTech Connect (OSTI)

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.

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

2004-10-03T23:59:59.000Z

112

Application of subgroup decomposition in diffusion theory to gas cooled thermal reactor problem  

SciTech Connect (OSTI)

In this paper, the accuracy and computational efficiency of the subgroup decomposition (SGD) method in diffusion theory is assessed in a ID benchmark problem characteristic of gas cooled thermal systems. This method can be viewed as a significant improvement in accuracy of standard coarse-group calculations used for VHTR whole core analysis in which core environmental effect and energy angle coupling are pronounced. It is shown that a 2-group SGD calculation reproduces fine-group (47) results with 1.5 to 6 times faster computational speed depending on the stabilizing schemes while it is as efficient as single standard 6-group diffusion calculation. (authors)

Yasseri, S.; Rahnema, F. [Nuclear and Radiological Engineering and Medical Physics Program, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA 30332-0405 (United States)

2013-07-01T23:59:59.000Z

113

Secondary Heat Exchanger Design and Comparison for Advanced High Temperature Reactor  

SciTech Connect (OSTI)

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

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

2012-06-01T23:59:59.000Z

114

Gas-cooled fast breeder reactor. Quarterly progress report, February 1-April 30, 1980  

SciTech Connect (OSTI)

Information is presented concerning the reactor vessel; reactivity control mechanisms and instrumentation; reactor internals; primary coolant circuits;core auxiliary cooling system; reactor core; systems engineering; and reactor safety and reliability;

Not Available

1980-05-01T23:59:59.000Z

115

Safety philosophy of gas turbine high temperature reactor (GTHTR300)  

SciTech Connect (OSTI)

Japan Atomic Energy Research Institute (JAERI) has undertaken the study of an original design concept of gas turbine high temperature reactor, the GTHTR300. The general concept of this study is development of a greatly simplified design that leads to substantially reduced technical and cost requirements. Newly proposed design features enable the GTHTR300 to be an efficient and economically competitive reactor in 2010's. Also, the GTHTR300 fully takes advantage of its inherent safety characteristics. The safety philosophy of the GTHTR300 is developed based on the HTTR (High Temperature Engineering Test Reactor) of JAERI which is the first HTGR in Japan. Major features of the newly proposed safety philosophy for the GTHTR300 are described in this article. (authors)

Shoji Katanishi; Kazuhiko Kunitomi; Shusaku Shiozawa [Department of Advanced Nuclear Heat Technology, Oarai Research Institute, Japan Atomic Energy Research Institute, Oarai-machi, Ibaraki-ken, 311-1394 (Japan)

2002-07-01T23:59:59.000Z

116

Gas-cooled reactor programs. Fuel-management positioning and accounting module: FUELMANG Version V1. 11, September 1981  

SciTech Connect (OSTI)

This report documents the code module FUELMANG for fuel management of a reactor. This code may be used to position fuel during the calculation of a reactor history, maintain a mass balance history of the fuel movement, and calculate the unit fuel cycle component of the electrical generation cost. In addition to handling fixed feed fuel without recycle, provision has been made for fuel recycle with various options applied to the recycled fuel. A continuous fueling option is also available with the code. A major edit produced by the code is a detailed summary of the mass balance history of the reactor and a fuel cost analysis of that mass balance history. This code is incorporated in the system containing the VENTURE diffusion theory neutronics code for routine use. Fuel movement according to prescribed instructions is performed without the access of additional user input data during the calculation of a reactor operating history. Local application has been primarily for analysis of the performance of gas-cooled thermal reactor core concepts.

Medlin, T.W.; Hill, K.L.; Johnson, G.L.; Jones, J.E.; Vondy, D.R.

1982-01-01T23:59:59.000Z

117

Changes in the mechanical properties of Hastelloy X when exposed to a typical gas-cooled reactor environment  

SciTech Connect (OSTI)

The helium used in a gas-cooled reactor will contain small amounts of H/sub 2/, CO, CH/sub 4/, H/sub 2/O, and N/sub 2/ which can lead to oxidation and carburization/decarburization of structural materials. Long-term creep tests were run on Hastelloy X to 30,000 h at 649 to 871/sup 0/C. It was found that extensive carburization occurred, the minimum creep rate and time to rupture were equal in air and impure helium environments, and the fracture strain was less in helium than in air. Thermal exposure in the temperature range of 538 to 871/sup 0/C resulted in the reduction of ductility in impact and tensile tests at ambient temperature, and this reduction was greater when the exposure was in impure helium rather than in air. A modified alloy with lower chromium and 2% titanium resisted carburization.

McCoy, H.E. Jr.

1981-01-01T23:59:59.000Z

118

Single Channel Testing for Characterization of the Direct Gas Cooled Reactor and the SAFE-100 Heat Exchanger  

SciTech Connect (OSTI)

Experiments have been designed to characterize the coolant gas flow in two space reactor concepts that are currently under investigation by NASA Marshall Space Flight Center and Los Alamos National Laboratory: the direct-drive gas-cooled reactor (DDG) and the SAFE-100 heatpipe-cooled reactor (HPR). For the DDG concept, initial tests have been completed to measure pressure drop versus flow rate for a prototypic core flow channel, with gas exiting to atmospheric pressure conditions. The experimental results of the completed DDG tests presented in this paper validate the predicted results to within a reasonable margin of error. These tests have resulted in a re-design of the flow annulus to reduce the pressure drop. Subsequent tests will be conducted with the re-designed flow channel and with the outlet pressure held at 150 psi (1 MPa). Design of a similar test for a nominal flow channel in the HPR heat exchanger (HPR-HX) has been completed and hardware is currently being assembled for testing this channel at 150 psi. When completed, these test programs will provide the data necessary to validate calculated flow performance for these reactor concepts (pressure drop and film temperature rise)

Bragg-Sitton, S.M. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Propulsion Research Center, NASA Marshall Space Flight Center, Huntsville, AL 35812 (United States); Kapernick, R. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Godfroy, T.J. [Propulsion Research Center, NASA Marshall Space Flight Center, Huntsville, AL 35812 (United States)

2004-02-04T23:59:59.000Z

119

Analysis of the Reactor Cavity Cooling System for Very High Temperature Gas-cooled Reactors Using Computational Fluid Dynamics Tools  

E-Print Network [OSTI]

the VHTR performance and safety analysis, one-dimensional (1-D) system type codes, like RELAP5 or MELCOR, and multi-dimensional CFD codes can be used. The choice of 1-D over multi-dimensional codes first involves identifying the main phenomena, and from...

Frisani, Angelo

2011-08-08T23:59:59.000Z

120

Fluoride Salt-Cooled High-Temperature Reactor Development Roadmap  

SciTech Connect (OSTI)

Fluoride salt-cooled high-temperature reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics and fully passive safety. This paper provides an overview of a technology development pathway for expeditious commercial deployment of first-generation FHRs. The paper describes the principal remaining FHR technology challenges and the development path needed to address the challenges. First-generation FHRs do not appear to require any technology breakthroughs, but will require significant technology development and demonstration. FHRs are currently entering early phase engineering development. As such, the development 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.

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

2014-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "high-temperature gas-cooled reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
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121

EVALUATION OF ZERO-POWER, ELEVATED-TEMPERATURE MEASUREMENTS AT JAPAN’S HIGH TEMPERATURE ENGINEERING TEST REACTOR  

SciTech Connect (OSTI)

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.

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

2011-03-01T23:59:59.000Z

122

Supercell Depletion Studies for Prismatic High Temperature Reactors  

SciTech Connect (OSTI)

The traditional two-step method of analysis is not accurate enough to represent the neutronic effects present in the prismatic high temperature reactor concept. The long range coupling of the various regions in high temperature reactors poses a set of challenges that are not seen in either LWRs or fast reactors. Unlike LWRs, which exhibit large, localized effects, the dominant effects in PMRs are, for the most part, distributed over larger regions, but with lower magnitude. The 1-D in-line treatment currently used in pebble bed reactor analysis is not sufficient because of the 2-D nature of the prismatic blocks. Considerable challenges exist in the modeling of blocks in the vicinity of reflectors, which, for current small modular reactor designs with thin annular cores, include the majority of the blocks. Additional challenges involve the treatment of burnable poisons, operational and shutdown control rods. The use of a large domain for cross section preparation provides a better representation of the neutron spectrum, enables the proper modeling of BPs and CRs, allows the calculation of generalized equivalence theory parameters, and generates a relative power distribution that can be used in compact power reconstruction. The purpose of this paper is to quantify the effects of the reflector, burnable poison, and operational control rods on an LEU design and to delineate an analysis approach for the Idaho National Laboratory. This work concludes that the use of supercells should capture these long-range effects in the preparation of cross sections and along with a set of triangular meshes to treat BPs, and CRs a high fidelity neutronics computation is attainable.

J. Ortensi

2012-10-01T23:59:59.000Z

123

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

SciTech Connect (OSTI)

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.

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

2011-08-01T23:59:59.000Z

124

Computational Analysis of Fluid Flow in Pebble Bed Modular Reactor  

E-Print Network [OSTI]

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...

Gandhir, Akshay

2012-10-19T23:59:59.000Z

125

Evaluation of Alternate Materials for Coated Particle Fuels for the Gas-Cooled Fast Reactor. Laboratory Directed Research and Development Program FY 2006 Final Report  

SciTech Connect (OSTI)

Candidate ceramic materials were studied to determine their suitability as Gas-Cooled Fast Reactor particle fuel coatings. The ceramics examined in this work were: TiC, TiN, ZrC, ZrN, AlN, and SiC. The studies focused on (i) chemical reactivity of the ceramics with fission products palladium and rhodium, (ii) the thermomechanical stresses that develop in the fuel coatings from a variety of causes during burnup, and (iii) the radiation resiliency of the materials. The chemical reactivity of TiC, TiN, ZrC, and ZrN with Pd and Rh were all found to be much lower than that of SiC. A number of important chemical behaviors were observed at the ceramic-metal interfaces, including the formation of specific intermetallic phases and a variation in reaction rates for the different ceramics investigated. Based on the data collected in this work, the nitride ceramics (TiN and ZrN) exhibit chemical behavior that is characterized by lower reaction rates with Pd and Rh than the carbides TiC and ZrC. The thermomechanical stresses in spherical fuel particle ceramic coatings were modeled using finite element analysis, and included contributions from differential thermal expansion, fission gas pressure, fuel kernel swelling, and thermal creep. In general the tangential stresses in the coatings during full reactor operation are tensile, with ZrC showing the lowest values among TiC, ZrC, and SiC (TiN and ZrN were excluded from the comprehensive calculations due to a lack of available materials data). The work has highlighted the fact that thermal creep plays a critical role in the development of the stress state of the coatings by relaxing many of the stresses at high temperatures. To perform ion irradiations of sample materials, an irradiation beamline and high-temperature sample irradiation stage was constructed at the University of Wisconsin’s 1.7MV Tandem Accelerator Facility. This facility is now capable of irradiating of materials to high dose while controlling sample temperature up to 800şC.

Paul A. Demkowicz; Karen Wright; Jian Gan; David Petti; Todd Allen; Jake Blanchard

2006-09-01T23:59:59.000Z

126

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

E-Print Network [OSTI]

if the large portion of the coolant flows into bypass gaps instead of coolant channels in which the cooling efficiency is much higher. A preliminary three dimensional steady-state CFD analysis was performed with commercial code STARCCM+ 6.04 to investigate...

Wang, Huhu 1985-

2012-12-13T23:59:59.000Z

127

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

E-Print Network [OSTI]

relates to validation of systems-level computer code modeling capabilities in anticipation of the eventual need to perform HTGR licensing analyses. Because the NRC has used MELCOR for LWR licensing in the past and because MELCOR was recently updated...

Beeny, Bradley Aaron 1988-

2012-11-12T23:59:59.000Z

128

Benchmarking of the MIT High Temperature Gas-Cooled Reactor TRISO-Coated Particle Fuel Performance Model  

E-Print Network [OSTI]

are specified. This thesis reports TIMCOAT's results from the benchmark study. As this was a blind benchmark and Engineering Thesis Supervisor Accepted by ___________________________________________________________________ Andrew C. Kadak Professor of the Practice of Nuclear Science and Engineering Thesis Reader Accepted

129

Weldability and weld performance of a special grade Hastelloy-X modified for high-temperature gas-cooled reactors  

SciTech Connect (OSTI)

The characteristics of weld defects in the electron beam (EB) welding and the tungsten inert gas (TIG) arc welding for Hastelloy-XR, a modified version of Hastelloy-X, are clarified through the bead-on-plate test and the Trans-Varestraint test. Based on the results, weldabilities on EB and TIG weldings for Hastelloy-XR are discussed and found to be almost the same as Hastelloy-X. The creep rupture behaviors of the welded joints are evaluated by employing data on creep properties of the base and the weld metals. According to the evaluation, the creep rupture strength of the EB-welded joint may be superior to that of the TIG-welded joint. The corrosion test in helium containing certain impurities is conducted for the weld metals. There is no significant difference of such corrosion characteristics as weight gain, internal oxidation, depleted zone, and so on between the base and the weld metals. Those are superior to Hastelloy-X.

Shimizu, S.; Mutoh, Y.

1984-07-01T23:59:59.000Z

130

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

SciTech Connect (OSTI)

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 3.05×10-3 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.

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

2011-05-01T23:59:59.000Z

131

Deterministic Modeling of the High Temperature Test Reactor  

SciTech Connect (OSTI)

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 INL’s 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 Green’s 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 2–3% 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.

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

2010-06-01T23:59:59.000Z

132

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

E-Print Network [OSTI]

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 ...

Minck, Matthew J. (Matthew Joseph)

2013-01-01T23:59:59.000Z

133

Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input  

SciTech Connect (OSTI)

A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

Monado, Fiber, E-mail: fiber.monado@gmail.com [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Aziz, Ferhat [National Nuclear Energy Agency of Indonesia (BATAN) (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)

2014-02-12T23:59:59.000Z

134

Tritium Formation and Mitigation in High Temperature Reactors  

SciTech Connect (OSTI)

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 450–750°C. Results of the diffusion model are presented for one steadystate value of tritium production in the reactor.

Piyush Sabharwall; Carl Stoots

2012-08-01T23:59:59.000Z

135

Review of High Temperature Water and Steam Cooled Reactor Concepts  

SciTech Connect (OSTI)

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

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

2002-07-01T23:59:59.000Z

136

Tritium Formation and Mitigation in High-Temperature Reactors  

SciTech Connect (OSTI)

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 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

Piyush Sabharwall; Carl Stoots

2012-10-01T23:59:59.000Z

137

Tritium Formation and Mitigation in High-Temperature Reactor Systems  

SciTech Connect (OSTI)

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 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

Piyush Sabharwall; Carl Stoots; Hans A. Schmutz

2013-03-01T23:59:59.000Z

138

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

SciTech Connect (OSTI)

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

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

2012-06-06T23:59:59.000Z

139

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

SciTech Connect (OSTI)

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

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

2012-07-01T23:59:59.000Z

140

Optimizing Neutron Thermal Scattering Effects in very High Temperature Reactors  

SciTech Connect (OSTI)

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.

Hawari, Ayman; Ougouag, Abderrafi

2014-07-08T23:59:59.000Z

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


141

Advanced High-Temperature, High-Pressure Transport Reactor Gasification  

SciTech Connect (OSTI)

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

Michael Swanson; Daniel Laudal

2008-03-31T23:59:59.000Z

142

High temperature ceramic membrane reactors for coal liquid upgrading  

SciTech Connect (OSTI)

In this project we will study a novel process concept, i.e., the use of ceramic membrane reactors in upgrading of coal model compounds and coal derived liquids. In general terms, the USC research team is responsible for constructing and operating the membrane reactor apparatus and for testing various inorganic membranes for the upgrading of coal derived asphaltenes and coal model compounds. The USC effort will involve the principal investigator of this project and two graduate research assistants. The ALCOA team is responsible for the preparation of the inorganic membranes, for construction and testing of the ceramic membrane modules, and for measurement of their transport properties. The ALCOA research effort will involve Dr. Paul K. T. Liu, who is the project manager of the ALCOA research team, an engineer and a technician. UNOCAL's contribution will be limited to overall technical assistance in catalyst preparation and the operation of the laboratory upgrading membrane reactor and for analytical back-up and expertise in oil analysis and materials characterization. UNOCAL is a no-cost contractor but will be involved in all aspects of the project, as deemed appropriate.

Tsotsis, T.T.

1992-01-01T23:59:59.000Z

143

High Temperature Gas Reactors: Assessment of Applicable Codes and Standards  

SciTech Connect (OSTI)

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.

