National Library of Energy BETA

Sample records for reactor demonstration case

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

    SciTech Connect (OSTI)

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

    2012-08-01

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

  2. Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Demonstration...

    Office of Scientific and Technical Information (OSTI)

    (KRUSTY) Demonstration. CEDT Phase 1 Preliminary Design Documentation Citation Details In-Document Search Title: Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Demonstration. ...

  3. Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Demonstration...

    Office of Scientific and Technical Information (OSTI)

    The reactor design will include heat pipes coupled to Stirling engines to demonstrate how one can generate electricity when extracting energy from a "nuclear generated" heat ...

  4. NEAC Nuclear Reactor Technology (NRT) Subcommittee Advanced Test and/or Demonstration Reactor Planning Study

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

    Nuclear Reactor Technology (NRT) Subcommittee Advanced Test and/or Demonstration Reactor Planning Study October 6 th , 2015 Meeting Summary and Comments Given direction from Congress, the Department of Energy's Office of Nuclear Energy (DOE- NE) is conducting a planning study for an advanced test and/or demonstration reactor (AT/DR study) in the United States. The Nuclear Energy Advisory Committee (NEAC) and specifically its Nuclear Reactor Technology (NRT) subcommittee has been asked to provide

  5. Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Demonstration. CEDT

    Office of Scientific and Technical Information (OSTI)

    Phase 1 Preliminary Design Documentation (Technical Report) | SciTech Connect Technical Report: Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Demonstration. CEDT Phase 1 Preliminary Design Documentation Citation Details In-Document Search Title: Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Demonstration. CEDT Phase 1 Preliminary Design Documentation The intent of the integral experiment request IER 299 (called KiloPower by NASA) is to assemble and evaluate the operational

  6. Simulator platform for fast reactor operation and safety technology demonstration

    SciTech Connect (OSTI)

    Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J.

    2012-07-30

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

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

    SciTech Connect (OSTI)

    Curtis Smith; Steven Prescott; Tony Koonce

    2014-04-01

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

  8. PRELIMINARY DEMONSTRATION REACTOR POINT DESIGN FOR THE FLUORIDE...

    Office of Scientific and Technical Information (OSTI)

    HIGH-TEMPERATURE REACTOR Authors: Qualls, A L 1 ; Betzler, Benjamin R 1 ; Brown, Nicholas R 1 ; Carbajo, Juan 1 ; Greenwood, Michael Scott 1 ; Hale, Richard...

  9. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    SciTech Connect (OSTI)

    Qualls, A. L.; Betzler, Benjamin R.; Brown, Nicholas R.; Carbajo, Juan; Greenwood, Michael Scott; Hale, Richard Edward; Harrison, Thomas J.; Powers, Jeffrey J.; Robb, Kevin R.; Terrell, Jerry W.

    2015-12-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  10. Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Demonstration...

    Office of Scientific and Technical Information (OSTI)

    KiloPower by NASA) is to assemble and evaluate the operational performance of a compact reactor configuration that closely resembles the flight unit to be used by NASA to execute a ...

  11. Efficient Solutions for New Homes Case Study: Demonstration House...

    Energy Savers [EERE]

    Demonstration House of Cold-Climate Solutions for Affordable Housing Efficient Solutions for New Homes Case Study: Demonstration House of Cold-Climate Solutions for Affordable ...

  12. Beyond ITER: Neutral beams for a demonstration fusion reactor (DEMO) (invited)

    SciTech Connect (OSTI)

    McAdams, R.

    2014-02-15

    In the development of magnetically confined fusion as an economically sustainable power source, International Tokamak Experimental Reactor (ITER) is currently under construction. Beyond ITER is the demonstration fusion reactor (DEMO) programme in which the physics and engineering aspects of a future fusion power plant will be demonstrated. DEMO will produce net electrical power. The DEMO programme will be outlined and the role of neutral beams for heating and current drive will be described. In particular, the importance of the efficiency of neutral beam systems in terms of injected neutral beam power compared to wallplug power will be discussed. Options for improving this efficiency including advanced neutralisers and energy recovery are discussed.

  13. Thermal-Hydraulic Design of a Fluoride High-Temperature Demonstration Reactor

    SciTech Connect (OSTI)

    Carbajo, Juan J; Qualls, A L

    2016-01-01

    INTRODUCTION The Fluoride High-Temperature Reactor (FHR) named the Demonstration Reactor (DR) is a novel reactor concept using molten salt coolant and TRIstructural ISOtropic (TRISO) fuel that is being developed at Oak Ridge National Laboratory (ORNL). The objective of the FHR DR is to advance the technology readiness level of FHRs. The FHR DR will demonstrate technologies needed to close remaining gaps to commercial viability. The FHR DR has a thermal power of 100 MWt, very similar to the SmAHTR, another FHR ORNL concept (Refs. 1 and 2) with a power of 125 MWt. The FHR DR is also a small version of the Advanced High Temperature Reactor (AHTR), with a power of 3400 MWt, cooled by a molten salt and also being developed at ORNL (Ref. 3). The FHR DR combines three existing technologies: (1) high-temperature, low-pressure molten salt coolant, (2) high-temperature coated-particle TRISO fuel, (3) and passive decay heat cooling systems by using Direct Reactor Auxiliary Cooling Systems (DRACS). This paper presents FHR DR thermal-hydraulic design calculations.

  14. Specification of the Advanced Burner Test Reactor Multi-Physics Coupling Demonstration Problem

    SciTech Connect (OSTI)

    Shemon, E. R.; Grudzinski, J. J.; Lee, C. H.; Thomas, J. W.; Yu, Y. Q.

    2015-12-21

    This document specifies the multi-physics nuclear reactor demonstration problem using the SHARP software package developed by NEAMS. The SHARP toolset simulates the key coupled physics phenomena inside a nuclear reactor. The PROTEUS neutronics code models the neutron transport within the system, the Nek5000 computational fluid dynamics code models the fluid flow and heat transfer, and the DIABLO structural mechanics code models structural and mechanical deformation. The three codes are coupled to the MOAB mesh framework which allows feedback from neutronics, fluid mechanics, and mechanical deformation in a compatible format.

  15. Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Demonstration. CEDT Phase 1 Preliminary Design Documentation

    SciTech Connect (OSTI)

    Sanchez, Rene Gerardo; Hutchinson, Jesson D.; Mcclure, Patrick Ray; Myers, William L.

    2015-08-20

    The intent of the integral experiment request IER 299 (called KiloPower by NASA) is to assemble and evaluate the operational performance of a compact reactor configuration that closely resembles the flight unit to be used by NASA to execute a deep space exploration mission. The reactor design will include heat pipes coupled to Stirling engines to demonstrate how one can generate electricity when extracting energy from a “nuclear generated” heat source. This series of experiments is a larger scale follow up to the DUFF series of experiments1,2 that were performed using the Flat-Top assembly.

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

    SciTech Connect (OSTI)

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

    2013-11-01

    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.

  17. FINAL SIMULATION RESULTS FOR DEMONSTRATION CASE 1 AND 2

    SciTech Connect (OSTI)

    David Sloan; Woodrow Fiveland

    2003-10-15

    The goal of this DOE Vision-21 project work scope was to develop an integrated suite of software tools that could be used to simulate and visualize advanced plant concepts. Existing process simulation software did not meet the DOE's objective of ''virtual simulation'' which was needed to evaluate complex cycles. The overall intent of the DOE was to improve predictive tools for cycle analysis, and to improve the component models that are used in turn to simulate equipment in the cycle. Advanced component models are available; however, a generic coupling capability that would link the advanced component models to the cycle simulation software remained to be developed. In the current project, the coupling of the cycle analysis and cycle component simulation software was based on an existing suite of programs. The challenge was to develop a general-purpose software and communications link between the cycle analysis software Aspen Plus{reg_sign} (marketed by Aspen Technology, Inc.), and specialized component modeling packages, as exemplified by industrial proprietary codes (utilized by ALSTOM Power Inc.) and the FLUENT{reg_sign} computational fluid dynamics (CFD) code (provided by Fluent Inc). A software interface and controller, based on an open CAPE-OPEN standard, has been developed and extensively tested. Various test runs and demonstration cases have been utilized to confirm the viability and reliability of the software. ALSTOM Power was tasked with the responsibility to select and run two demonstration cases to test the software--(1) a conventional steam cycle (designated as Demonstration Case 1), and (2) a combined cycle test case (designated as Demonstration Case 2). Demonstration Case 1 is a 30 MWe coal-fired power plant for municipal electricity generation, while Demonstration Case 2 is a 270 MWe, natural gas-fired, combined cycle power plant. Sufficient data was available from the operation of both power plants to complete the cycle configurations. Three runs

  18. Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit

    SciTech Connect (OSTI)

    Merzari, E.; Shemon, E. R.; Yu, Y. Q.; Thomas, J. W.; Obabko, A.; Jain, Rajeev; Mahadevan, Vijay; Tautges, Timothy; Solberg, Jerome; Ferencz, Robert Mark; Whitesides, R.

    2015-12-21

    This report describes to employ SHARP to perform a first-of-a-kind analysis of the core radial expansion phenomenon in an SFR. This effort required significant advances in the framework Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit used to drive the coupled simulations, manipulate the mesh in response to the deformation of the geometry, and generate the necessary modified mesh files. Furthermore, the model geometry is fairly complex, and consistent mesh generation for the three physics modules required significant effort. Fully-integrated simulations of a 7-assembly mini-core test problem have been performed, and the results are presented here. Physics models of a full-core model of the Advanced Burner Test Reactor have also been developed for each of the three physics modules. Standalone results of each of the three physics modules for the ABTR are presented here, which provides a demonstration of the feasibility of the fully-integrated simulation.