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

2011-10-31T23:59:59.000Z

144

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)

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)

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

145

Risk-informed design guidance for a Generation-IV gas-cooled fast reactor emergency core cooling system  

E-Print Network [OSTI]

Fundamental objectives of sustainability, economics, safety and reliability, and proliferation resistance, physical protection and stakeholder relations must be considered during the design of an advanced reactor. However, ...

Delaney, Michael J. (Michael James), 1979-

2004-01-01T23:59:59.000Z

146

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

E-Print Network [OSTI]

The Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is a test reactor concept that aims to demonstrate the neutronics, thermal-hydraulics, materials, tritium management, and to address other reactor operational ...

Rodriguez, Judy N

2013-01-01T23:59:59.000Z

147

Advanced High-Temperature, High-Pressure Transport Reactor Gasification  

SciTech Connect (OSTI)

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

Michael L. Swanson

2005-08-30T23:59:59.000Z

148

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

SciTech Connect (OSTI)

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

Sterbentz, James W

2007-05-01T23:59:59.000Z

149

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

SciTech Connect (OSTI)

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)

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

2011-07-01T23:59:59.000Z

150

Proceedings HTR2006: International Topical Meeting on High Temperature Reactor Technology  

E-Print Network [OSTI]

Proceedings HTR2006: 3rd International Topical Meeting on High Temperature Reactor Technology be effectively modeled using computational fluid dynamics. The NACOK test facility at the Julich Research Center TESTS USING COMPUTATIONAL FLUID DYNAMICS Marie-Anne Brudieu Department of Nuclear Engineering

151

High-temperature gas-cooled reactor helium compatibility studies: results of 10,000-hour exposure of selected alloys in simulated reactor helium  

SciTech Connect (OSTI)

Work on the HTGR Helium Compatibility Task accomplished during the period March 31, 1977 through September 30, 1979, is documented in this report. Emphasis is on the results and analyses of creep data to 10,000 h and the detailed metallurgical evaluations performed on candidate alloy specimens tested for up to 10,000 h. Long-term creep and unstressed aging data in controlled-impurity helium and in air at 800, 900, and 1000/sup 0/C are reported for alloys included in the program in FY-76, including the wrought solid-solution-strengthened alloys, Hastelloy X, Hastelloy S, RA 333, and HD 556, and the centrifugally cast austenitic alloys, HK 40, Supertherm, Manaurite 36X, Manaurite 36XS, and Manaurite 900.

Lechtenberg, T.A.; Stevenson, R.D.; Johnson, W.R.

1980-05-01T23:59:59.000Z

152

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

SciTech Connect (OSTI)

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.

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

2011-01-01T23:59:59.000Z

153

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

SciTech Connect (OSTI)

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.

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

154

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

SciTech Connect (OSTI)

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.

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

1993-06-01T23:59:59.000Z

155

THE INTEGRATION OF PROCESS HEAT APPLICATIONS TO HIGH TEMPERATURE GAS REACTORS  

SciTech Connect (OSTI)

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.

Michael G. McKellar

2011-11-01T23:59:59.000Z

156

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

SciTech Connect (OSTI)

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)

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

2006-07-01T23:59:59.000Z

157

The decomposition of methyltrichlorosilane: Studies in a high-temperature flow reactor  

SciTech Connect (OSTI)

Experimental measurements of the decomposition of methyltrichlorosilane (MTS), a common silicon carbide precursor, in a high-temperature flow reactor are presented. The results indicate that methane and hydrogen chloride are major products of the decomposition. No chlorinated silane products were observed. Hydrogen carrier gas was found to increase the rate of MTS decomposition. The observations suggest a radical-chain mechanism for the decomposition. The implications for silicon carbide chemical vapor deposition are discussed.

Allendorf, M.D.; Osterheld, T.H.; Melius, C.F.

1994-01-01T23:59:59.000Z

158

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

SciTech Connect (OSTI)

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)

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

159

Gas Cooling Through Galaxy Formations  

E-Print Network [OSTI]

Abstract-- Gas cooling was studied in two different boxes of sizes and by simulation at same redshifts. The gas cooling is shown in four different redshifts (z=1.15, 0.5, 0.1 and 0). In the simulation the positions of the clumps of cooled gas were studied with slices of the two volumes and also the density of cooled gas of the two volumes shown in the simulation. From the process of gas cooling it is clear that this process gives different results in the two cases. Index Term- Gas Cooling, Simulation, galaxy Formation. I.

Mariwan A. Rasheed; Mohamad A. Brza

160

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

SciTech Connect (OSTI)

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.

Trond Bjornard; John Hockert

2011-08-01T23:59:59.000Z

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


161

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

SciTech Connect (OSTI)

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.

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

162

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

SciTech Connect (OSTI)

The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge burnup level, while retaining its inherent safety characteristics. Using generic pebble bed reactor cores, this task will perform physics calculations to evaluate the capabilities of the pebble bed reactor to perform utilization and destruction of LWR used-fuel transuranics. The task will use established benchmarked models, and will introduce modeling advancements appropriate to the nature of the fuel considered (high TRU content and high burn-up).

B. Boer; A. M. Ougouag

2010-09-01T23:59:59.000Z

163

Advanced High-Temperature Reactor Dynamic System Model Development: April 2012 Status  

SciTech Connect (OSTI)

The Advanced High-Temperature Reactor (AHTR) is a large-output fluoride-salt-cooled high-temperature reactor (FHR). An early-phase preconceptual design of a 1500 MW(e) power plant was developed in 2011 [Refs. 1 and 2]. An updated version of this plant is shown as Fig. 1. FHRs feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR is designed to be a “walk away” reactor that requires no action to prevent large off-site releases following even severe reactor accidents. This report describes the development of dynamic system models used to further the AHTR design toward that goal. These models predict system response during warmup, startup, normal operation, and limited off-normal operating conditions. Severe accidents that include a loss-of-fluid inventory are not currently modeled. The scope of the models is limited to the plant power system, including the reactor, the primary and intermediate heat transport systems, the power conversion system, and safety-related or auxiliary heat removal systems. The primary coolant system, the intermediate heat transport system and the reactor building structure surrounding them are shown in Fig. 2. These systems are modeled in the most detail because the passive interaction of the primary system with the surrounding structure and heat removal systems, and ultimately the environment, protects the reactor fuel and the vessel from damage during severe reactor transients. The reactor silo also plays an important role during system warmup. The dynamic system modeling tools predict system performance and response. The goal is to accurately predict temperatures and pressures within the primary, intermediate, and power conversion systems and to study the impacts of design changes on those responses. The models are design tools and are not intended to be used in reactor qualification. The important details to capture in the primary system relate to flows within the reactor vessel during severe events and the resulting temperature profiles (temperature and duration) for major components. Critical components include the fuel, reactor vessel, primary piping, and the primary-to-intermediate heat exchangers (P-IHXs). The major AHTR power system loops are shown in Fig. 3. The intermediate heat transfer system is a group of three pumped salt loops that transports the energy produced in the primary system to the power conversion system. Two dynamic system models are used to analyze the AHTR. A Matlab/Simulink?-based model initiated in 2011 has been updated to reflect the evolving design parameters related to the heat flows associated with the reactor vessel. The Matlab model utilizes simplified flow assumptions within the vessel and incorporates an empirical representation of the Direct Reactor Auxiliary Cooling System (DRACS). A Dymola/Modelica? model incorporates a more sophisticated representation of primary coolant flow and a physics-based representation of the three-loop DRACS thermal hydraulics. This model is not currently operating in a fully integrated mode. The Matlab model serves as a prototype and provides verification for the Dymola model, and its use will be phased out as the Dymola model nears completion. The heat exchangers in the system are sized using spreadsheet-based, steady-state calculations. The detail features of the heat exchangers are programmed into the dynamic models, and the overall dimensions are used to generate realistic plant designs. For the modeling cases where the emphasis is on understanding responses within the intermediate and primary systems, the power conversion system may be modeled as a simple boundary condition at the intermediate-to-power conversion system heat exchangers.

Qualls, A.L.; Cetiner, M.S.; Wilson, T.L., Jr.

2012-04-30T23:59:59.000Z

164

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

SciTech Connect (OSTI)

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.

T. Munsat

1999-10-01T23:59:59.000Z

165

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

SciTech Connect (OSTI)

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 540°C and 900°C, 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%.

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

2008-08-01T23:59:59.000Z

166

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

E-Print Network [OSTI]

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 ...

Macdonald, Ruaridh (Ruaridh R.)

2014-01-01T23:59:59.000Z

167

Work Breakdown Structure and Plant/Equipment Designation System Numbering Scheme for the High Temperature Gas- Cooled Reactor (HTGR) Component Test Capability (CTC)  

SciTech Connect (OSTI)

This white paper investigates the potential integration of the CTC work breakdown structure numbering scheme with a plant/equipment numbering system (PNS), or alternatively referred to in industry as a reference designation system (RDS). Ideally, the goal of such integration would be a single, common referencing system for the life cycle of the CTC that supports all the various processes (e.g., information, execution, and control) that necessitate plant and equipment numbers be assigned. This white paper focuses on discovering the full scope of Idaho National Laboratory (INL) processes to which this goal might be applied as well as the factors likely to affect decisions about implementation. Later, a procedure for assigning these numbers will be developed using this white paper as a starting point and that reflects the resolved scope and outcome of associated decisions.

Jeffrey D Bryan

2009-09-01T23:59:59.000Z

168

Maintenance building structural design description: 4 x 350 MW(t) Modular HTGR [High-Temperature Gas-Cooled Reactor] Plant  

SciTech Connect (OSTI)

The Maintenance Building is a grade-founded, two-story, steel-framed structure, located adjacent to the Turbine Building in the Energy Conversion Area. It has a reinforced concrete foundation and slab on grade, and insulated sheet metal exterior walls and roof decking.

NONE

1986-06-01T23:59:59.000Z

169

Assessment of Candidate Molten Salt Coolants for the Advanced High Temperature Reactor (AHTR)  

SciTech Connect (OSTI)

The Advanced High-Temperature Reactor (AHTR) is a novel reactor design that utilizes the graphite-matrix high-temperature fuel of helium-cooled reactors, but provides cooling with a high-temperature fluoride salt. For applications at temperatures greater than 900 C the AHTR is also referred to as a Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR). This report provides an assessment of candidate salts proposed as the primary coolant for the AHTR based upon a review of physical properties, nuclear properties, and chemical factors. The physical properties most relevant for coolant service were reviewed. Key chemical factors that influence material compatibility were also analyzed for the purpose of screening salt candidates. Some simple screening factors related to the nuclear properties of salts were also developed. The moderating ratio and neutron-absorption cross-section were compiled for each salt. The short-lived activation products, long-lived transmutation activity, and reactivity coefficients associated with various salt candidates were estimated using a computational model. Table A presents a summary of the properties of the candidate coolant salts. Certain factors in this table, such as melting point, vapor pressure, and nuclear properties, can be viewed as stand-alone parameters for screening candidates. Heat-transfer properties are considered as a group in Sect. 3 in order to evaluate the combined effects of various factors. In the course of this review, it became apparent that the state of the properties database was strong in some areas and weak in others. A qualitative map of the state of the database and predictive capabilities is given in Table B. It is apparent that the property of thermal conductivity has the greatest uncertainty and is the most difficult to measure. The database, with respect to heat capacity, can be improved with modern instruments and modest effort. In general, ''lighter'' (low-Z) salts tend to exhibit better heat transfer and nuclear performance metrics. Lighter salts also tend to have more favorable (larger) moderating ratios, and thus should have a more favorable coolant-voiding behavior in-core. Heavy (high-Z) salts tend to have lower heat capacities and thermal conductivities and more significant activation and transmutation products. However, all of the salts are relatively good heat-transfer agents. A detailed discussion of each property and the combination of properties that served as a heat-transfer metric is presented in the body of this report. In addition to neutronic metrics, such as moderating ratio and neutron absorption, the activation properties of the salts were investigated (Table C). Again, lighter salts tend to have more favorable activation properties compared to salts with high atomic-number constituents. A simple model for estimating the reactivity coefficients associated with a reduction of salt content in the core (voiding or thermal expansion) was also developed, and the primary parameters were investigated. It appears that reasonable design flexibility exists to select a safe combination of fuel-element design and salt coolant for most of the candidate salts. Materials compatibility is an overriding consideration for high-temperature reactors; therefore the question was posed whether any one of the candidate salts was inherently, or significantly, more corrosive than another. This is a very complex subject, and it was not possible to exclude any fluoride salts based on the corrosion database. The corrosion database clearly indicates superior container alloys, but the effect of salt identity is masked by many factors which are likely more important (impurities, redox condition) in the testing evidence than salt identity. Despite this uncertainty, some reasonable preferences can be recommended, and these are indicated in the conclusions. The reasoning to support these conclusions is established in the body of this report.

Williams, D.F.

2006-03-24T23:59:59.000Z

170

Lessons Learned From Gen I Carbon Dioxide Cooled Reactors  

SciTech Connect (OSTI)

This paper provides a review of early gas cooled reactors including the Magnox reactors originating in the United Kingdom and the subsequent development of the Advanced Gas-cooled Reactors (AGR). These early gas cooled reactors shared a common coolant medium, namely carbon dioxide (CO2). A framework of information is provided about these early reactors and identifies unique problems/opportunities associated with use of CO2 as a coolant. Reactor designers successfully rose to these challenges. After years of successful use of the CO2 gas cooled reactors in Europe, the succeeding generation of reactors, called the High Temperature Gas Reactors (HTGR), were designed with Helium gas as the coolant. Again, in the 21st century, with the latest reactor designs under investigation in Generation IV, there is a revived interest in developing Gas Cooled Fast Reactors that use CO2 as the reactor coolant. This paper provides a historical perspective on the 52 CO2 reactors and the reactor programs that developed them. The Magnox and AGR design features and safety characteristics were reviewed, as well as the technologies associated with fuel storage, reprocessing, and disposal. Lessons-learned from these programs are noted to benefit the designs of future generations of gas cooled nuclear reactors.

David E. Shropshire

2004-04-01T23:59:59.000Z

171

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

E-Print Network [OSTI]

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 900°C and 950°C, so the core temperature is one of the key features in the design...

Kanjanakijkasem, Worasit 1975-

2012-11-29T23:59:59.000Z

172

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

SciTech Connect (OSTI)

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.

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

2007-09-01T23:59:59.000Z

173

Computational Fluid Dynamics Analyses on Very High Temperature Reactor Air Ingress  

SciTech Connect (OSTI)

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.

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

2009-07-01T23:59:59.000Z

174

Evaporation behavior of Hastelloy-X alloys in simulated very high temperature reactor environments  

SciTech Connect (OSTI)

A sequential analysis was made on the material degradations during exposure of nickel-base corrosionresistant austenitic alloys to simulated very high temperature reactor environments. The materials tested were two modified versions of Hastelloy-X in terms of both increased manganese content for improved compatibility and decreased manganese content for possible adverse effects. Quantitative analysis of the specimens after exposure for 1000 h at several temperature steps from 850 to 1050/sup 0/C have revealed the temperature-dependent aspects of the processes including the depletion of chromium and manganese due to oxidation, evaporation, and carbon transfer into and/or from the materials. The material with enriched manganese, developed and specified as Hastelloy-XR, showed enhanced resistance to loss of chromium in terms of both oxidation and evaporation.

Shindo, M.; Kondo, T.

1984-08-01T23:59:59.000Z

175

Development and Verification of Tritium Analyses Code for a Very High Temperature Reactor  

SciTech Connect (OSTI)

A tritium permeation analyses code (TPAC) has been developed by Idaho National Laboratory for the purpose of analyzing tritium distributions in the VHTR systems including integrated hydrogen production systems. A MATLAB SIMULINK software package was used for development of the code. The TPAC is based on the mass balance equations of tritium-containing species and a various form of hydrogen (i.e., HT, H2, HTO, HTSO4, and TI) coupled with a variety of tritium source, sink, and permeation models. In the TPAC, ternary fission and neutron reactions with 6Li, 7Li 10B, 3He were taken into considerations as tritium sources. Purification and leakage models were implemented as main tritium sinks. Permeation of HT and H2 through pipes, vessels, and heat exchangers were importantly considered as main tritium transport paths. In addition, electroyzer and isotope exchange models were developed for analyzing hydrogen production systems including both high-temperature electrolysis and sulfur-iodine process. The TPAC has unlimited flexibility for the system configurations, and provides easy drag-and-drops for making models by adopting a graphical user interface. Verification of the code has been performed by comparisons with the analytical solutions and the experimental data based on the Peach Bottom reactor design. The preliminary results calculated with a former tritium analyses code, THYTAN which was developed in Japan and adopted by Japan Atomic Energy Agency were also compared with the TPAC solutions. This report contains descriptions of the basic tritium pathways, theory, simple user guide, verifications, sensitivity studies, sample cases, and code tutorials. Tritium behaviors in a very high temperature reactor/high temperature steam electrolysis system have been analyzed by the TPAC based on the reference indirect parallel configuration proposed by Oh et al. (2007). This analysis showed that only 0.4% of tritium released from the core is transferred to the product hydrogen. The amount of tritium in the product hydrogen was estimated to be approximately an order less than the gaseous effluent limit for tritium.

Chang H. Oh; Eung S. Kim

2009-09-01T23:59:59.000Z

176

Vortex Diode Analysis and Testing for Fluoride Salt-Cooled High-Temperature Reactors  

SciTech Connect (OSTI)

Fluidic diodes are presently being considered for use in several fluoride salt-cooled high-temperature reactor designs. A fluidic diode is a passive device that acts as a leaky check valve. These devices are installed in emergency heat removal systems that are designed to passively remove reactor decay heat using natural circulation. The direct reactor auxiliary cooling system (DRACS) uses DRACS salt-to-salt heat exchangers (DHXs) that operate in a path parallel to the core flow. Because of this geometry, under normal operating conditions some flow bypasses the core and flows through the DHX. A flow diode, operating in reverse direction, is-used to minimize this flow when the primary coolant pumps are in operation, while allowing forward flow through the DHX under natural circulation conditions. The DRACSs reject the core decay heat to the environment under loss-of-flow accident conditions and as such are a reactor safety feature. Fluidic diodes have not previously been used in an operating reactor system, and therefore their characteristics must be quantified to ensure successful operation. This report parametrically examines multiple design parameters of a vortex-type fluidic diode to determine the size of diode needed to reject a particular amount of decay heat. Additional calculations were performed to size a scaled diode that could be tested in the Oak Ridge National Laboratory Liquid Salt Flow Loop. These parametric studies have shown that a 152.4 mm diode could be used as a test article in that facility. A design for this diode is developed, and changes to the loop that will be necessary to test the diode are discussed. Initial testing of a scaled flow diode has been carried out in a water loop. The 150 mm diode design discussed above was modified to improve performance, and the final design tested was a 171.45 mm diameter vortex diode. The results of this testing indicate that diodicities of about 20 can be obtained for diodes of this size. Experimental results show similar trends as the computational fluid dynamics (CFD) results presented in this report; however, some differences exist that will need to be assessed in future studies. The results of this testing will be used to improve the diode design to be tested in the liquid salt loop system.

Yoder Jr, Graydon L [ORNL; Elkassabgi, Yousri M. [Texas A& M University, Kingsville; De Leon, Gerardo I. [Texas A& M University, Kingsville; Fetterly, Caitlin N. [Texas A& M University, Kingsville; Ramos, Jorge A. [Texas A& M University, Kingsville; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)

2012-02-01T23:59:59.000Z

177

Reactor siting risk comparisons related to recommendations of NUREG-0625  

SciTech Connect (OSTI)

This document evaluates how implementing the remote siting recommendations for nuclear reactors (NUREG-0625) made by the Siting Policy Task Force of the US Nuclear Regulatory Commission (NRC) can reduce potential public risk. The document analyzes how population density affects site-specific risk for both light water reactors (LWRs) and high-temperature gas-cooled reactors (HTGRs).

Barsell, A.W.; Dombek, F.S.; Orvis, D.D.

1980-11-01T23:59:59.000Z

178

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

SciTech Connect (OSTI)

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)

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

179

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

SciTech Connect (OSTI)

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 â??warm boreâ?ť 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 â??spiderâ?ť 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 â??Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limitersâ?ť was to stimulate the HTS wire industry with, first 1G, then 2G, HTS wire applications. Over the approximately 5 years of ZPâ??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.

Frank Darmann; Robert Lombaerde; Franco Moriconi; Albert Nelson

2011-10-31T23:59:59.000Z

180

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

SciTech Connect (OSTI)

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.

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

2012-11-05T23:59:59.000Z

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


181

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

SciTech Connect (OSTI)

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 INL’s 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 2–3% 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.

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

2010-10-01T23:59:59.000Z

182

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

SciTech Connect (OSTI)

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.

Hassan, Yassin

2013-10-22T23:59:59.000Z

183

Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation  

E-Print Network [OSTI]

Gas Expansion Module Gas-cooled Fast Reactor High Enrichedfast reactors: gas-cooled fast reactor (GFR), sodium-cooledderived from the Gas cooled Fast Reactor (GFR). This core

Heidet, Florent

2010-01-01T23:59:59.000Z

184

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

E-Print Network [OSTI]

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.

S. I. Fisenko; I. S. Fisenko

2007-05-27T23:59:59.000Z

185

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...

186

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

SciTech Connect (OSTI)

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

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

2010-09-01T23:59:59.000Z

187

Transmutation Analysis of Enriched Uranium and Deep Burn High Temperature Reactors  

SciTech Connect (OSTI)

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.

Michael A. Pope

2012-07-01T23:59:59.000Z

188

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

SciTech Connect (OSTI)

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.

Richard Schultz

2012-04-01T23:59:59.000Z

189

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

SciTech Connect (OSTI)

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.

Curtis Smith; Scott Beck; Bill Galyean

2005-09-01T23:59:59.000Z

190

Gas cooled traction drive inverter  

SciTech Connect (OSTI)

The present invention provides a modular circuit card configuration for distributing heat among a plurality of circuit cards. Each circuit card includes a housing adapted to dissipate heat in response to gas flow over the housing. In one aspect, a gas-cooled inverter includes a plurality of inverter circuit cards, and a plurality of circuit card housings, each of which encloses one of the plurality of inverter cards.

Chinthavali, Madhu Sudhan

2013-10-08T23:59:59.000Z

191

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

SciTech Connect (OSTI)

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.

Hawari, Ayman I

2011-08-30T23:59:59.000Z

192

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)

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.

Hans Gougar

2010-02-01T23:59:59.000Z

193

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

SciTech Connect (OSTI)

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 IAEA’s 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.

Philip Casey Durst; Mark Schanfein

2012-08-01T23:59:59.000Z

194

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

SciTech Connect (OSTI)

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 (390°C 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

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

2012-09-01T23:59:59.000Z

195

Development of hollow fiber catalytic membrane reactors for high temperature gas cleanup. Final report, September 1989--March 1994  

SciTech Connect (OSTI)

The objective of this project was to develop economically and technically viable catalytic membrane reactors for high temperature, high pressure gaseous contaminant control in Integrated Gasification Combined Cycle (IGCC) systems. These catalytic membrane reactors decompose H{sub 2}S and separate the reaction products. The reactors were designed to operate in the hostile process environment of the IGCC systems, and at temperatures ranging from 500 to 1000{degrees}C. Severe conditions encountered in the IGCC process (e.g., 900{degrees}C, containing of H{sub 2}S, CO{sub 2} and H{sub 2}O) make it impossible to use polymeric membranes in the process. A list of inorganic membranes that can be employed in the membrane reactor includes Pd metallic membranes, molecular-sieve glass membranes (PPG Industries), porous Vycor glass membranes and porous sol-gel derived membranes such as alumina, zirconia. Alumina and zirconia membranes, however, cannot withstand for a long time at high temperatures in the presence of water vapors. Palladium membranes are a very promising class of inorganic membranes for gas separations that is currently under development. In this project two different types of membranes were used in the design of the membrane reactor -- molecular-sieve glass membrane and Vycor glass porous membrane.

Ma, Yi Hua; Moser, W.R.; Pien, S.; Shelekhin, A.B.

1994-07-01T23:59:59.000Z

196

E-Print Network 3.0 - advanced gas-cooled nuclear Sample Search...  

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

Energy 9 UNM engineering s p r i N g 2 0 0 6 Summary: Re); the gas cooled Submersion-Subcritical Safe Space reactor (S4) designed to operate with multiple Closed... for reducing...

197

IDAHO NATIONAL LABORATORY PROGRAM TO OBTAIN BENCHMARK DATA ON THE FLOW PHENOMENA IN A SCALED MODEL OF A PRISMATIC GAS-COOLED REACTOR LOWER PLENUM FOR THE VALIDATION OF CFD CODES  

SciTech Connect (OSTI)

The experimental program that is being conducted at the Matched Index-of-Refraction (MIR) Flow Facility at Idaho National Laboratory (INL) to obtain benchmark data on measurements of flow phenomena in a scaled model of a typical prismatic gas-cooled (GCR) reactor lower plenum using 3-D Particle Image Velocimetry (PIV) is presented. A detailed description of the model, scaling, the experimental facility, 3-D PIV system, measurement uncertainties and analysis, experimental procedures and samples of the data sets that have been obtained are included. Samples of the data set that are presented include mean-velocity-field and turbulence data in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic GCR design. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. Results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). Inlet jet Reynolds numbers (based on the jet diameter and the time-mean average flow rate) are approximately 4,300 and 12,400. The measurements reveal undeveloped, non-uniform flow in the inlet jets and complicated flow patterns in the model lower plenum. Data include three-dimensional vector plots, data displays along the coordinate planes (slices) and charts that describe the component flows at specific regions in the model. Information on inlet flow is also presented.

Hugh M. McIlroy Jr.; Donald M. McEligot; Robert J. Pink

2008-09-01T23:59:59.000Z

198

Macroscopic Mechanistic Modeling and Optimization of a Self-Initiated High-Temperature Polymerization Reactor  

E-Print Network [OSTI]

and optimization study of a batch polymerization reactor in which self-initiated free-radical poly- merization of n to calculate an optimal batch-reactor temperature profile that yields an end-batch polymer product with desiredMacroscopic Mechanistic Modeling and Optimization of a Self-Initiated High

Rappe, Andrew M.

199

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

E-Print Network [OSTI]

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...

Ames, David E, II

2006-10-30T23:59:59.000Z

200

Assessment of passive decay heat removal in the General Atomics Modular Helium Reactor  

E-Print Network [OSTI]

/ATHENA. The MHR is a high temperature gas cooled reactor. It is a prismatic core concept for New Generation Nuclear Plant (NGNP). Very few reactors of that kind have been designed in the past. Furthermore, the MHR is supposed to be a highly passively safe concept...

Cocheme, Francois Guilhem

2005-02-17T23:59:59.000Z

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


201

"A New Class od Functionally Graded Cearamic-Metal Composites for Next Generation Very High Temperature Reactors"  

SciTech Connect (OSTI)

Generation IV Very High Temperature power generating nuclear reactors will operate at temperatures greater than 900 oC. At these temperatures, the components operating in these reactors need to be fabricated from materials with excellent thermo-mechanical properties. Conventional pure or composite materials have fallen short in delivering the desired performance. New materials, or conventional materials with new microstructures, and associated processing technologies are needed to meet these materials challenges. Using the concept of functionally graded materials, we have fabricated a composite material which has taken advantages of the mechanical and thermal properties of ceramic and metals. Functionally-graded composite samples with various microstructures were fabricated. It was demonstrated that the composition and spatial variation in the composition of the composite can be controlled. Some of the samples were tested for irradiation resistance to neutrons. The samples did not degrade during initial neutron irradiation testing.

Dr. Mohit Jain; Dr. Ganesh Skandan; Dr. Gordon E. Khose; Mrs. Judith Maro, Nuclear Reactor Laboratory, MIT

2008-05-01T23:59:59.000Z

202

Application of a moving bed biofilm reactor for tertiary ammonia treatment in high temperature industrial wastewater  

E-Print Network [OSTI]

industrial wastewater Jennifer L. Shore a,b , William S. M'Coy b , Claudia K. Gunsch a , Marc A. Deshusses a 2012 Available online 17 February 2012 Keywords: Moving bed biofilm reactor Industrial wastewater and industrial wastewater. No biotreatment was observed at 45 °C, although effective nitrification was rapidly

203

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

SciTech Connect (OSTI)

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

Farzad Rahnema; Dingkang Zhang; Abderrafi Ougouag; Frederick Gleicher

2011-04-04T23:59:59.000Z

204

Research and Development of High Temperature Light Water Cooled Reactor Operating at Supercritical-Pressure in Japan  

SciTech Connect (OSTI)

This paper summarizes the status and future plans of research and development of the high temperature light water cooled reactor operating at supercritical-pressure in Japan. It includes; the concept development; material for the fuel cladding; water chemistry under supercritical pressure; thermal hydraulics of supercritical fluid; and the conceptual design of core and plant system. Elements of concept development of the once-through coolant cycle reactor are described, which consists of fuel, core, reactor and plant system, stability and safety. Material studies include corrosion tests with supercritical water loops and simulated irradiation tests using a high-energy transmission electron microscope. Possibilities of oxide dispersion strengthening steels for the cladding material are studied. The water chemistry research includes radiolysis and kinetics of supercritical pressure water, influence of radiolysis and radiation damage on corrosion and behavior on the interface between water and material. The thermal hydraulic research includes heat transfer tests of single tube, single rod and three-rod bundles with a supercritical Freon loop and numerical simulations. The conceptual designs include core design with a three-dimensional core simulator and sub-channel analysis, and balance of plant. (authors)

Yoshiaki Oka [Nuclear Engineering Research Laboratory, The University of Tokyo, 7-3-1, Hongo, Bunkyo-ku, Tokyo, 112-0006 (Japan); Katsumi Yamada [Isogo Nuclear Engineering Center, Toshiba Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan)

2004-07-01T23:59:59.000Z

205

Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors  

E-Print Network [OSTI]

systems for the Gas Cooled Fast Reactor (GCFR) includes theThey are 1) gas cooled fast reactors (GFR), 2) very high

Galvez, Cristhian

2011-01-01T23:59:59.000Z

206

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

SciTech Connect (OSTI)

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.

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

207

Catalytic membrane program novation: High temperature catalytic membrane reactors. Final report  

SciTech Connect (OSTI)

The original objective was to develop an energy-efficient hydrocarbon dehydrogenation process based on catalytic membrane reactors. Golden Technologies determined that the goals of this contract would be best served by novating the contract to an end user or other interested party which is better informed on the economic justification aspects of petrochemical refining processes to carry out the remaining work. In light of the Chevron results, the program objective was broadened to include development of inorganic membranes for applications in the chemical industry. The proposed membrane technologies shall offer the potential to improve chemical production processes via conversion increase and energy savings. The objective of this subcontract is to seek a party that would serve as a prime contractor to carry out the remaining tasks on the agreement and bring the agreement to a successful conclusion. Four tasks were defined to select the prime contractor. They were (1) prepare a request for proposal, (2) solicit companies as potential prime contractors as well as team members, (3) discuss modifications requested by the potential prime contractors, and (4) obtain, review and rank the proposals. The accomplishments on the tasks is described in detail in the following sections.

Kleiner, R.N. [Kleiner (Richard N.), Englewood, CO (United States)] [Kleiner (Richard N.), Englewood, CO (United States)

1998-08-28T23:59:59.000Z

208

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

SciTech Connect (OSTI)

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

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

2014-03-01T23:59:59.000Z

209

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

E-Print Network [OSTI]

shaft. With the impeller in place and rotating, gas was drawn into the top port and ejected at the impeller mount. The reactor pressure was monitored via the transducer port. The transducer was a Viatran Pressure Transducer, model 103. The liquid...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...

Hess, Richard Kenneth

2012-06-07T23:59:59.000Z

210

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

SciTech Connect (OSTI)

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)

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

211

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)

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.

Michael A. Pope

2011-10-01T23:59:59.000Z

212

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

SciTech Connect (OSTI)

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.

Ingersoll, DT

2005-12-15T23:59:59.000Z

213

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

E-Print Network [OSTI]

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

Stanislav Fisenko; Igor Fisenko

2010-06-27T23:59:59.000Z

214

Accelerator-Based Irradiation Creep of Pyrolytic Carbon Used in TRISO Fuel Particles for the (VHTR) Very Hight Temperature Reactors  

SciTech Connect (OSTI)

Pyrolytic carbon (PyC) is one of the important structural materials in the TRISO fuel particles which will be used in the next generation of gas-cooled very-high-temperature reactors (VHTR). When the TRISO particles are under irradiation at high temperatures, creep of the PyC layers may cause radial cracking leading to catastrophic particle failure. Therefore, a fundamental understanding of the creep behavior of PyC during irradiation is required to predict the overall fuel performance.

Lumin Wang; Gary Was

2010-07-30T23:59:59.000Z

215

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)

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.

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

2005-06-01T23:59:59.000Z

216

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

SciTech Connect (OSTI)

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.

Köck, Eva-Maria; Kogler, Michaela; Pramsoler, Reinhold; Klötzer, 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

217

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

E-Print Network [OSTI]

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

Short, Michael Philip

2010-01-01T23:59:59.000Z

218

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

E-Print Network [OSTI]

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...