  19. Experimental Development and Demonstration of Ultrasonic Measurement Diagnostics for Sodium Fast Reactor Thermal-hydraulics

    SciTech Connect (OSTI)

    Tokuhiro, Akira; Jones, Byron

    2013-09-13

    This research project will address some of the principal technology issues related to sodium-cooled fast reactors (SFR), primarily the development and demonstration of ultrasonic measurement diagnostics linked to effective thermal convective sensing under normatl and off-normal conditions. Sodium is well-suited as a heat transfer medium for the SFR. However, because it is chemically reactive and optically opaque, it presents engineering accessibility constraints relative to operations and maintenance (O&M) and in-service inspection (ISI) technologies that are currently used for light water reactors. Thus, there are limited sensing options for conducting thermohydraulic measurements under normal conditions and off-normal events (maintenance, unanticipated events). Acoustic methods, primarily ultrasonics, are a key measurement technology with applications in non-destructive testing, component imaging, thermometry, and velocimetry. THis project would have yielded a better quantitative and qualitative understanding of the thermohydraulic condition of solium under varied flow conditions. THe scope of work will evaluate and demonstrate ultrasonic technologies and define instrumentation options for the SFR.

  20. Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

    SciTech Connect (OSTI)

    Gougar, Hans David

    2015-10-01

    The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each of the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.

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

    SciTech Connect (OSTI)

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

    2011-05-31

    Development and deployment of small-scale nuclear power reactors and their maintenance, monitoring, and control are part of the mission under the Small Modular Reactor (SMR) program. The objectives of this NERI-consortium research project are to investigate, develop, and validate advanced methods for sensing, controlling, monitoring, diagnosis, and prognosis of these reactors, and to demonstrate the methods with application to one of the proposed integral pressurized water reactors (IPWR). For this project, the IPWR design by Westinghouse, the International Reactor Secure and Innovative (IRIS), has been used to demonstrate the techniques developed under this project. The research focuses on three topical areas with the following objectives. Objective 1 - Develop and apply simulation capabilities and sensitivity/uncertainty analysis methods to address sensor deployment analysis and small grid stability issues. Objective 2 - Develop and test an autonomous and fault-tolerant control architecture and apply to the IRIS system and an experimental flow control loop, with extensions to multiple reactor modules, nuclear desalination, and optimal sensor placement strategy. Objective 3 - Develop and test an integrated monitoring, diagnosis, and prognosis system for SMRs using the IRIS as a test platform, and integrate process and equipment monitoring (PEM) and process and equipment prognostics (PEP) toolboxes. The research tasks are focused on meeting the unique needs of reactors that may be deployed to remote locations or to developing countries with limited support infrastructure. These applications will require smaller, robust reactor designs with advanced technologies for sensors, instrumentation, and control. An excellent overview of SMRs is described in an article by Ingersoll (2009). The article refers to these as deliberately small reactors. Most of these have modular characteristics, with multiple units deployed at the same plant site. Additionally, the topics focus

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

    SciTech Connect (OSTI)

    Frank Darmann; Robert Lombaerde; Franco Moriconi; Albert Nelson

    2011-10-31

    Zenergy Power has successfully designed, built, tested, and installed in the US electrical grid a saturable reactor Fault Current Limiter. Beginning in 2007, first as SC Power Systems and from 2008 as Zenergy Power, Inc., ZP used DOE matching grant and ARRA funds to help refine the design of the saturated reactor fault current limiter. ZP ultimately perfected the design of the saturated reactor FCL to the point that ZP could reliably design a suitable FCL for most utility applications. Beginning with a very basic FCL design using 1G HTS for a coil housed in a LN2 cryostat for the DC bias magnet, the technology progressed to a commercial system that was offered for sale internationally. Substantial progress was made in two areas. First, the cryogenics cooling system progressed from a sub-cooled liquid nitrogen container housing the HTS coils to cryostats utilizing dry conduction cooling and reaching temperatures down to less than 20 degrees K. Large, round cryostats with “warm bore” diameters of 1.7 meters enabled the design of large tanks to hold the AC components. Second, the design of the AC part of the FCL was refined from a six legged “spider” design to a more compact and lighter design with better fault current limiting capability. Further refinement of the flux path and core shape led to an efficient saturated reactor design requiring less Ampere-turns to saturate the core. In conclusion, the development of the saturable reactor FCL led to a more efficient design not requiring HTS magnets and their associated peripheral equipment, which yielded a more economical product in line with the electric utility industry expectations. The original goal for the DOE funding of the ZP project “Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters” was to stimulate the HTS wire industry with, first 1G, then 2G, HTS wire applications. Over the approximately 5 years of ZP’s product

  3. Building America Case Study: Demonstration House of Cold-Climate...

    Energy Savers [EERE]

    Demonstration House of Cold-Climate Solutions for Affordable Housing Minneapolis, ... America (BA) team that conducted a demonstration project to help nonproft developers ...