Alajo, Ayodeji Babatunde

2011-08-08T23:59:59.000Z

219

PBMR as an Ideal Heat Source for High-Temperature Process Heat Applications  

SciTech Connect (OSTI)

The Pebble Bed Modular Reactor (PBMR) is an advanced helium-cooled, graphite-moderated High Temperature Gas-cooled Reactor (HTGR). A 400 MWt PBMR Demonstration Power Plant (DPP) for the production of electricity is being developed in South Africa. This PBMR technology is also an ideal heat source for process heat applications, including Steam Methane Reforming, steam for Oil Sands bitumen recovery, Hydrogen Production and co-generation (process heat and/or electricity and/or process steam) for petrochemical industries. The cycle configuration used to transport the heat of the reactor to the process plant or to convert the reactor's heat into electricity or steam directly influences the cycle efficiency and plant economics. The choice of cycle configuration depends on the process requirements and is influenced by practical considerations, component and material limitations, maintenance, controllability, safety, performance, risk and cost. This paper provides an overview of the use of a PBMR reactor for process applications and possible cycle configurations are presented for applications which require high temperature process heat and/or electricity. (authors)

Correia, Michael; Greyvenstein, Renee [PBMR - Pty Ltd., 1279 Mike Crawford Avenue, Centurion, 0046 (South Africa); Silady, Fred; Penfield, Scott [Technology Insights, 6540 Lusk Blvd, Suite C-102, San Diego, California 92121 (United States)

2006-07-01T23:59:59.000Z

220

Safety and core design of large liquid-metal cooled fast breeder reactors  

E-Print Network [OSTI]

fission gas plenum212 Conventional fast reactor core designGUPTA. “A Compact Gas-Cooled Fast Reactor with an Ultra-Longbreed and burn gas-cooled fast reactor”. Ph.D. Thesis. MIT,

Qvist, Staffan Alexander

2013-01-01T23:59:59.000Z

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


221

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

222

High temperature ceramic membrane reactors for coal liquid upgrading. Quarterly report No. 10, December 21, 1991--March 20, 1992  

SciTech Connect (OSTI)

In this project we will study a novel process concept, i.e., the use of ceramic membrane reactors in upgrading of coal model compounds and coal derived liquids. In general terms, the USC research team is responsible for constructing and operating the membrane reactor apparatus and for testing various inorganic membranes for the upgrading of coal derived asphaltenes and coal model compounds. The USC effort will involve the principal investigator of this project and two graduate research assistants. The ALCOA team is responsible for the preparation of the inorganic membranes, for construction and testing of the ceramic membrane modules, and for measurement of their transport properties. The ALCOA research effort will involve Dr. Paul K. T. Liu, who is the project manager of the ALCOA research team, an engineer and a technician. UNOCAL`s contribution will be limited to overall technical assistance in catalyst preparation and the operation of the laboratory upgrading membrane reactor and for analytical back-up and expertise in oil analysis and materials characterization. UNOCAL is a no-cost contractor but will be involved in all aspects of the project, as deemed appropriate.

Tsotsis, T.T.

1992-07-01T23:59:59.000Z

223

Novel, fiber optic, hybrid pressure and temperature sensor designed for high-temperature gen-IV reactor applications  

SciTech Connect (OSTI)

A novel, fiber optic, hybrid pressure-temperature sensor is presented. The sensor is designed for reliable operation up to 1050 C, and is based on the high-temperature fiber optic sensors already demonstrated during previous work. The novelty of the sensors presented here lies in the fact that pressure and temperature are measured simultaneously with a single fiber and a single transducer. This hybrid approach will enable highly accurate active temperature compensation and sensor self-diagnostics not possible with other platforms. Hybrid pressure and temperature sensors were calibrated by varying both pressure and temperature. Implementing active temperature compensation resulted in a ten-fold reduction in the temperature-dependence of the pressure measurement. Sensors were also tested for operability in a relatively high neutron radiation environment up to 6.9x10{sup 17} n/cm{sup 2}. In addition to harsh environment survivability, fiber optic sensors offer a number of intrinsic advantages for nuclear power applications including small size, immunity to electromagnetic interference, self diagnostics / prognostics, and smart sensor capability. Deploying fiber optic sensors on future nuclear power plant designs would provide a substantial improvement in system health monitoring and safety instrumentation. Additional development is needed, however, before these advantages can be realized. This paper will highlight recent demonstrations of fiber optic sensors in environments relevant to emerging nuclear power plants. Successes and lessons learned will be highlighted. (authors)

Palmer, M. E.; Fielder, R. S.; Davis, M. A. [Luna Innovations, Incorporated, 2851 Commerce St., Blacksburg, VA 24060 (United States)

2006-07-01T23:59:59.000Z

224

The Modular Helium Reactor for Hydrogen Production  

SciTech Connect (OSTI)

For electricity and hydrogen production, an advanced reactor technology receiving considerable international interest is a modular, passively-safe version of the high-temperature, gas-cooled reactor (HTGR), known in the U.S. as the Modular Helium Reactor (MHR), which operates at a power level of 600 MW(t). For hydrogen production, the concept is referred to as the H2-MHR. Two concepts that make direct use of the MHR high-temperature process heat are being investigated in order to improve the efficiency and economics of hydrogen production. The first concept involves coupling the MHR to the Sulfur-Iodine (SI) thermochemical water splitting process and is referred to as the SI-Based H2-MHR. The second concept involves coupling the MHR to high-temperature electrolysis (HTE) and is referred to as the HTE-Based H2-MHR.

E. Harvego; M. Richards; A. Shenoy; K. Schultz; L. Brown; M. Fukuie

2006-10-01T23:59:59.000Z

225

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

SciTech Connect (OSTI)

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.

John D. Bess; Nozomu Fujimoto

2014-11-01T23:59:59.000Z

226

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)

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.

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

227

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

SciTech Connect (OSTI)

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%.

Ronald G. Ballinger Chunyun Wang Andrew Kadak Neil Todreas

2004-08-30T23:59:59.000Z

228

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

SciTech Connect (OSTI)

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

Simon Phillpot; James Tulenko

2011-09-08T23:59:59.000Z

229

Buffer-Gas Cooled Bose-Einstein Condensate  

E-Print Network [OSTI]

We report the creation of a Bose-Einstein condensate using buffer-gas cooling, the first realization of Bose-Einstein condensation using a broadly general method which relies neither on laser cooling nor unique atom-surface ...

Ketterle, Wolfgang

230

Evaluating the income and employment impacts of gas cooling technologies  

SciTech Connect (OSTI)

The purpose of this study is to estimate the potential employment and income benefits of the emerging market for gas cooling products. The emphasis here is on exports because that is the major opportunity for the U.S. heating, ventilating, and air-conditioning (HVAC) industry. But domestic markets are also important and considered here because without a significant domestic market, it is unlikely that the plant investments, jobs, and income associated with gas cooling exports would be retained within the United States. The prospects for significant gas cooling exports appear promising for a variety of reasons. There is an expanding need for cooling in the developing world, natural gas is widely available, electric infrastructures are over-stressed in many areas, and the cost of building new gas infrastructure is modest compared to the cost of new electric infrastructure. Global gas cooling competition is currently limited, with Japanese and U.S. companies, and their foreign business partners, the only product sources. U.S. manufacturers of HVAC products are well positioned to compete globally, and are already one of the faster growing goods-exporting sectors of the U.S. economy. Net HVAC exports grew by over 800 percent from 1987 to 1992 and currently exceed $2.6 billion annually (ARI 1994). Net gas cooling job and income creation are estimated using an economic input-output model to compare a reference case to a gas cooling scenario. The reference case reflects current policies, practices, and trends with respect to conventional electric cooling technologies. The gas cooling scenario examines the impact of accelerated use of natural gas cooling technologies here and abroad.

Hughes, P.J. [Oak Ridge National Lab., TN (United States); Laitner, S.

1995-03-01T23:59:59.000Z

231

Understanding Fundamental Material Degradation Processes in High Temperature Aggressive Chemomechanical Environments  

SciTech Connect (OSTI)

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 Next–Generation 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.

Stubbins, James; Gewirth, Andrew; Sehitoglu, Huseyin; Sofronis, Petros; Robertson, Ian

2014-01-16T23:59:59.000Z

232

Decommissioning the UHTREX Reactor Facility at Los Alamos, New Mexico  

SciTech Connect (OSTI)

The Ultra-High Temperature Reactor Experiment (UHTREX) facility was constructed in the late 1960s to advance high-temperature and gas-cooled reactor technology. The 3-MW reactor was graphite moderated and helium cooled and used 93% enriched uranium as its fuel. The reactor was run for approximately one year and was shut down in February 1970. The decommissioning of the facility involved removing the reactor and its associated components. This document details planning for the decommissioning operations which included characterizing the facility, estimating the costs of decommissioning, preparing environmental documentation, establishing a system to track costs and work progress, and preplanning to correct health and safety concerns in the facility. Work to decommission the facility began in 1988 and was completed in September 1990 at a cost of $2.9 million. The facility was released to Department of Energy for other uses in its Los Alamos program.

Salazar, M.; Elder, J.

1992-08-01T23:59:59.000Z

233

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

SciTech Connect (OSTI)

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

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

2010-03-01T23:59:59.000Z

234

System Evaluation and Economic Analysis of a HTGR Powered High-Temperature Electrolysis Hydrogen Production Plant  

SciTech Connect (OSTI)

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 322°C and 750°C, 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.

Michael G. McKellar; Edwin A. Harvego; Anastasia A. Gandrik

2010-10-01T23:59:59.000Z

235

GAS COOLED ELECTRICAL LEADS FOR USE ON FORCED COOLED SUPERCONDUCTING MAGNETS  

E-Print Network [OSTI]

11-14, 1981 GAS COOLED ELECTRICAL LEADS FOR USE ON FORCEDim mumii P mm GAS COOLED ELECTRICAL LEADS FOR USE ON FORCEDD. Henning, "Cryogenic Electrical Leads," Proceedings of the

Smits, R.G.

2010-01-01T23:59:59.000Z

236

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

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

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

237

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

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

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

238

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

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

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

239

High Temperature Heat Exchanger Project  

SciTech Connect (OSTI)

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

Anthony E. Hechanova, Ph.D.

2008-09-30T23:59:59.000Z

240

High-Temperature Superconductivity  

ScienceCinema (OSTI)

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

Peter Johnson

2010-01-08T23:59:59.000Z

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


241

High temperature pressure gauge  

DOE Patents [OSTI]

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.

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

1981-01-01T23:59:59.000Z

242

MHTGR: New production reactor summary of experience base  

SciTech Connect (OSTI)

Worldwide interest in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) stems from the capability of the system to retain the advanced fuel and thermal performance while providing unparalleled levels of safety. The small power level of the MHTGR and its passive systems give it a margin of safety not attained by other concepts being developed for power generation. This report covers the experience base for the key nuclear system, components, and processes related to the MHTGR-NPR. 9 refs., 39 figs., 9 tabs.

Not Available

1988-03-01T23:59:59.000Z

243

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)

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.

Scheele, Randall D.; Casella, Andrew M.

2010-09-28T23:59:59.000Z

244

Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 4: High-Temperature Materials PIRTs  

SciTech Connect (OSTI)

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.

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

245

High-temperature low-cycle fatigue and tensile properties of Hastelloy X and alloy 617 in air and HTGR-helium  

SciTech Connect (OSTI)

Results of strain controlled fatigue and tensile tests are presented for two nickel base solution hardened alloys which are reference structural alloys for use in several high temperature gas cooled reactor concepts. These alloys, Hastelloy X Inconel 617, were tested at temperatures ranging from room temperature to 871/sup 0/C in air and impure helium. Materials were tested in the solution annealed as well as in the pre-aged condition where aging consisted of isothermal exposure at one of several temperatures for periods of up to 20,000 h. Comparisons are also given between the strain controlled fatigue lives of these alloys and several other commonly used alloys all tested at 538/sup 0/C.

Strizak, J.P.; Brinkman, C.R.; Rittenhouse, P.L.

1981-01-01T23:59:59.000Z

246

NSTX High Temperature Sensor Systems  

SciTech Connect (OSTI)

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.

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

1999-11-01T23:59:59.000Z

247

PROCESS MODEL FOR THE PRODUCTION OF SYNGAS VIA HIGH TEMPERATURE CO-ELECTROLYSIS  

SciTech Connect (OSTI)

A process model has been developed to evaluate the potential performance of a large-scale high-temperature coelectrolysis plant for the production of syngas from steam and carbon dioxide. The coelectrolysis process allows for direct electrochemical reduction of the steam – carbon dioxide gas mixture, yielding hydrogen and carbon monoxide, or syngas. The process model has been developed using the HYSYS systems analysis code. Using this code, a detailed process flowsheet has been defined that includes all the components that would be present in an actual plant such as pumps, compressors, heat exchangers, turbines, and the electrolyzer. Since the electrolyzer is not a standard HYSYS component, a custom one-dimensional coelectrolysis model was developed for incorporation into the overall HYSYS process flowsheet. The 1-D coelectrolysis model assumes local chemical equilibrium among the four process-gas species via the shift reaction. The electrolyzer model allows for the determination of coelectrolysis outlet temperature, composition (anode and cathode sides), mean Nernst potential, operating voltage and electrolyzer power based on specified inlet gas flow rates, heat loss or gain, current density, and cell area-specific resistance. The one-dimensional electrolyzer model was validated by comparison with results obtained from a fully 3-D computational fluid dynamics model developed using FLUENT, and by comparison to experimental data. This paper provides representative results obtained from the HYSYS flowsheet model for a 300 MW coelectrolysis plant, coupled to a high-temperature gas-cooled nuclear reactor. The coelectrolysis process, coupled to a nuclear reactor, provides a means of recycling carbon dioxide back into a useful liquid fuel. If the carbon dioxide source is based on biomass, the entire process would be climate neutral.

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

2007-11-01T23:59:59.000Z

248

Parametric Study Of Large-Scale Production Of Syngas Via High Temperature Co-Electrolysis  

SciTech Connect (OSTI)

A process model has been developed to evaluate the potential performance of a largescale high-temperature co-electrolysis plant for the production of syngas from steam and carbon dioxide. The co-electrolysis process allows for direct electrochemical reduction of the steam – carbon dioxide gas mixture, yielding hydrogen and carbon monoxide, or syngas. The process model has been developed using the Honeywell UniSim systems analysis code. Using this code, a detailed process flow sheet has been defined that includes all the components that would be present in an actual plant such as pumps, compressors, heat exchangers, turbines, and the electrolyzer. Since the electrolyzer is not a standard UniSim component, a custom one-dimensional co-electrolysis model was developed for incorporation into the overall UniSim process flow sheet. The one dimensional co-electrolysis model assumes local chemical equilibrium among the four process-gas species via the gas shift reaction. The electrolyzer model allows for the determination of co-electrolysis outlet temperature, composition (anode and cathode sides); mean Nernst potential, operating voltage and electrolyzer power based on specified inlet gas flow rates, heat loss or gain, current density, and cell area-specific resistance. The one-dimensional electrolyzer model was validated by comparison with results obtained from a fully three dimensional computational fluid dynamics model developed using FLUENT, and by comparison to experimental data. This paper provides representative results obtained from the UniSim flow sheet model for a 300 MW co-electrolysis plant, coupled to a high-temperature gas-cooled nuclear reactor. The coelectrolysis process, coupled to a nuclear reactor, provides a means of recycling carbon dioxide back into a useful liquid fuel. If the carbon dioxide source is based on biomass, the overall process, from production through utilization, would be climate neutral.

J. E. O'Brien; M. G. McKellar; C. M. Stoots; J. S. Herring; G. L. Hawkes

2007-11-01T23:59:59.000Z

249

Investigation of the Feasibility of a Small Scale Transmutation Device  

E-Print Network [OSTI]

27], and the Gas-Cooled Fast Transmutation Reactor (GCFTR)fast breeder reactors (LMFBRs) and high temperature gas

Sit, Roger Carson

2009-01-01T23:59:59.000Z

250

Thermal hydraulic performance testing of printed circuit heat exchangers in a high-temperature helium test facility  

SciTech Connect (OSTI)

In high-temperature gas-cooled reactors, such as a very high temperature reactor (VHTR), an intermediate heat exchanger (IHX) is required to efficiently transfer the core thermal output to a secondary fluid for electricity generation with an indirect power cycle and/or process heat applications. Currently, there is no proven high-temperature (750–800 °C or higher) compact heat exchanger technology for high-temperature reactor design concepts. In this study, printed circuit heat exchanger (PCHE), a potential IHX concept for high-temperature applications, has been investigated for their heat transfer and pressure drop characteristics under high operating temperatures and pressures. Two PCHEs, each having 10 hot and 10 cold plates with 12 channels (semicircular cross-section) in each plate are fabricated using Alloy 617 plates and tested for their performance in a high-temperature helium test facility (HTHF). The PCHE inlet temperature and pressure were varied from 85 to 390 °C/1.0–2.7 MPa for the cold side and 208–790 °C/1.0–2.7 MPa for the hot side, respectively, while the mass flow rate of helium was varied from 15 to 49 kg/h. This range of mass flow rates corresponds to PCHE channel Reynolds numbers of 950 to 4100 for the cold side and 900 to 3900 for the hot side (corresponding to the laminar and laminar-to-turbulent transition flow regimes). The obtained experimental data have been analyzed for the pressure drop and heat transfer characteristics of the heat transfer surface of the PCHEs and compared with the available models and correlations in the literature. In addition, a numerical treatment of hydrodynamically developing and hydrodynamically fully-developed laminar flow through a semicircular duct is presented. Relations developed for determining the hydrodynamic entrance length in a semicircular duct and the friction factor (or pressure drop) in the hydrodynamic entry length region for laminar flow through a semicircular duct are given. Various hydrodynamic entrance region parameters, such as incremental pressure drop number, apparent Fanning friction factor, and hydrodynamic entrance length in a semicircular duct have been numerically estimated.