  4. Assigning Seismic Design Category to Large Reactors: A Case Study of the ATR

    Broader source: Energy.gov [DOE]

    Assigning Seismic Design Category to Large Reactors: A Case Study of the ATR Stuart Jensen October 21, 2014

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

    SciTech Connect (OSTI)

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

    2013-05-17

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

  6. The demonstration of continuous stirred tank reactor operations with high level waste

    SciTech Connect (OSTI)

    Peterson, R.A.

    2000-07-19

    This report contains the results of testing performed at the request of High Level Waste Engineering. These tests involved the operation of two continuous stirred tank reactors with high level waste.

  7. Business Case for Small Modular Reactors Report on Findings | Department of

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

    Energy Business Case for Small Modular Reactors Report on Findings Business Case for Small Modular Reactors Report on Findings This study assesses the market for SMRs and develops a business case to identify incentives, policies, and programs that can be effectively implemented and have significant impact on the commercialization of SMRs. Business Case for Small Modular Reactors Report on Findings to the U.S. Department of Energy (666.81 KB) More Documents & Publications A Strategic

  8. Continuous-flow stirred-tank reactor 20-L demonstration test: Final report

    SciTech Connect (OSTI)

    Lee, D.D.; Collins, J.L.

    2000-02-01

    One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test using the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required.

  9. Processing Tritiated Water at the Savannah River Site: A Production-Scale Demonstration of a Palladium Membrane Reactor

    SciTech Connect (OSTI)

    Sessions, Kevin L. [Westinghouse Savannah River Company (United States)

    2005-07-15

    The Palladium Membrane Reactor (PMR) process was installed in the Tritium Facilities at the Savannah River Site to perform a production-scale demonstration for the recovery of tritium from tritiated water adsorbed on molecular sieve (zeolite). Unlike the current recovery process that utilizes magnesium, the PMR offers a means to process tritiated water in a more cost effective and environmentally friendly manner. The design and installation of the large-scale PMR process was part of a collaborative effort between the Savannah River Site and Los Alamos National Laboratory.The PMR process operated at the Savannah River Site between May 2001 and April 2003. During the initial phase of operation the PMR processed thirty-four kilograms of tritiated water from the Princeton Plasma Physics Laboratory. The water was processed in fifteen separate batches to yield approximately 34,400 liters (STP) of hydrogen isotopes. Each batch consisted of round-the-clock operations for approximately nine days. In April 2003 the reactor's palladium-silver membrane ruptured resulting in the shutdown of the PMR process. Reactor performance, process performance and operating experiences have been evaluated and documented. A performance comparison between PMR and current magnesium process is also documented.

  10. Demonstration Assessment of Light-Emitting Diode (LED) Freezer Case Lighting

    SciTech Connect (OSTI)

    Rishman, E. E.; Tuenge, J. R.

    2009-10-01

    This report describes the process and results of a demonstration of solid-state lighting (SSL) technology combined with occupancy sensors in a set of upright grocery store freezer cases.

  11. Demonstrating and implementing innovative technologies: Case studies from the USDOE Office of Technology Development

    SciTech Connect (OSTI)

    Brouns, T.M.; Koegler, K.J.; Mamiya, L.S.

    1995-02-01

    This paper describes elements of success for demonstration, evaluation, and transfer for deployment of innovative technologies for environmental restoration. They have been compiled from lessons learned through the US Department of Energy (DOE) Office of Technology Development`s Volatile Organic Compounds in Arid Soil Integrated Demonstration (VOC-Arid ID). The success of the VOC-Arid ID program was determined by the rapid development demonstration, and transfer for deployment of technologies to operational sites that improve on safety, cost, and/or schedule of performance over baseline technologies. The VOC-Arid ID successfully fielded more than 25 innovative technology field demonstrations; several of the technologies demonstrated have been successfully transferred for deployment Field demonstration is a critical element in the successful transfer of innovative technologies into environmental restoration operations. The measures of success for technology demonstrations include conducting the demonstration in a safe and controlled environment and generating the appropriate information by which to evaluate the technology. However, field demonstrations alone do not guarantee successful transfer for deployment There are many key elements throughout the development and demonstration process that have a significant impact on the success of a technology. This paper presents key elements for a successful technology demonstration and transfer for deployment identified through the experiences of the VOC-Arid ID. Also, several case studies are provided as examples.

  12. The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

    SciTech Connect (OSTI)

    Nevinitsa, V. A. Dudnikov, A. A.; Blandinskiy, V. Yu.; Balanin, A. L.; Alekseev, P. N.; Titarenko, Yu. E.; Batyaev, V. F.; Pavlov, K. V.; Titarenko, A. Yu.

    2015-12-15

    A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing {sup 233}U from {sup 232}Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.

  13. An Overview of the Safety Case for Small Modular Reactors

    SciTech Connect (OSTI)

    Ingersoll, Daniel T

    2011-01-01

    Several small modular reactor (SMR) designs emerged in the late 1970s and early 1980s in response to lessons learned from the many technical and operational challenges of the large Generation II light-water reactors. After the accident at the Three Mile Island plant in 1979, an ensuing reactor redesign effort spawned the term inherently safe designs, which later evolved into passively safe terminology. Several new designs were engineered to be deliberately small in order to fully exploit the benefits of passive safety. Today, new SMR designs are emerging with a similar philosophy of offering highly robust and resilient designs with increased safety margins. Additionally, because these contemporary designs are being developed subsequent to the September 11, 2001, terrorist attack, they incorporate a number of intrinsic design features to further strengthen their safety and security. Several SMR designs are being developed in the United States spanning the full spectrum of reactor technologies, including water-, gas-, and liquid-metal-cooled ones. Despite a number of design differences, most of these designs share a common set of design principles to enhance plant safety and robustness, such as eliminating plant design vulnerabilities where possible, reducing accident probabilities, and mitigating accident consequences. An important consequence of the added resilience provided by these design approaches is that the individual reactor units and the entire plant should be able to survive a broader range of extreme conditions. This will enable them to not only ensure the safety of the general public but also help protect the investment of the owner and continued availability of the power-generating asset. Examples of typical SMR design features and their implications for improved plant safety are given for specific SMR designs being developed in the United States.

  14. Case Study for the ARRA-funded GSHP Demonstration at University at Albany

    SciTech Connect (OSTI)

    Liu, Xiaobing; Malhotra, Mini; Xiong, Zeyu

    2015-03-01

    High initial costs and lack of public awareness of ground-source heat pump (GSHP) technology are the two major barriers preventing rapid deployment of this energy-saving technology in the United States. Under the American Recovery and Reinvestment Act (ARRA), 26 GSHP projects have been competitively selected and carried out to demonstrate the benefits of GSHP systems and innovative technologies for cost reduction and/or performance improvement. This report highlights the findings of a case study of one of the ARRA-funded GSHP demonstration projects—a distributed GSHP system at a new 500-bed apartment-style student residence hall at the University at Albany. This case study is based on the analysis of detailed design documents, measured performance data, published catalog data of heat pump equipment, and actual construction costs. Simulations with a calibrated computer model are performed for both the demonstrated GSHP system and a baseline heating, ventilation, and airconditioning (HVAC) system to determine the energy savings and other related benefits achieved by the GSHP system. The evaluated performance metrics include the energy efficiency of the heat pump equipment and the overall GSHP system, as well as the pumping performance, energy savings, carbon emission reductions, and cost-effectiveness of the demonstrated GSHP system compared with the baseline HVAC system. This case study also identifies opportunities for improving the operational efficiency of the demonstrated GSHP system.

  15. REACTOR

    DOE Patents [OSTI]

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  16. REACTOR

    DOE Patents [OSTI]

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  17. Precision Information Environment (PIE) for International Safeguards: Pre-Demonstration Development Use Cases

    SciTech Connect (OSTI)

    Gastelum, Zoe N.; Henry, Michael J.

    2013-11-13

    In FY2013, the PIE International Safeguards team demonstrated our development progress to U.S. Department of Energy (DOE) staff from the Office of Nonproliferation and International Security (NA-24, our client) and the Office of Defense Nuclear Nonproliferation Research and Development (NA-22). Following the demonstration, the team was asked by our client to complete additional development prior to a planned demonstration at the International Atomic Energy Agency (IAEA), scheduled tentatively for January or spring of 2014. The team discussed four potential areas for development (in priority order), and will develop them as time and funding permit prior to an IAEA demonstration. The four capability areas are: 1. Addition of equipment manuals to PIE-accessible files 2. Optical character recognition (OCR) of photographed text 3. Barcode reader with information look-up from a database 4. Add Facilities to Data Model 5. Geospatial capabilities with information integration Each area will be described below in a use case.

  18. Case study for ARRA-funded ground-source heat pump (GSHP) demonstration at Oakland University

    SciTech Connect (OSTI)

    Im, Piljae; Liu, Xiaobing

    2015-09-01

    High initial costs and lack of public awareness of ground-source heat pump (GSHP) technology are the two major barriers preventing rapid deployment of this energy-saving technology in the United States. Under the American Recovery and Reinvestment Act (ARRA), 26 GSHP projects have been competitively selected and carried out to demonstrate the benefits of GSHP systems and innovative technologies for cost reduction and/or performance improvement. This paper highlights the findings of a case study of one of the ARRA-funded GSHP demonstration projects, a ground-source variable refrigerant flow (GS-VRF) system installed at the Human Health Building at Oakland University in Rochester, Michigan. This case study is based on the analysis of measured performance data, maintenance records, construction costs, and simulations of the energy consumption of conventional central heating, ventilation, and air-conditioning (HVAC) systems providing the same level of space conditioning as the demonstrated GS-VRF system. The evaluated performance metrics include the energy efficiency of the heat pump equipment and the overall GS-VRF system, pumping performance, energy savings, carbon emission reductions, and cost-effectiveness of the GS-VRF system compared with conventional HVAC systems. This case study also identified opportunities for reducing uncertainties in the performance evaluation, improving the operational efficiency, and reducing the installed cost of similar GSHP systems in the future.