Sai K. Mylavarapu; Xiaodong Sun; Richard E. Glosup; Richard N. Christensen; Michael W. Patterson

2014-04-01T23:59:59.000Z

251

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...

252

Regulatory Safety Issues in the Structural Design Criteria of ASME Section III Subsection NH and for Very High Temperatures for VHTR & GEN IV  

SciTech Connect (OSTI)

The objective of this task is to identify issues relevant to ASME Section III, Subsection NH [1], and related Code Cases that must be resolved for licensing purposes for VHTGRs (Very High Temperature Gas Reactor concepts such as those of PBMR, Areva, and GA); and to identify the material models, design criteria, and analysis methods that need to be added to the ASME Code to cover the unresolved safety issues. Subsection NH was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. This Report describes: (1) NRC and ACRS safety concerns raised during the licensing process of CRBR , (2) how some of these issues are addressed by the current Subsection NH of the ASME Code; and (3) the material models, design criteria, and analysis methods that need to be added to the ASME Code and Code Cases to cover unresolved regulatory issues for very high temperature service.

William J. O’Donnell; Donald S. Griffin

2007-05-07T23:59:59.000Z

253

Reactor safety method  

DOE Patents [OSTI]

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

Vachon, Lawrence J. (Clairton, PA)

1980-03-11T23:59:59.000Z

254

High temperature probe  

DOE Patents [OSTI]

A high temperature probe for sampling, for example, smokestack fumes, and is able to withstand temperatures of 3000.degree. F. The probe is constructed so as to prevent leakage via the seal by placing the seal inside the water jacket whereby the seal is not exposed to high temperature, which destroys the seal. The sample inlet of the probe is also provided with cooling fins about the area of the seal to provide additional cooling to prevent the seal from being destroyed. Also, a heated jacket is provided for maintaining the temperature of the gas being tested as it passes through the probe. The probe includes pressure sensing means for determining the flow velocity of an efficient being sampled. In addition, thermocouples are located in various places on the probe to monitor the temperature of the gas passing there through.

Swan, Raymond A. (Fremont, CA)

1994-01-01T23:59:59.000Z

255

High Temperature Capacitor Development  

SciTech Connect (OSTI)

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.

John Kosek

2009-06-30T23:59:59.000Z

256

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.

257

Evaluation Metrics for Intermediate Heat Exchangers for Next Generation Nuclear Reactors  

SciTech Connect (OSTI)

The Department of Energy (DOE) is working with industry to develop a next generation, high-temperature gas-cooled reactor (HTGR) as a part of the effort to supply the United States with abundant, clean, and secure energy as initiated by the Energy Policy Act of 2005 (EPAct; Public Law 109-58,2005). The NGNP Project, led by the Idaho National Laboratory (INL), will demonstrate the ability of the HTGR to generate hydrogen, electricity, and/or high-quality process heat for a wide range of industrial applications.

Piyush Sabharwall; Eung Soo Kim; Nolan Anderson

2011-06-01T23:59:59.000Z

258

Small-Particle Solar Receiver for High-Temperature Brayton Power Cycles  

Broader source: Energy.gov [DOE]

This fact sheet describes a scattering solar thermal concentrators project awarded under the DOE's 2012 SunShot Concentrating Solar Power R&D award program. The team, led by San Diego State University, is working to validate, through on-sun testing, the viability of the Small Particle Heat Exchange Receiver concept. If successful, this project team would build the first large-scale, pressurized, high-temperature, gas-cooled solar receiver capable of being deployed commercially.

259

High temperature thermometric phosphors  

DOE Patents [OSTI]

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.

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

1999-03-23T23:59:59.000Z

260

High temperature thermometric phosphors  

DOE Patents [OSTI]

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.

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

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


261

Reactor physics design of supercritical CO?-cooled fast reactors  

E-Print Network [OSTI]

Gas-Cooled Fast Reactors (GFRs) are among the GEN-IV designs proposed for future deployment. Driven by anticipated plant cost reduction, the use of supercritical CO? (S-CO?) as a Brayton cycle working fluid in a direct ...

Pope, Michael A. (Michael Alexander)

2004-01-01T23:59:59.000Z

262

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

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

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

263

Testing of Performance of Optical Fibers Under Irradiation in Intense Radiation Fields, When Subjected to Very High Temperatures  

SciTech Connect (OSTI)

The primary objective of this project is to measure and model the performance of optical fibers in intense radiation fields when subjected to very high temperatures. This research will pave the way for fiber optic and optically based sensors under conditions expected in future high-temperature gas-cooled reactors. Sensor life and signal-to-noise ratios are susceptible to attenuation of the light signal due to scattering and absorbance in the fibers. This project will provide an experimental and theoretical study of the darkening of optical fibers in high-radiation and high-temperature environments. Although optical fibers have been studied for moderate radiation fluence and flux levels, the results of irradiation at very high temperatures have not been published for extended in-core exposures. Several previous multi-scale modeling efforts have studied irradiation effects on the mechanical properties of materials. However, model-based prediction of irradiation-induced changes in silica�s optical transport properties has only recently started to receive attention due to possible applications as optical transmission components in fusion reactors. Nearly all damage-modeling studies have been performed in the molecular-dynamics domain, limited to very short times and small systems. Extended-time modeling, however, is crucial to predicting the long-term effects of irradiation at high temperatures, since the experimental testing may not encompass the displacement rate that the fibers will encounter if they are deployed in the VHTR. The project team will pursue such extended-time modeling, including the effects of the ambient and recrystallization. The process will be based on kinetic MC modeling using the concept of amorphous material consisting of building blocks of defect-pairs or clusters, which has been successfully applied to kinetic modeling in amorphized and recrystallized silicon. Using this procedure, the team will model compensation for rate effects, and the interplay of rate effects with the effects of annealing, to accurately predict the fibers� reliability and expected lifetime

Thomas Blue; Wolfgang Windl; Bryan Dickerson

2013-01-03T23:59:59.000Z

264

High temperature interfacial superconductivity  

DOE Patents [OSTI]

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.

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

2012-06-19T23:59:59.000Z

265

A vectorized heat transfer model for solid reactor cores  

SciTech Connect (OSTI)

The new generation of nuclear reactors includes designs that are significantly different from light water reactors. Among these new reactor designs is the Modular High-Temperature Gas-Cooled Reactor (MHTGR). In addition, nuclear thermal rockets share a number of similarities with terrestrial HTGRs and would be amenable to similar types of analyses. In these reactors, the heat transfer in the solid core mass is of primary interest in design and safety assessment. One significant safety feature of these reactors is the capability to withstand a loss of pressure and forced cooling in the primary system and still maintain peak fuel temperatures below the safe threshold for retaining the fission products. To accurately assess the performance of gas-cooled reactors during these types of transients, a Helium/Hydrogen Cooled Reactor Analysis (HERA) computer code has been developed. HERA has the ability to model arbitrary geometries in three dimensions, which allows the user to easily analyze reactor cores constructed of prismatic graphite elements. The code accounts for heat generation in the fuel, control rods and other structures; conduction and radiation across gaps; convection to the coolant; and a variety of boundary conditions. The numerical solution scheme has been optimized for vector computers, making long transient analyses economical. Time integration is either explicit or implicit, which allows the use of the model to accurately calculate both short- or long-term transients with an efficient use of computer time. Both the basic spatial and temporal integration schemes have been benchmarked against analytical solutions. Also, HERA has been used to analyze a depressurized loss of forced cooling transient in a HTGR with a very detailed three-dimensional input model. The results compare favorably with other means of analysis and provide further validation of the models and methods. 18 refs., 11 figs.

Rider, W.J.; Cappiello, M.W.; Liles, D.R.

1990-01-01T23:59:59.000Z

266

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...

267

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

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

ARIES Program Collection: Plasma Physics and Fusion 37 HIGH EFFICIENCY GENERATION OF HYDROGEN FUELS Summary: , Heavy metal-cooled, Gas-cooled, Molten salt-cooled, Liquid- core and...

268

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

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

Report Detecting Fractures Using Technology at High Temperatures and Depths - Geothermal Ultrasonic Fracture Imager (GUFI); 2010 Geothermal Technology Program Peer Review Report...

269

Safety Issues for High Temperature Gas Reactors  

E-Print Network [OSTI]

Consequences AOO AC SPC Challenges DESIGN BASIS * Severe challenge to the Fission Products Confinement Function Significant Accident Sequences ­ Air Ingress ­ Water Ingress (reactivity insertion) ­ Seismic Events effects and chemical attack on graphite · Blow down loads and timing of accident event sequences

270

High Temperature Processing Symposium 2014  

E-Print Network [OSTI]

} High temperature recycling operations } Materials sustainability } New furnace technology (including solar) We look forward to seeing you in February 2014. Dr M Akbar Rhamdhani (Chairman HTPS 2014) Prof

Liley, David

271

Philosophy 26 High Temperature Superconductivity  

E-Print Network [OSTI]

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

Callender, Craig

272

High temperature Seebeck coefficient metrology  

SciTech Connect (OSTI)

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.

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

273

High temperature Seebeck coefficient metrology  

SciTech Connect (OSTI)

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.

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

2010-01-01T23:59:59.000Z

274

High temperature superconductor current leads  

DOE Patents [OSTI]

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.

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

1995-06-20T23:59:59.000Z

275

Microsoft Word - inl-ext-10-20460.doc  

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

evaluation for the technology readiness of the interface components that are required to transfer high-temperature heat from a High Temperature Gas-cooled Reactor (HTGR) to...

276

High temperature turbine engine structure  

DOE Patents [OSTI]

A high temperature turbine engine includes a rotor portion having axially stacked adjacent ceramic rotor parts. A ceramic/ceramic joint structure transmits torque between the rotor parts while maintaining coaxial alignment and axially spaced mutually parallel relation thereof despite thermal and centrifugal cycling.

Boyd, Gary L. (Tempe, AZ)

1991-01-01T23:59:59.000Z

277

High temperature lightweight foamed cements  

DOE Patents [OSTI]

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.

Sugama, Toshifumi.

1989-10-03T23:59:59.000Z

278

High temperature lightweight foamed cements  

DOE Patents [OSTI]

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.

Sugama, Toshifumi (Mastic Beach, NY)

1989-01-01T23:59:59.000Z

279

HIGH TEMPERATURE GEOTHERMAL RESERVOIR ENGINEERING  

E-Print Network [OSTI]

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.

Schroeder, R.C.

2009-01-01T23:59:59.000Z

280

Experiment Hazard Class 3 - High Temperatures  

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

Operation * APS Base Low Temperatures * Cryogenic Systems High Temperatures * Electric Furnace * Optical Furnace * Other High Temperature Lasers * Laser, Class 2 * Laser,...

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


281

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...

282

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...

283

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.

284

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...

285

An evaluation of the fast-mixed spectrum reactor  

E-Print Network [OSTI]

An independent evaluation of the neutronic characteristics of a gas-cooled fast-mixed spectrum reactor (FMSR) core design has been performed. A benchmark core configuration for an early FMSR design was provided by Brookhaven ...

Loh, Wee Tee

1980-01-01T23:59:59.000Z

286

High temperature structural insulating material  

DOE Patents [OSTI]

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.

Chen, W.Y.

1984-07-27T23:59:59.000Z

287

High temperature structural insulating material  

DOE Patents [OSTI]

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.

Chen, Wayne Y. (Munster, IN)

1987-01-01T23:59:59.000Z

288

High temperature turbine engine structure  

DOE Patents [OSTI]

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.

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

1993-01-01T23:59:59.000Z

289

High temperature turbine engine structure  

DOE Patents [OSTI]

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.

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

1992-01-01T23:59:59.000Z

290

High temperature turbine engine structure  

DOE Patents [OSTI]

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.

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

1994-01-01T23:59:59.000Z

291

High Temperature Test Laboratory Accomplishments  

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

Knudson, K. Condie, and B. Sencer, "In-situ Creep Testing Capability for the Advanced Test Reactor," Nuclear Technology, 179, 3, September 2012, pp 413-428. B. Geslot, T. Unruh,...

292

advanced gas-cooled reactor: Topics by E-print Network  

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

. . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

293

advanced gas-cooled reactors: Topics by E-print Network  

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

. . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

294

Gas-cooled fast breeder reactor fuel element thermal-hydraulic investigations : final report  

E-Print Network [OSTI]

Experimental and analytical work was performed to determine the influence of rod surface roughening on the thermal-hydraulic behavior of rod array type, nuclear fuel elements. Experimental data was obtained using a ...

Eaton, Thomas Eldon

1975-01-01T23:59:59.000Z

295

E-Print Network 3.0 - advanced-gas-cooled-nuclear-reactor materials...  

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

for all job... with specially handled documents Physical processing of materialsbook marking Collect & process ... Source: Purdue University, Combinatorial Scientific...

296

Design of compact intermediate heat exchangers for gas cooled fast reactors  

E-Print Network [OSTI]

Two aspects of an intermediate heat exchanger (IHX) for GFR service have been investigated: (1) the intrinsic characteristics of the proposed compact printed circuit heat exchanger (PCHE); and (2) a specific design optimizing ...

Gezelius, Knut, 1978-

2004-01-01T23:59:59.000Z

297

Microsoft Word - PLN-4086 Rev 1.docx  

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

Test Reactor IAEA International Atomic Energy Agency MHTGR modular high temperature gas-cooled reactor NGNP Next Generation Nuclear Plant PSID preliminary safety information...

298

High-temperature thermocouples and related methods  

DOE Patents [OSTI]

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.

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

299

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

300

High Temperature Superconducting Underground Cable  

SciTech Connect (OSTI)

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 world’s 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.

Farrell, Roger, A.

2010-02-28T23:59:59.000Z

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


301

Considerations Associated with Reactor Technology Selection for the Next Generation Nuclear Plant Project  

SciTech Connect (OSTI)

At the inception of the Next Generation Nuclear Plant Project and during predecessor activities, alternative reactor technologies have been evaluated to determine the technology that best fulfills the functional and performance requirements of the targeted energy applications and market. Unlike the case of electric power generation where the reactor performance is primarily expressed in terms of economics, the targeted energy applications involve industrial applications that have specific needs in terms of acceptable heat transport fluids and the associated thermodynamic conditions. Hence, to be of interest to these industrial energy applications, the alternative reactor technologies are weighed in terms of the reactor coolant/heat transport fluid, achievable reactor outlet temperature, and practicality of operations to achieve the very high reliability demands associated with the petrochemical, petroleum, metals and related industries. These evaluations have concluded that the high temperature gas-cooled reactor (HTGR) can uniquely provide the required ranges of energy needs for these target applications, do so with promising economics, and can be commercialized with reasonable development risk in the time frames of current industry interest – i.e., within the next 10-15 years.

L.E. Demick

2010-09-01T23:59:59.000Z

302

Deposition method for producing silicon carbide high-temperature semiconductors  

DOE Patents [OSTI]

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.

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

1987-01-01T23:59:59.000Z

303

Implementation of vented fuel assemblies in the supercritical CO?-cooled fast reactor  

E-Print Network [OSTI]

Analysis has been undertaken to investigate the utilization of fuel assembly venting in the reference design of the gas-cooled fast reactor under study as part of the larger research effort at MIT under Gen-IV NERI Project ...

McKee, Stephanie A

2008-01-01T23:59:59.000Z

304

Optimized core design of a supercritical carbon dioxide-cooled fast reactor  

E-Print Network [OSTI]

Spurred by the renewed interest in nuclear power, Gas-cooled Fast Reactors (GFRs) have received increasing attention in the past decade. Motivated by the goals of the Generation-IV International Forum (GIF), a GFR cooled ...

Handwerk, Christopher S. (Christopher Stanley), 1974-

2007-01-01T23:59:59.000Z

305

High Temperature Materials Interim Data Qualification Report FY 2011  

SciTech Connect (OSTI)

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.