  19. REACTOR

    DOE Patents [OSTI]

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  20. REACTORS

    DOE Patents [OSTI]

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  1. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

  2. Reactor

    DOE Patents [OSTI]

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  3. Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor

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

    Demonstration Case Study | Department of Energy (RISMC) Advanced Test Reactor Demonstration Case Study Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A

  4. Existing Homes Retrofit Case Study: SMUD's 32nd Avenue Remodel Demonstration Program, Sacramento, CA

    SciTech Connect (OSTI)

    none,

    2011-03-01

    This case study describes a retrofit project between SMUD and Building America to create a high-performance home.

  5. Existing Homes Retrofit Case Study: SMUD's Jean Avenue Remodel Demonstration Program, Sacramento, CA

    SciTech Connect (OSTI)

    none,

    2011-03-01

    This case study describes how SMUD teamed with Building America partners to retrofit aging, foreclosed homes into high-performance homes.

  6. Demonstration Assessment of Light-Emitting Diode (LED) Freezer Case Lighting

    Broader source: Energy.gov [DOE]

    This document is a report of observations and results obtained from a lighting demonstration project conducted under a U.S. Department of Energy program. The program supports demonstrations of high-performance solid-state lighting (SSL) products in order to develop empirical data and experience with in-the-field applications of this advanced lighting technology for plant-wide improvement.

  7. Case Study of The ARRA-Funded GSHP Demonstration at the Natural Sources Building, Montana Tech

    SciTech Connect (OSTI)

    Malhotra, Mini; Liu, Xiaobing

    2015-04-01

    Under the American Recovery and Reinvestment Act (ARRA), 26 ground source heat pump (GSHP) projects were competitively selected in 2009 to demonstrate the benefits of GSHP systems and innovative technologies for cost reduction and/or performance improvement. One of the selected demonstration projects was proposed by Montana Tech of the University of Montana for a 56,000 sq ft, newly constructed, on-campus research facility – the Natural Resources Building (NRB) located in Butte, Montana. This demonstrated GSHP system consists of a 50 ton water-to-water heat pump and a closed-loop ground heat exchanger with two redundant 7.5 hp constant-speed pumps to use water in the nearby flooded mines as a heat source or heat sink. It works in conjunction with the originally installed steam HX and an aircooled chiller to provide space heating and cooling. It is coupled with the existing hot water and chilled water piping in the building and operates in the heating or cooling mode based on the outdoor air temperature. The ground loop pumps operate in conjunction with the existing pumps in the building hot and chilled water loops for the operation of the heat pump unit. The goal of this demonstration project is to validate the technical and economic feasibility of the demonstrated commercial-scale GSHP system in the region, and illustrate the feasibility of using mine waters as the heat sink and source for GSHP systems. Should the demonstration prove satisfactory and feasible, it will encourage similar GSHP applications using mine water, thus help save energy and reduce carbon emissions. The actual performance of the system is analyzed with available measured data for January through July 2014. The annual energy performance is predicted and compared with a baseline scenario, with the heating and cooling provided by the originally designed systems. The comparison is made in terms of energy savings, operating cost savings, cost-effectiveness, and environmental benefits. Finally

  8. Evaluation of a Business Case for Safeguards by Design in Nuclear Power Reactors

    SciTech Connect (OSTI)

    Wood, Thomas W.; Seward, Amy M.; Lewis, Valerie A.; Gitau, Ernest TN; Zentner, Michael D.

    2012-12-01

    Safeguards by Design (SbD) is a well-known paradigm for consideration and incorporation of safeguards approaches and associated design features early in the nuclear facility development process. This paradigm has been developed as part of the Next Generation Safeguards Initiative (NGSI), and has been accepted as beneficial in many discussions and papers on NGSI or specific technologies under development within NGSI. The Office of Nuclear Safeguards and Security funded the Pacific Northwest National Laboratory to examine the business case justification of SbD for nuclear power reactors. Ultimately, the implementation of SbD will rely on the designers of nuclear facilities. Therefore, it is important to assess the incentives which will lead designers to adopt SbD as a standard practice for nuclear facility design. This report details the extent to which designers will have compelling economic incentives to adopt SbD.

  9. Posttraumatic tibia valga: a case demonstrating asymmetric activity at the proximal growth plate on technetium bone scan

    SciTech Connect (OSTI)

    Zionts, L.E.; Harcke, H.T.; Brooks, K.M.; MacEwen, G.D.

    1987-07-01

    Posttraumatic tibia valga is a well-recognized complication following fracture of the upper tibial metaphysis in young children. We present a case of a child who developed a valgus deformity following fracture of the proximal tibia and fibula in which quantitative bone scintigraphy at 5 months after injury demonstrated increased uptake at the proximal tibial growth plate with proportionally greater uptake on the medial side. This finding suggests that the valgus deformity in this patient was due to a relative increase in vascularity and consequent overgrowth of the medial portion of the proximal tibial physis.

  10. Comparing Simulation Results with Traditional PRA Model on a Boiling Water Reactor Station Blackout Case Study

    SciTech Connect (OSTI)

    Zhegang Ma; Diego Mandelli; Curtis Smith

    2011-07-01

    A previous study used RELAP and RAVEN to conduct a boiling water reactor station black-out (SBO) case study in a simulation based environment to show the capabilities of the risk-informed safety margin characterization methodology. This report compares the RELAP/RAVEN simulation results with traditional PRA model results. The RELAP/RAVEN simulation run results were reviewed for their input parameters and output results. The input parameters for each simulation run include various timing information such as diesel generator or offsite power recovery time, Safety Relief Valve stuck open time, High Pressure Core Injection or Reactor Core Isolation Cooling fail to run time, extended core cooling operation time, depressurization delay time, and firewater injection time. The output results include the maximum fuel clad temperature, the outcome, and the simulation end time. A traditional SBO PRA model in this report contains four event trees that are linked together with the transferring feature in SAPHIRE software. Unlike the usual Level 1 PRA quantification process in which only core damage sequences are quantified, this report quantifies all SBO sequences, whether they are core damage sequences or success (i.e., non core damage) sequences, in order to provide a full comparison with the simulation results. Three different approaches were used to solve event tree top events and quantify the SBO sequences: W process flag, default process flag without proper adjustment, and default process flag with adjustment to account for the success branch probabilities. Without post-processing, the first two approaches yield incorrect results with a total conditional probability greater than 1.0. The last approach accounts for the success branch probabilities and provides correct conditional sequence probabilities that are to be used for comparison. To better compare the results from the PRA model and the simulation runs, a simplified SBO event tree was developed with only four top

  11. Case Study for the ARRA-funded Ground Source Heat Pump Demonstration at Denver Museum of Nature & Science

    SciTech Connect (OSTI)

    Im, Piljae; Liu, Xiaobing

    2015-11-30

    High initial costs and lack of public awareness of ground-source heat pump (GSHP) technology are the two major barriers preventing rapid deployment of this energy-saving technology in the United States. Under the American Recovery and Reinvestment Act (ARRA), 26 GSHP projects were competitively selected and carried out to demonstrate the benefits of GSHP systems and innovative technologies for cost reduction and/or performance improvement. This report highlights the findings of a case study of one such GSHP demonstration projects that uses a recycled water heat pump (RWHP) system installed at the Denver Museum of Nature & Science in Denver, Colorado. The RWHP system uses recycled water from the city’s water system as the heat sink and source for a modular water-to-water heat pump (WWHP). This case study was conducted based on the available measured performance data from December 2014 through August 2015, utility bills of the building in 2014 and 2015, construction drawings, maintenance records, personal communications, and construction costs. The annual energy consumption of the RWHP system was calculated based on the available measured data and other related information. It was compared with the performance of a baseline scenario— a conventional VAV system using a water-cooled chiller and a natural gas fired boiler, both of which have the minimum energy efficiencies allowed by ASHRAE 90.1-2010. The comparison was made to determine energy savings, operating cost savings, and CO2 emission reductions achieved by the RWHP system. A cost analysis was performed to evaluate the simple payback of the RWHP system. Summarized below are the results of the performance analysis, the learned lessons, and recommended improvement in the operation of the RWHP system.

  12. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Starr, C.

    1963-01-01

    This patent relates to a combination useful in a nuclear reactor and is comprised of a casing, a mass of graphite irapregnated with U compounds in the casing, and at least one coolant tube extending through the casing. The coolant tube is spaced from the mass, and He is irtroduced irto the space between the mass and the coolant tube. (AEC)

  13. Results of tests to demonstrate a six-inch diameter coater for production of TRISO-coated particles for advanced gas reactor experiments

    SciTech Connect (OSTI)

    Barnes, Charles M; Marshall, Douglas W; Keeley, Joseph T; Hunn, John D

    2009-01-01

    The Next Generation Nuclear Plant (NGNP)/Advanced Gas Reactor (AGR) Fuel Development and Qualification Program includes a series of irradiation experiments in Idaho National Laboratory's (INL's) Advanced Test Reactor. TRISO-coated particles for the first AGR experiment, AGR-1, were produced at Oak Ridge National Laboratory (ORNL) in a two-inch diameter coater. A requirement of the NGNP/AGR Program is to produce coated particles for later experiments in coaters more representative of industrial scale. Toward this end, tests have been performed by Babcock and Wilcox (B&W) in a six-inch diameter coater. These tests are expected to lead to successful fabrication of particles for the second AGR experiment, AGR-2.