Nancy Lybeck

2011-08-01T23:59:59.000Z

306

Electronic Applications of High Temperature Superconductors  

E-Print Network [OSTI]

ELECfRONIC APPLICAnONS OF HIGH TEMPERATURE SUPERCONDUCTORS HARRY KROGER and ROBERT F. MIRACKY Superconductivity Program MCC Austin, Texas ABSTRACT The possible uses of high temperature superconductors in electronics applications... attempts a sober appraisal of the potential ap plications of high temperature superconductors to electronics. Al though we believe that these applications are very promising, and in some sense unlimited, we offer here an opinion which runs contrary...

Kroger, H.; Miracky, R. F.

307

Deep Trek High Temperature Electronics Project  

SciTech Connect (OSTI)

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.

Bruce Ohme

2007-07-31T23:59:59.000Z

308

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...

309

Intertwined Orders in High Temperature Superconductors  

E-Print Network [OSTI]

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

Ostoja-Starzewski, Martin

310

Quantitative Modeling of High Temperature Magnetization Dynamics  

SciTech Connect (OSTI)

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

Zhang, Shufeng

2009-03-01T23:59:59.000Z

311

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...

312

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...

313

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...

314

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...

315

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...

316

Photonic crystals for high temperature applications  

E-Print Network [OSTI]

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 ...

Yeng, Yi Xiang

2014-01-01T23:59:59.000Z

317

Manufacturing Barriers to High Temperature PEM Commercialization...  

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

Barriers to High Temperature PEM Commercialization Presented at the NREL Hydrogen and Fuel Cell Manufacturing R&D Workshop in Washington, DC, August 11-12, 2011....

318

High temperature superconducting fault current limiter  

DOE Patents [OSTI]

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.

Hull, J.R.

1997-02-04T23:59:59.000Z

319

High Temperature Solar Splitting of Methane  

E-Print Network [OSTI]

-term commercialization opportunities #12;Why Use Solar Energy?Why Use Solar Energy? · High concentrations possible (>1000High Temperature Solar Splitting of Methane to Hydrogen and Carbon High Temperature Solar Splitting and worldwide) ­ Sufficient to power the world (if we choose to) · Advantages tradeoff against collection area

320

Thermal hydraulic design of a 2400 MW t?h? direct supercritical CO?-cooled fast reactor  

E-Print Network [OSTI]

The gas cooled fast reactor (GFR) has received new attention as one of the basic concepts selected by the Generation-IV International Forum (GIF) for further investigation. Currently, the reference GFR is a helium-cooled ...

Pope, Michael A. (Michael Alexander)

2006-01-01T23:59:59.000Z

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


321

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

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

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

322

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

E-Print Network [OSTI]

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

California at San Diego, University of

323

Liquid Fuel Production from Biomass via High Temperature Steam Electrolysis  

SciTech Connect (OSTI)

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.

Grant L. Hawkes; Michael G. McKellar

2009-11-01T23:59:59.000Z

324

Microsoft Word - MPR HTGRs for US H2.doc  

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

U.S. Hydrogen Production Purpose This evaluation estimates how many High Temperature Gas-cooled Reactors (HTGRs) would be needed to provide energy for different aspects of...

325

Microsoft Word - NGNP_Licensing Study Report_4-30-2008-FINAL...  

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

Company LLC 8 of 55 Abbreviation or Acronym Definition MHTGR Modular High Temperature Gas-Cooled Reactor NEPA National Environmental Policy Act NGNP Next Generation Nuclear...

326

Microsoft Word - TEV-962 Rev 0.docx  

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

via Steam Methane Reforming to High Temperature Gas-Cooled Reactor Outlet Temperature Economic Analysis Form 412.09 (Rev. 10) Idaho National Laboratory SENSITIVITY OF HYDROGEN...

327

Microsoft Word - CTF WBS&PEDS_white paper_v1.doc  

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

WHITE PAPER Work Breakdown Structure and PlantEquipment Designation System Numbering Scheme for the High Temperature Gas- Cooled Reactor (HTGR) Component Test Capability (CTC)...

328

NGNP-NHS 100-RXBLDG 0 Final 091608.doc  

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

Production System HSS Helium Service System HTGR High Temperature Gas Cooled Reactor HVAC Heating, Ventilation, and Air Conditioning IBC International Building Code ICM Interim...

329

Microsoft Word - NGNP-PLD-GEN-RPT-N-00007_Rev0.doc  

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

Pressure Boundary HPS Helium Purification System HTGR High Temperature Gas-cooled Reactor HVAC Heating, Ventilation, Air Conditioning ICDS In-core Distribution System IFI Initial...

330

12-9152036-000 Pebble Bed Design Requirements 01_05_11  

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

new requirements. 3.0 REQUIREMENT SELECTION PROCESSCRITERIA The NGNP System Requirements Manual 2 and Key Design Requirements for the High Temperature Gas- cooled Reactor...

331

Microsoft Word - TEV-961 Rev 0.docx  

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

Evaluation Study Project No. 23843 Sensitivity of Hydrogen Production via Steam Methane Reforming to High Temperature Gas-Cooled Reactor Outlet Temperature Process Analysis...

332

Microsoft Word - TEV 704 Rev 2.docx  

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

Generation Nuclear Plant (NGNP) Project to evaluate the integration of high-temperature gas-cooled reactor (HTGR) technology with conventional chemical processes. This TEV...

333

Microsoft Word - ~4787852.docx  

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

high temperature gas-cooled reactor (HTGR)-integrated processes that recover oil from oil shale, desalinate seawater, and upgrade bitumen to synthetic crude oil. An economic...

334

Microsoft Word - 12-9051191-001-FINALforRecordsMgt.doc  

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

Note to Commissioner Assistants, August 17, 2001, "Summary of the NRC Delegation Visit to Germany on Safety Aspects of High Temperature Gas Cooled Reactor Design Technology,"...

335

Investigations into High Temperature Components and Packaging  

SciTech Connect (OSTI)

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.

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

2007-12-31T23:59:59.000Z

336

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

SciTech Connect (OSTI)

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.

Michael G. McKellar; Edwin A. Harvego

2010-05-01T23:59:59.000Z

337

A new chemodynamical tool to study the evolution of galaxies in the local Universe: a quick and accurate numerical technique to compute gas cooling rate for any chemical composition  

E-Print Network [OSTI]

We have developed a quick and accurate numerical tool to compute gas cooling whichever its chemical composition.

Nicolas Champavert; Hervé Wozniak

2007-03-13T23:59:59.000Z

338

Silicon carbide oxidation in high temperature steam  

E-Print Network [OSTI]

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 ...

Arnold, Ramsey Paul

2011-01-01T23:59:59.000Z

339

Process Heat Exchanger Options for the Advanced High Temperature Reactor  

SciTech Connect (OSTI)

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.

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

2011-06-01T23:59:59.000Z

340

Process Heat Exchanger Options for Fluoride Salt High Temperature Reactor  

SciTech Connect (OSTI)

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.

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

2011-04-01T23:59:59.000Z

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


341

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation ProposedUsing ZirconiaPolicy andExsolutionFES6FY 2011 OIG(SC) 2FY98 ToFuel | ornl.gov

342

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartmentNationalRestart of the Review of theOFFICE OF CIVIL to Mod 0099 datedMod0/%2A2 *31)5)

343

High temperature crystalline superconductors from crystallized glasses  

DOE Patents [OSTI]

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.

Shi, Donglu (Downers Grove, IL)

1992-01-01T23:59:59.000Z

344

Frustrated phase separation and high temperature superconductivity  

SciTech Connect (OSTI)

A dilute system of neutral holes in an antiferromagnet separates into a hole-rich and a hole-poor phase. The phase separation is frustrated by long-range Coulomb interactions but, provided the dielectric constant is sufficiently large, there remain large-amplitude low-energy fluctuations in the hole density at intermediate length scales. The extensive experimental evidence showing that this behavior giver, a reasonable picture of high temperature superconductors is surveyed. Further, it is shown that the scattering of mobile holes from the local density fluctuations may account for the anomalous normal-state properties of high temperature superconductors and also provide the mechanism of pairing.

Emery, V.J. [Brookhaven National Lab., Upton, NY (United States); Kivelson, S.A. [California Univ., Los Angeles, CA (United States). Dept. of Physics

1992-09-01T23:59:59.000Z

345

Frustrated phase separation and high temperature superconductivity  

SciTech Connect (OSTI)

A dilute system of neutral holes in an antiferromagnet separates into a hole-rich and a hole-poor phase. The phase separation is frustrated by long-range Coulomb interactions but, provided the dielectric constant is sufficiently large, there remain large-amplitude low-energy fluctuations in the hole density at intermediate length scales. The extensive experimental evidence showing that this behavior giver, a reasonable picture of high temperature superconductors is surveyed. Further, it is shown that the scattering of mobile holes from the local density fluctuations may account for the anomalous normal-state properties of high temperature superconductors and also provide the mechanism of pairing.

Emery, V.J. (Brookhaven National Lab., Upton, NY (United States)); Kivelson, S.A. (California Univ., Los Angeles, CA (United States). Dept. of Physics)

1992-01-01T23:59:59.000Z

346

Sustained Growth of the Ex Vivo Ablation Zones' Critical Short Axis Using Gas-cooled Radiofrequency Applicators  

SciTech Connect (OSTI)

Purpose: To evaluate the ablation zones created with a gas-cooled bipolar radiofrequency applicator performed on ex vivo bovine liver tissue. Materials and Methods: A total of 320 ablations with an internally gas-cooled bipolar radiofrequency applicator were performed on fresh ex vivo bovine liver tissue, varying the ablation time (5, 10, 15, and 20 min), power (20, 30, 40, and 50 W), and gas pressure of the CO{sub 2} used for cooling (585, 600, 615, 630, 645 psi), leading to a total of 80 different parameter combinations. Size and shape of the white coagulation zone were assessed. Results: The largest complete ablation zone was achieved after 20 min of implementing 50 W and 645 psi, resulting in a short axis of mean 46 {+-} 1 mm and a long axis of 56 {+-} 2 mm (mean {+-} standard deviation). Short-axis diameters increased between 5 and 20 min of ablation time at 585 psi (increase of the short axis was 45% at 30 W, 29% at 40 W, and 39% at 50 W). This increase was larger at 645 psi (113% at 30 W, 67% at 40 W, and 70% at 50 W). Macroscopic assessment and NADH (nicotinamide adenine dinucleotide) staining revealed incompletely ablated tissue along the needle track in 18 parameter combinations including low-power settings (20 and 30 W) and different cooling levels and ablation times. Conclusion: Gas-cooled radiofrequency applicators increase the short-axis diameter of coagulation in an ex vivo setting if appropriate parameters are selected.

Rempp, Hansjoerg, E-mail: hansjoerg.rempp@med.uni-tuebingen.de [Eberhard Karls University of Tuebingen, Department of Diagnostic and Interventional Radiology (Germany); Scharpf, Marcus [Insitute of Pathology, Eberhard Karls University of Tuebingen, Department of General Pathology and Pathological Anatomy (Germany); Voigtlaender, Matthias [ERBE Elektromedizin GmbH (Germany); Schraml, Christina; Schmidt, Diethard [Eberhard Karls University of Tuebingen, Department of Diagnostic and Interventional Radiology (Germany); Fend, Falko [Insitute of Pathology, Eberhard Karls University of Tuebingen, Department of General Pathology and Pathological Anatomy (Germany); Claussen, Claus D. [Eberhard Karls University of Tuebingen, Department of Diagnostic and Interventional Radiology (Germany); Enderle, Markus D. [ERBE Elektromedizin GmbH (Germany); Pereira, Philippe L. [Klinik fuer Radiologie, Minimalinvasive Therapien und Nuklearmedizin (Germany); Clasen, Stephan [Eberhard Karls University of Tuebingen, Department of Diagnostic and Interventional Radiology (Germany)

2011-02-15T23:59:59.000Z

347

E-Print Network 3.0 - advanced gas cooled graphite moderated...  

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

at Chicago's Stagg Field and the X-IO graphite reactor at Oak Ridge). The Los Alamos Water Boiler... for neutron multiplication experiments at moderate power. They also...

348

The second and third NGNP advanced gas reactor fuel irradiation experiments  

SciTech Connect (OSTI)

The United States Dept. of Energy's Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is currently scheduled to irradiate a total of five low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The irradiations are being accomplished to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas cooled reactors. The experiments will each consist of at least six separate capsules, and will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The effluent sweep gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and completed a very successful irradiation in early November 2009. The second experiment (AGR-2) started irradiation in June 2010, and the third and fourth experiments have been combined into a single larger irradiation (AGR-3/4) that is currently being assembled. The design and status of the second through fourth experiments as well as the irradiation results of the second experiment to date are discussed. (authors)

Grover, S. B.; Petti, D. A. [Idaho National Laboratory, 2525 N. Fremont Ave., Idaho Falls, ID 83415 (United States)

2012-07-01T23:59:59.000Z

349

Advanced Core Design And Fuel Management For Pebble-Bed Reactors  

SciTech Connect (OSTI)

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

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

2004-10-01T23:59:59.000Z

350

QED3 Theory of High Temperature Superconductors  

E-Print Network [OSTI]

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

Tesanovic, Zlatko

351

High Temperature, Permanent Magnet Biased Magnetic Bearings  

E-Print Network [OSTI]

performance, high speed and high temperature applications like space vehicles, jet engines and deep sea equipment. The bearing system had a target design to carry a load equal to 500 lb-f (2225N). Another objective was to design and build a test rig fixture...

Gandhi, Varun R.

2010-07-14T23:59:59.000Z

352

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

353

High temperature hot water systems: A primer  

SciTech Connect (OSTI)

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.

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

1998-01-01T23:59:59.000Z

354

Potential applications of high temperature helium  

SciTech Connect (OSTI)

This paper discusses the DOE MHTGR-SC program`s recent activity to improve the economics of the MHTGR without sacrificing safety performance and two potential applications of high temperature helium, the MHTGR gas turbine plant and a process heat application for methanol production from coal.

Schleicher, R.W. Jr.; Kennedy, A.J.

1992-09-01T23:59:59.000Z

355

Potential applications of high temperature helium  

SciTech Connect (OSTI)

This paper discusses the DOE MHTGR-SC program's recent activity to improve the economics of the MHTGR without sacrificing safety performance and two potential applications of high temperature helium, the MHTGR gas turbine plant and a process heat application for methanol production from coal.

Schleicher, R.W. Jr.; Kennedy, A.J.

1992-09-01T23:59:59.000Z

356

Thermal disconnect for high-temperature batteries  

DOE Patents [OSTI]

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.

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

2000-01-01T23:59:59.000Z

357

Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)  

SciTech Connect (OSTI)

The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

J. K. Wright; R. N. Wright

2010-07-01T23:59:59.000Z

358

Updated Generation IV Reactors Integrated Materials Technology Program Plan, Revision 2  

SciTech Connect (OSTI)

The Department of Energy's (DOE's) Generation IV Nuclear Energy Systems Program will address the research and development (R&D) necessary to support next-generation nuclear energy systems. Such R&D will be guided by the technology roadmap developed for the Generation IV International Forum (GIF) over two years with the participation of over 100 experts from the GIF countries. The roadmap evaluated over 100 future systems proposed by researchers around the world. The scope of the R&D described in the roadmap covers the six most promising Generation IV systems. The effort ended in December 2002 with the issue of the final Generation IV Technology Roadmap [1.1]. The six most promising systems identified for next generation nuclear energy are described within the 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. 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. Accordingly, DOE has identified materials as one of the focus areas for Gen IV technology development.

Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Halsey, William [Lawrence Livermore National Laboratory (LLNL); Hayner, George [Idaho National Laboratory (INL); Katoh, Yutai [ORNL; Klett, James William [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Stoller, Roger E [ORNL; Wilson, Dane F [ORNL

2005-12-01T23:59:59.000Z

359

Microchannel High-Temperature Recuperator for Fuel Cell Systems...  

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

Microchannel High-Temperature Recuperator for Fuel Cell Systems - Fact Sheet, 2011 Microchannel High-Temperature Recuperator for Fuel Cell Systems - Fact Sheet, 2011 FuelCell...

360

High Reliability, High TemperatureThermoelectric Power Generation...  

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

Reliability, High TemperatureThermoelectric Power Generation Materials and Technologies High Reliability, High TemperatureThermoelectric Power Generation Materials and Technologies...

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


361

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...

362

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...

363

High Temperature Polymer Membrane Development at Argonne National...  

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

Polymer Membrane Development at Argonne National Laboratory High Temperature Polymer Membrane Development at Argonne National Laboratory Summary of ANL's high temperature polymer...

364

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.

365

Low and high Temperature Dual Thermoelectric Generation Waste...  

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

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

366

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...

367

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

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

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

368

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.

369

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...

370

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...

371

High Temperature Polymer Membrane Development at Argonne National Laboratory  

Broader source: Energy.gov [DOE]

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

372

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

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

Generator Development of a 100-Watt High Temperature Thermoelectric Generator Test results for low and high temperature thermoelectric generators (TEG) those for a...

373

Microchannel High-Temperature Recuperator for Fuel Cell Systems...  

Energy Savers [EERE]

Microchannel High-Temperature Recuperator for Fuel Cell Systems - Fact Sheet, 2014 Microchannel High-Temperature Recuperator for Fuel Cell Systems - Fact Sheet, 2014 FuelCell...

374

Summary of space nuclear reactor power systems, 1983--1992  

SciTech Connect (OSTI)

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

Buden, D.

1993-08-11T23:59:59.000Z

375

Drart environmental impact statement siting, construction, and operation of New Production Reactor capacity. Volume 4, Appendices D-R  

SciTech Connect (OSTI)

This Environmental Impact Statement (EIS) assesses the potential environmental impacts, both on a broad programmatic level and on a project-specific level, concerning a proposed action to provide new tritium production capacity to meet the nation`s nuclear defense requirements well into the 21st century. A capacity equivalent to that of about a 3,000-megawatt (thermal) heavy-water reactor was assumed as a reference basis for analysis in this EIS; this is the approximate capacity of the existing production reactors at DOE`s Savannah River Site near Aiken, South Carolina. The EIS programmatic alternatives address Departmental decisions to be made on whether to build new production facilities, whether to build one or more complexes, what size production capacity to provide, and when to provide this capacity. Project-specific impacts for siting, constructing, and operating new production reactor capacity are assessed for three alternative sites: the Hanford Site near Richland, Washington; the Idaho National Engineering Laboratory near Idaho Falls, Idaho; and the Savannah River Site. For each site, the impacts of three reactor technologies (and supporting facilities) are assessed: a heavy-water reactor, a light-water reactor, and a modular high-temperature gas-cooled reactor. Impacts of the no-action alternative also are assessed. The EIS evaluates impacts related to air quality; noise levels; surface water, groundwater, and wetlands; land use; recreation; visual environment; biotic resources; historical, archaeological, and cultural resources; socioeconomics; transportation; waste management; and human health and safety. The EIS describes in detail the potential radioactive releases from new production reactors and support facilities and assesses the potential doses to workers and the general public. This volume contains 15 appendices.

Not Available

1991-04-01T23:59:59.000Z

376

Brazing Refractory Metals Used In High-Temperature Nuclear Instrumentation  

SciTech Connect (OSTI)

As part of the U. S. Department of Energy (DOE) sponsored Next Generation Nuclear Project (NGNP) currently ongoing at Idaho National Laboratory (INL), the irradiation performance of candidate high-temperature gas reactor fuels and materials is being evaluated at INL’s Advanced Test Reactor (ATR). The design of the first Advanced Gas Reactor (AGR 1) experiment, currently being irradiated in the ATR, required development of special techniques for brazing niobium and molybdenum. Brazing is one technique used to join refractory metals to each other and to stainless steel alloys. Although brazing processes are well established, it is difficult to braze niobium, molybdenum, and most other refractory metals because they quickly develop adherent oxides when exposed to room-temperature air. Specialized techniques and methods were developed by INL to overcome these obstacles. This paper describes the techniques developed for removing these oxides, as well as the ASME Section IX-qualified braze procedures that were developed as part of the AGR-1 project. All brazes were made using an induction coil with an inert or reducing atmosphere at low pressure. Other parameters, such as filler metals, fluxes used, and general setup procedures, are also discussed.

A. J. Palmer; C. J. Woolstenhulme

2009-06-01T23:59:59.000Z

377

Thermal fuse for high-temperature batteries  

DOE Patents [OSTI]

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.

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

2000-01-01T23:59:59.000Z

378

Establishment of Harrop, High-Temperature Viscometer  

SciTech Connect (OSTI)

This report explains how the Harrop, High-Temperature Viscometer was installed, calibrated, and operated. This report includes assembly and alignment of the furnace, viscometer, and spindle, and explains the operation of the Brookfield Viscometer, the Harrop furnace, and the UDC furnace controller. Calibration data and the development of the spindle constant from NIST standard reference glasses is presented. A simple operational procedure is included.

Schumacher, R.F.

1999-11-05T23:59:59.000Z

379

High Temperature Materials for Aerospace Applications  

E-Print Network [OSTI]

below 430 ?C for exposure times up to 20 minutes. Transition-metal carbides were initially synthesized by carbothermal reduction of transition-metal halides and polymer precursor mixtures, at temperatures that range from 900 to 1500 ?C in an argon... ........................................ 20 2.3 Present/Future Aerospace Applications ......................................... 24 2.4 Ultra-High Temperature Materials ................................................. 27 2.4.1 Transition-Metal Carbides...

Adamczak, Andrea Diane

2011-08-08T23:59:59.000Z

380

Corrosion Resistant Coatings for High Temperature Applications  

SciTech Connect (OSTI)

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.

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

1998-12-01T23:59:59.000Z

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


381

High temperature hot water distribution system study  

SciTech Connect (OSTI)

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.

NONE

1996-12-01T23:59:59.000Z

382

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

DOE Patents [OSTI]

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.

Yunker, Wayne H. (Richland, WA); Christiansen, David W. (Kennewick, WA)

1987-01-01T23:59:59.000Z

383

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

E-Print Network [OSTI]

-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...

Barner, Robert Buckner

2007-04-25T23:59:59.000Z

384

A HIGH TEMPERATURE GAS RECEIVER UTILIZING SMALL PARTICLES  

E-Print Network [OSTI]

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

Hunt, Arlon

2012-01-01T23:59:59.000Z

385

Using risk-based regulations for licensing nuclear power plants : case study of gas-cooled fast reactor  

E-Print Network [OSTI]

The strategy adopted for national energy supply is one of the most important policy choice for the US. Although it has been dismissed in the past decades, nuclear power today has key assets when facing concerns on energy ...

Jourdan, Grégoire

2005-01-01T23:59:59.000Z

386

Compatibility of aluminide-coated Hastelloy x and Inconel 617 in a simulated gas-cooled reactor environment  

SciTech Connect (OSTI)

Commercially prepared aluminide coatings on Hastelloy X and Inconel 617 substrates were exposed to controlled-impurity helium at 850/sup 0/ and 950/sup 0/C for 3000 h. Optical and scanning electron (SEM) microscopy, electron microprobe profiles, and SEM X-ray mapping were used to evaluate and compare exposed and unexposed control samples. Four coatings were evaluated: aluminide, aluminide with platinum, aluminide with chromium, and aluminide with rhodium. With extended time at elevated temperature, nickel diffused into the aluminide coatings to form epsilon-phase (Ni/sub 3/Al). This diffusion was the primary cause of porosity formation at the aluminide/alloy interface.

Chin, J.; Johnson, W. R.; Chen, K.

1982-03-01T23:59:59.000Z

387

Design of an experimental loop for post-LOCA heat transfer regimes in a Gas-cooled Fast Reactor  

E-Print Network [OSTI]

The goal of this thesis is to design an experimental thermal-hydraulic loop capable of generating accurate, reliable data in various convection heat transfer regimes for use in the formulation of a comprehensive convection ...

Cochran, Peter A. (Peter Andrew)

2005-01-01T23:59:59.000Z

388

Application of Molten Salt Reactor Technology to MMW In-Space NEP and Surface Power Missions  

SciTech Connect (OSTI)

Anticipated manned nuclear electric propulsion (NEP) and planetary surface power missions will require multi-megawatt nuclear reactors that are lightweight, operationally robust, and sealable in power for widely varying scientific mission objectives. Molten salt reactor technology meets all of these requirements and offers an interesting alternative to traditional multi-megawatt gas-cooled and liquid metal concepts. (authors)

Patton, Bruce; Sorensen, Kirk [Propulsion Research Center, Marshall Space Flight Center, Huntsville, AL 35812 (United States)

2002-07-01T23:59:59.000Z

389

Fermi liquid theory for high temperature superconductors  

SciTech Connect (OSTI)

In this article the Fermi liquid theory of metals is discussed starting from Luttinger's theorem. The content of Luttinger's Theorem and its implications for microscopic theories of high temperature superconductors are discussed. A simple quasi-2d Fermi liquid theory is introduced and some of its properties are calculated. It is argued that a number of experiments on YBa/sub 2/Cu/sub 3/O/sub 6+x/, x > 0.5, strongly suggest the existence of a Fermi surface and thereby a Fermi liquid normal state. 25 refs., 1 fig.

Bedell, K.S.

1988-01-01T23:59:59.000Z

390

High-temperature directional drilling turbodrill  

SciTech Connect (OSTI)

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.

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

1982-02-01T23:59:59.000Z

391

High temperature regenerable hydrogen sulfide removal agents  

DOE Patents [OSTI]

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.

Copeland, Robert J. (Wheat Ridge, CO)

1993-01-01T23:59:59.000Z

392

High Temperature Materials Laboratory third annual report  

SciTech Connect (OSTI)

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.

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

1990-12-01T23:59:59.000Z

393

High temperature intermetallic binders for HVOF carbides  

SciTech Connect (OSTI)

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.

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

394

High temperature lined conduits, elbows and tees  

DOE Patents [OSTI]

A high temperature lined conduit comprising, a liner, a flexible insulating refractory blanket around and in contact with the liner, a pipe member around the blanket and spaced therefrom, and castable rigid refractory material between the pipe member and the blanket. Anchors are connected to the inside diameter of the pipe and extend into the castable material. The liner includes male and female slip joint ends for permitting thermal expansion of the liner with respect to the castable material and the pipe member. Elbows and tees of the lined conduit comprise an elbow liner wrapped with insulating refractory blanket material around which is disposed a spaced elbow pipe member with castable refractory material between the blanket material and the elbow pipe member. A reinforcing band is connected to the elbow liner at an intermediate location thereon from which extend a plurality of hollow tubes or pins which extend into the castable material to anchor the lined elbow and permit thermal expansion. A method of fabricating the high temperature lined conduit, elbows and tees is also disclosed which utilizes a polyethylene layer over the refractory blanket after it has been compressed to maintain the refractory blanket in a compressed condition until the castable material is in place. Hot gases are then directed through the interior of the liner for evaporating the polyethylene and setting the castable material which permits the compressed blanket to come into close contact with the castable material.

De Feo, Angelo (Passaic, NJ); Drewniany, Edward (Bergen, NJ)

1982-01-01T23:59:59.000Z

395

High temperature electrochemical corrosion rate probes  

SciTech Connect (OSTI)

Corrosion occurs in the high temperature sections of energy production plants due to a number of factors: ash deposition, coal composition, thermal gradients, and low NOx conditions, among others. Electrochemical corrosion rate (ECR) probes have been shown to operate in high temperature gaseous environments that are similar to those found in fossil fuel combustors. ECR probes are rarely used in energy production plants at the present time, but if they were more fully understood, corrosion could become a process variable at the control of plant operators. Research is being conducted to understand the nature of these probes. Factors being considered are values selected for the Stern-Geary constant, the effect of internal corrosion, and the presence of conductive corrosion scales and ash deposits. The nature of ECR probes will be explored in a number of different atmospheres and with different electrolytes (ash and corrosion product). Corrosion rates measured using an electrochemical multi-technique capabilities instrument will be compared to those measured using the linear polarization resistance (LPR) technique. In future experiments, electrochemical corrosion rates will be compared to penetration corrosion rates determined using optical profilometry measurements.

Bullard, Sophie J.; Covino, Bernard S., Jr.; Holcomb, Gordon R.; Ziomek-Moroz, M.

2005-09-01T23:59:59.000Z

396

Evaluation of the Gas Turbine Modular Helium Reactor  

SciTech Connect (OSTI)

Recent advances in gas-turbine and heat exchanger technology have enhanced the potential for a Modular Helium Reactor (MHR) incorporating a direct gas turbine (Brayton) cycle for power conversion. The resulting Gas Turbine Modular Helium Reactor (GT-MHR) power plant combines the high temperature capabilities of the MHR with the efficiency and reliability of modern gas turbines. While the passive safety features of the steam cycle MHR (SC-MHR) are retained, generation efficiencies are projected to be in the range of 48% and steam power conversion systems, with their attendant complexities, are eliminated. Power costs are projected to be reduced by about 20%, relative to the SC-MHR or coal. This report documents the second, and final, phase of a two-part evaluation that concluded with a unanimous recommendation that the direct cycle (DC) variant of the GT-MHR be established as the commercial objective of the US Gas-Cooled Reactor Program. This recommendation has been endorsed by industrial and utility participants and accepted by the US Department of Energy (DOE). The Phase II effort, documented herein, concluded that the DC GT-MHR offers substantial technical and economic advantages over both the IDC and SC systems. Both the DC and IDC were found to offer safety advantages, relative to the SC, due to elimination of the potential for water ingress during power operations. This is the dominant consequence event for the SC. The IDC was judged to require somewhat less development than the direct cycle, while the SC, which has the greatest technology base, incurs the least development cost and risk. While the technical and licensing requirements for the DC were more demanding, they were judged to be incremental and feasible. Moreover, the DC offers significant performance and cost improvements over the other two concepts. Overall, the latter were found to justify the additional development needs.

Not Available

1994-02-01T23:59:59.000Z

397

Draft environmental impact statement for the siting, construction, and operation of New Production Reactor capacity. Volume 3, Sections 7-12, Appendices A-C  

SciTech Connect (OSTI)

This Environmental Impact Statement (EIS) assesses the potential environmental impacts, both on a broad programmatic level and on a project-specific level, concerning a proposed action to provide new tritium production capacity to meet the nation`s nuclear defense requirements well into the 21st century. A capacity equivalent to that of about a 3,000-megawatt (thermal) heavy-water reactor was assumed as a reference basis for analysis in this EIS; this is the approximate capacity of the existing production reactors at DOE`s Savannah River Site near Aiken, South Carolina. The EIS programmatic alternatives address Departmental decisions to be made on whether to build new production facilities, whether to build one or more complexes, what size production capacity to provide, and when to provide this capacity. Project-specific impacts for siting, constructing, and operating new production reactor capacity are assessed for three alternative sites: the Hanford Site near Richland, Washington; the Idaho National Engineering Laboratory near Idaho Falls, Idaho; and the Savannah River Site. For each site, the impacts of three reactor technologies (and supporting facilities) are assessed: a heavy-water reactor, a light-water reactor, and a modular high-temperature gas-cooled reactor. Impacts of the no-action alternative also are assessed. The EIS evaluates impacts related to air quality; noise levels; surface water, groundwater, and wetlands; land use; recreation; visual environment; biotic resources; historical, archaeological, and cultural resources; socioeconomics; transportation; waste management; and human health and safety. The EIS describes in detail the potential radioactive releases from new production reactors and support facilities and assesses the potential doses to workers and the general public. This volume contains references; a list of preparers and recipients; acronyms, abbreviations, and units of measure; a glossary; an index and three appendices.

Not Available

1991-04-01T23:59:59.000Z

398

Draft environmental impact statement for the siting, construction, and operation of New Production Reactor capacity. Volume 2, Sections 1-6  

SciTech Connect (OSTI)

This (EIS) assesses the potential environmental impacts, both on a broad programmatic level and on a project-specific level, concerning a proposed action to provide new tritium production capacity to meet the nation`s nuclear defense requirements well into the 21st century. A capacity equivalent to that of about a 3,000-megawatt (thermal) heavy-water reactor was assumed as a reference basis for analysis in this EIS; this is the approximate capacity of the existing production reactors at DOE`s Savannah River Site. The EIS programmatic alternatives address Departmental decisions to be made on whether to build new production facilities, whether to build one or more complexes, what size production capacity to provide, and when to provide this capacity. Project-specific impacts for siting, constructing, and operating new production reactor capacity are assessed for three alternative sites: the Hanford Site near Richland, Washington; the Idaho National Engineering Laboratory near Idaho Falls, Idaho; and the Savannah River Site. For each site, the impacts of three reactor technologies (and supporting facilities) are assessed: a heavy-water reactor, a light-water reactor, and a modular high-temperature gas-cooled reactor. Impacts of the no-action alternative also are assessed. The EIS evaluates impacts related to air quality; noise levels; surface water, groundwater, and wetlands; land use; recreation; visual environment; biotic resources; historical, archaeological, and cultural resources; socioeconomics; transportation; waste management; and human health and safety. The EIS describes in detail the potential radioactive releases from new production reactors and support facilities and assesses the potential doses to workers and the general public. This volume contains the analysis of programmatic alternatives, project alternatives, affected environment of alternative sites, environmental consequences, and environmental regulations and permit requirements.

Not Available

1991-04-01T23:59:59.000Z

399

THERMODYNAMIC CONSIDERATIONS FOR THERMAL WATER SPLITTING PROCESSES AND HIGH TEMPERATURE ELECTROLYSIS  

SciTech Connect (OSTI)

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.

J. E. O'Brien

2008-11-01T23:59:59.000Z

400

High temperature ceramic composition for hydrogen retention  

DOE Patents [OSTI]

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)

Webb, R.W.

1974-01-01T23:59:59.000Z

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


401

Compliant high temperature seals for dissimilar materials  

DOE Patents [OSTI]

A high temperature, gas-tight seal is formed by utilizing one or more compliant metallic toroidal ring sealing elements, where the applied pressure serves to activate the seal, thus improving the quality of the seal. The compliant nature of the sealing element compensates for differences in thermal expansion between the materials to be sealed, and is particularly useful in sealing a metallic member and a ceramic tube art elevated temperatures. The performance of the seal may be improved by coating the sealing element with a soft or flowable coating such as silver or gold and/or by backing the sealing element with a bed of fine powder. The material of the sealing element is chosen such that the element responds to stress elastically, even at elevated temperatures, permitting the seal to operate through multiple thermal cycles.

Rynders, Steven Walton (Fogelsville, PA); Minford, Eric (Laurys Station, PA); Tressler, Richard Ernest (Boalsburg, PA); Taylor, Dale M. (Salt Lake City, UT)

2001-01-01T23:59:59.000Z

402

High temperature coatings for gas turbines  

DOE Patents [OSTI]

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.

Zheng, Xiaoci Maggie

2003-10-21T23:59:59.000Z

403

Thermochemistry of high-temperature corrosion  

SciTech Connect (OSTI)

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.

Natesan, K.

1980-01-01T23:59:59.000Z

404

Multilayer ultra-high-temperature ceramic coatings  

DOE Patents [OSTI]

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.

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

2012-03-20T23:59:59.000Z

405

High temperature insulation for ceramic matrix composites  

SciTech Connect (OSTI)

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.

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

2000-01-01T23:59:59.000Z

406

High temperature insulation for ceramic matrix composites  

DOE Patents [OSTI]

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.

Merrill, Gary B.; Morrison, Jay Alan

2004-01-13T23:59:59.000Z

407

High temperature insulation for ceramic matrix composites  

SciTech Connect (OSTI)

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.

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

2001-01-01T23:59:59.000Z

408

Turbine vane with high temperature capable skins  

DOE Patents [OSTI]

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.

Morrison, Jay A. (Oviedo, FL)

2012-07-10T23:59:59.000Z

409

High temperature low friction surface coating  

DOE Patents [OSTI]

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.

Bhushan, Bharat (Watervliet, NY)

1980-01-01T23:59:59.000Z

410

Expansion Joint Concepts for High Temperature Insulation Systems  

E-Print Network [OSTI]

EXPANSION JOINT CONCEPTS FOR HIGH TEMPERATURE INSULATION SYSTEMS Michael R. Harrison Johns-Manville Sales Corporation ";.,' Denver, Colorado ABSTRACT As high temperature steam and process piping expands with heat, joints beg in to open...

Harrison, M. R.

1980-01-01T23:59:59.000Z

411

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...

412

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...

413

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...

414

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...

415

Phenotyping of High Temperature Susceptibility in Garden Roses (Rosa xhybrida)  

E-Print Network [OSTI]

cultivars. Adaptation to high temperature stress is viewed as high priority in breeding programs of all major crops. High temperature stress negatively affects garden rose performance and the quality of flowers produced. The work described...

Greyvenstein, Ockert Frederick

2013-12-10T23:59:59.000Z

416

NGNP High Temperature Materials White Paper  

SciTech Connect (OSTI)

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

Lew Lommers; George Honma

2012-08-01T23:59:59.000Z

417

High Temperature Integrated Thermoelectric Ststem and Materials  

SciTech Connect (OSTI)

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.

Mike S. H. Chu

2011-06-06T23:59:59.000Z

418

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.

419

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.

420

Nanostructured High-Temperature Bulk Thermoelectric Energy Conversion...  

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

More Documents & Publications Nanostructured High-Temperature Bulk Thermoelectric Energy Conversion for Efficient Automotive Waste Heat Recovery Vehicle Technologies Office...

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


421

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

SciTech Connect (OSTI)

In the nuclear energy renaissance, driven by fission reactor concepts utilizing very high temperatures and fast neutron spectra, materials with enhanced performance that exceeds are expected to play a central role. With the operating temperatures of the Generation III reactors bringing the classical reactor materials close to their performance limits there is an urgent need to develop and qualify new alloys and composites. Efforts have been focused on the intricate relations and the high demands placed on materials at the anticipated extreme states within the next generation fusion and fission reactors which combine high radiation fluxes, elevated temperatures and aggressive environments. While nuclear reactors have been in operation for several decades, the structural materials associated with the next generation options need to endure much higher temperatures (1200 C), higher neutron doses (tens of displacements per atom, dpa), and extremely corrosive environments, which are beyond the experience on materials accumulated to-date. The most important consideration is the performance and reliability of structural materials for both in-core and out-of-core functions. While there exists a great body of nuclear materials research and operating experience/performance from fission reactors where epithermal and thermal neutrons interact with materials and alter their physio-mechanical properties, a process that is well understood by now, there are no operating or even experimental facilities that will facilitate the extreme conditions of flux and temperature anticipated and thus provide insights into the behaviour of these well understood materials. Materials, however, still need to be developed and their interaction and damage potential or lifetime to be quantified for the next generation nuclear energy. Based on material development advances, composites, and in particular ceramic composites, seem to inherently possess properties suitable for key functions within the operating envelope of both fission and fusion reactors. In advanced fission reactors composite materials are being designed in an effort to extend the life and improve the reliability of fuel rod cladding as well as structural materials. Composites are being considered for use as core internals in the next generation of gas-cooled reactors. Further, next-generation plasma-fusion reactors, such as the International Thermonuclear Experimental Reactor (ITER) will rely on the capabilities of advanced composites to safely withstand extremely high neutron fluxes while providing superior thermal shock resistance.

Simos, N.

2011-05-01T23:59:59.000Z

422

Heat Exchangers for the Next Generation of Nuclear Reactors  

SciTech Connect (OSTI)

The realisation that fossil fuel resources are finite, the associated rising price and a growing concern about greenhouse gas emissions, has resulted in renewed interest in nuclear energy. Generation IV and other programmes are looking at a variety of new reactors. These reactors vary in type from Very High Temperature Gas Cooled Reactors (VHTR) to Liquid Metal Fast Reactors (LFR and SFR) with cooling mediums that include: - Helium, - Supercritical carbon dioxide, - Sodium, - Lead, - Molten salts. In addition interest is not just focused on production of electrical power with an efficiency greater than that associated with the Rankine Cycle (typically 30 -35%); there is now genuine interest in nuclear energy as a heat source for hydrogen production, via the Sulphur Iodine Process (SI) or high temperature electrolysis. The production of electrical power at higher efficiency via a Brayton Cycle, and hydrogen production requires both heat at higher temperatures, up to 1000 deg C and high effectiveness heat exchange to transfer the heat to either the power or process cycle. This presents new challenges for the heat exchangers. If plant efficiencies are to be improved there is a need for: - High effectiveness heat exchange at minimal pressure drop; - Compact heat exchange to improve safety and economics; - An ability to build coded heat exchangers in a variety of nickel based alloys, oxide dispersion strengthened alloys (ODS) and ceramic materials to address the temperature, life and corrosion issues associated with these demanding duties. Heatric has already given consideration to many of these challenges. Their Print Circuit Heat Exchanger (PCHE) and Formed Plate Heat Exchanger (FPHE) technology which are commercially available today, will fulfill all of the duties up to temperatures of 950 deg C. In addition products currently under development will further increase the temperature and pressure range, while offering greater corrosion resistance and operational life. This paper outlines the challenges for the heat exchangers and the development required, with particular attention given to material selection. It is further the objective of this study to demonstrate that heat exchangers such as PCHE and FPHE are able to meet the above challenges. (authors)

Xiuqing, Li; Le Pierres, Renaud; Dewson, Stephen John [Heatric Division of Meggitt (UK) Ltd., 46 Holton Road, Holton Heath, Poole, Dorset BH16 6LT (United Kingdom)

2006-07-01T23:59:59.000Z

423

High Temperature Borehole Televiewer software user manual  

SciTech Connect (OSTI)

The High Temperature Borehole Televiewer is a downhole instrument which provides acoustic pictures of the borehole walls that are suitable for casing inspection and fracture detection in geothermal wells. The Geothermal Drilling Organization has funded the development of a commercial tool survivable to temperatures of 275{degree}C and pressures of 5000 psi. A real-time display on an IBM-compatible PC was included as part of the development effort. This report contains a User Manual which describes the operation of this software. The software is designed in a menu format allowing the user to change many of the parameters which control both the acquisition and the display of the Televiewer data. An internal data acquisition card digitizes the waveform from the tool at a rate of 100,000 samples per second. The data from the tool, both the range or arrival time and the amplitude of the return signal, are displayed in color on the CRT screen of the computer during the logging operation. This data may be stored on the hard disk for later display and analysis. The software incorporates many features which aid in the setup of the tool for proper operation. These features include displaying and storing the captured waveform data to check the voltage and time windows selected by the user. 17 refs., 28 figs., 15 tabs.

Duda, L.E.

1989-11-01T23:59:59.000Z

424

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

SciTech Connect (OSTI)

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.

Morris, J. F.

1985-03-19T23:59:59.000Z

425

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

DOE Patents [OSTI]

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.

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

1983-11-25T23:59:59.000Z

426

Thermal stability of high temperature structural alloys  

SciTech Connect (OSTI)

High temperature structural alloys were evaluated for suitability for long term operation at elevated temperatures. The effect of elevated temperature exposure on the microstructure and mechanical properties of a number of alloys was characterized. Fe-based alloys (330 stainless steel, 800H, and mechanically alloyed MA 956), and Ni-based alloys (Hastelloy X, Haynes 230, Alloy 718, and mechanically alloyed MA 758) were evaluated for room temperature tensile and impact toughness properties after exposure at 750 C for 10,000 hours. Of the Fe-based alloys evaluated, 330 stainless steel and 800H showed secondary carbide (M{sub 23}C{sub 6}) precipitation and a corresponding reduction in ductility and toughness as compared to the as-received condition. Within the group of Ni-based alloys tested, Alloy 718 showed the most dramatic structure change as it formed delta phase during 10,000 hours of exposure at 750 C with significant reductions in strength, ductility, and toughness. Haynes 230 and Hastelloy X showed significant M{sub 23}C{sub 6} carbide precipitation and a resulting reduction in ductility and toughness. Haynes 230 was also evaluated after 10,000 hours of exposure at 850, 950, and 1050 C. For the 750--950 C exposures the M{sub 23}C{sub 6} carbides in Haynes 230 coarsened. This resulted in large reductions in impact strength and ductility for the 750, 850 and 950 C specimens. The 1050 C exposure specimens showed the resolution of M{sub 23}C{sub 6} secondary carbides, and mechanical properties similar to the as-received solution annealed condition.

Jordan, C.E.; Rasefske, R.K.; Castagna, A. [Lockheed Martin Corp., Schenectady, NY (United States)

1999-03-01T23:59:59.000Z

427

Geochemistry of Aluminum in High Temperature Brines  

SciTech Connect (OSTI)

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-250°C). 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.

Benezeth, P.; Palmer, D.A.; Wesolowski, D.J.

1999-05-18T23:59:59.000Z

428

BNL Pulsed Magnet Inertially Cooled , LN2 or 30K He Gas Cooled Between Shots MIT test will use only LN2  

E-Print Network [OSTI]

BNL Pulsed Magnet ­Inertially Cooled , LN2 or 30K He Gas Cooled Between Shots ­MIT test will use 2.0 Objectives 3.0 Test Location 4.0 Critical Lifts 5.0 Power Supplies 6.0 Cryogenic System for MIT field (exclusive of the proton beam) Cost issues dictated a modest coil design. Power supply limitations

McDonald, Kirk

429

Pressure Resistance Welding of High Temperature Metallic Materials  

SciTech Connect (OSTI)

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.

N. Jerred; L. Zirker; I. Charit; J. Cole; M. Frary; D. Butt; M. Meyer; K. L. Murty

2010-10-01T23:59:59.000Z

430

Status of the INL high-temperature electrolysis research program –experimental and modeling  

SciTech Connect (OSTI)

This paper provides a status update on the high-temperature electrolysis (HTE) research and development program at the Idaho National Laboratory (INL), with an overview of recent large-scale system modeling results and the status of the experimental program. System analysis results have been obtained using the commercial code UniSim, augmented with a custom high-temperature electrolyzer module. 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 coolant outlet temperatures. In terms of experimental research, the INL has recently completed an Integrated Laboratory Scale (ILS) HTE test at the 15 kW level. The initial hydrogen production rate for the ILS test was in excess of 5000 liters per hour. Details of the ILS design and operation will be presented. Current small-scale experimental research is focused on improving the degradation characteristics of the electrolysis cells and stacks. Small-scale testing ranges from single cells to multiple-cell stacks. The INL is currently in the process of testing several state-of-the-art anode-supported cells and is working to broaden its relationship with industry in order to improve the long-term performance of the cells.

J. E. O'Brien; C. M. Stoots; M. G. McKellar; E. A. Harvego; K. G. Condie; G. K. Housley; J. S. Herring; J. J. Hartvigsen

2009-04-01T23:59:59.000Z

431

Microsoft Word - EXT-12-24761 AGR-1 TC Data Analysis - FINAL...  

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

the Advanced Test Reactor (ATR) at INL to obtain data on fuel behavior under normal and accident conditions of a new generation of very high-temperature gas-cooled 2 reactors...

432

High Temperature Oxidation Performance of Aluminide Coatings  

SciTech Connect (OSTI)

Aluminide coatings are of interest for many high temperature applications because of the possibility of improving the oxidation resistance of structural alloys by forming a protective external alumina scale. Steam and exhaust gas environments are of particular interest because alumina is less susceptible to the accelerated attack due to hydroxide formation observed for chromia- and silica-forming alloys and ceramics. For water vapor testing, one ferritic (Fe-9Cr-1Mo) and one austenitic alloy (304L) have been selected as substrate materials and CVD coatings have been used in order to have a well-controlled, high purity coating. It is anticipated that similar aluminide coatings could be made by a higher-volume, commercial process such as pack cementation. Previous work on this program has examined as-deposited coatings made by high and low Al activity CVD processes and the short-term performance of these coatings. The current work is focusing on the long term behavior in both diffusion tests16 and oxidation tests of the thicker, high Al activity coatings. For long-term coating durability, one area of concern has been the coefficient of thermal expansion (CTE) mismatch between coating and substrate. This difference could cause cracking or deformation that could reduce coating life. Corrosion testing using thermal cycling is of particular interest because of this potential problem and results are presented where a short exposure cycle (1h) severely degraded aluminide coatings on both types of substrates. To further study the potential role of aluminide coatings in fossil energy applications, several high creep strength Ni-base alloys were coated by CVD for testing in a high pressure (20atm) steam-CO{sub 2} environment for the ZEST (zero-emission steam turbine) program. Such alloys would be needed as structural and turbine materials in this concept. For Ni-base alloys, CVD produces a {approx}50{mu}m {beta}-NiAl outer layer with an underlying interdiffusion zone. Specimens of HR160, alloy 601 and alloy 230 were tested with and without coatings at 900 C and preliminary post-test characterization is reported.

Pint, Bruce A [ORNL; Zhang, Ying [Tennessee Technological University; Haynes, James A [ORNL; Wright, Ian G [ORNL

2004-01-01T23:59:59.000Z

433

3D CFD Model of High Temperature H2O/CO2 Co-electrolysis  

SciTech Connect (OSTI)

3D CFD Model of High Temperature H2O/CO2 Co-Electrolysis Grant Hawkes1, James O’Brien1, 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%.

Grant Hawkes; James O'Brien; Carl Stoots; Stephen Herring; Joe Hartvigsen

2007-06-01T23:59:59.000Z

434

Integrated High Temperature Coal-to-Hydrogen System with CO2 Separation  

SciTech Connect (OSTI)

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.

James A. Ruud; Anthony Ku; Vidya Ramaswamy; Wei Wei; Patrick Willson

2007-05-31T23:59:59.000Z

435

First high-temperature electronics products survey 2005.  

SciTech Connect (OSTI)

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.

Normann, Randy Allen

2006-04-01T23:59:59.000Z

436

Institute of Chemical Engineering and High Temperature Chemical...  

Open Energy Info (EERE)

Chemical Processes ICEHT Jump to: navigation, search Name: Institute of Chemical Engineering and High Temperature Chemical Processes (ICEHT) Place: Hellas, Greece Zip:...