  14. Light Water Reactor Sustainability Program Support and Modeling for the Boiling Water Reactor Station Black Out Case Study Using RELAP and RAVEN

    SciTech Connect (OSTI)

    Diego Mandelli; Curtis Smith; Thomas Riley; John Schroeder; Cristian Rabiti; Aldrea Alfonsi; Joe Nielsen; Dan Maljovec; Bie Wang; Valerio Pascucci

    2013-09-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated. In order to evaluate the impact of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the impact of power uprate on the safety of a boiled water reactor system. The case study considered is a loss of off-site power followed by the loss of diesel generators, i.e., a station black out (SBO) event. Analysis is performed by using a thermo-hydraulic code, i.e. RELAP-5, and a stochastic analysis tool currently under development at INL, i.e. RAVEN. Starting from the event tree models contained in SAPHIRE, we built the input file for RELAP-5 that models in great detail system dynamics under SBO conditions. We also interfaced RAVEN with RELAP-5 so that it would be possible to run multiple RELAP-5 simulation runs by changing specific keywords of the input file. We both employed classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. We also employed advanced data analysis and visualization tools that helped us to correlate simulation outcome such as maximum core temperature with a set of input uncertain parameters. Results obtained gave a detailed overview of the issues associated to power uprate for a SBO accident scenario. We were able to quantify how timing of safety related events were impacted by a higher reactor core power. Such insights can provide useful material to the decision makers to perform risk-infomed safety margins management.

  15. RESULTS OF TESTS TO DEMONSTRATE A SIX-INCH DIAMETER COATER FOR PRODUCTION OF TRISO-COATED PARTICLES FOR ADVANCED GAS REACTOR EXPERIMENTS

    SciTech Connect (OSTI)

    Douglas W. Marshall

    2008-09-01

    The Next Generation Nuclear Plant (NGNP)/Advanced Gas Reactor (AGR) Fuel Development and Qualification Program includes a series of irradiation experiments in Idaho National Laboratory's (INL's) Advanced Test Reactor. TRISOcoated particles for the first AGR experiment, AGR-1, were produced at Oak Ridge National Laboratory (ORNL) in a twoinch diameter coater. A requirement of the NGNP/AGR Program is to produce coated particles for later experiments in coaters more representative of industrial scale. Toward this end, tests have been performed by Babcock and Wilcox (B&W) in a six-inch diameter coater. These tests are expected to lead to successful fabrication of particles for the second AGR experiment, AGR-2. While a thorough study of how coating parameters affect particle properties was not the goal of these tests, the test data obtained provides insight into process parameter/coated particle property relationships. Most relationships for the six-inch diameter coater followed trends found with the ORNL two-inch coater, in spite of differences in coater design and bed hydrodynamics. For example the key coating parameters affecting pyrocarbon anisotropy were coater temperature, coating gas fraction, total gas flow rate and kernel charge size. Anisotropy of the outer pyrolytic carbon (OPyC) layer also strongly correlates with coater differential pressure. In an effort to reduce the total particle fabrication run time, silicon carbide (SiC) was deposited with methyltrichlorosilane (MTS) concentrations up to 3 mol %. Using only hydrogen as the fluidizing gas, the high concentration MTS tests resulted in particles with lower than desired SiC densities. However when hydrogen was partially replaced with argon, high SiC densities were achieved with the high MTS gas fraction.

  16. RESULTS OF TESTS TO DEMONSTRATE A SIX-INCH-DIAMETER COATER FOR PRODUCTION OF TRISO-COATED PARTICLES FOR ADVANCED GAS REACTOR EXPERIMENTS

    SciTech Connect (OSTI)

    Charles M Barnes

    2008-09-01

    The Next Generation Nuclear Plant (NGNP)/Advanced Gas Reactor (AGR) Fuel Development and Qualification Program includes a series of irradiation experiments in Idaho National Laboratory’s (INL’s) Advanced Test Reactor. TRISOcoated particles for the first AGR experiment, AGR-1, were produced at Oak Ridge National Laboratory (ORNL) in a two inch diameter coater. A requirement of the NGNP/AGR Program is to produce coated particles for later experiments in coaters more representative of industrial scale. Toward this end, tests have been performed by Babcock and Wilcox (B&W) in a six-inch diameter coater. These tests are expected to lead to successful fabrication of particles for the second AGR experiment, AGR-2. While a thorough study of how coating parameters affect particle properties was not the goal of these tests, the test data obtained provides insight into process parameter/coated particle property relationships. Most relationships for the six-inch diameter coater followed trends found with the ORNL two-inch coater, in spite of differences in coater design and bed hydrodynamics. For example the key coating parameters affecting pyrocarbon anisotropy were coater temperature, coating gas fraction, total gas flow rate and kernel charge size. Anisotropy of the outer pyrolytic carbon (OPyC) layer also strongly correlates with coater differential pressure. In an effort to reduce the total particle fabrication run time, silicon carbide (SiC) was deposited with methyltrichlorosilane (MTS) concentrations up to 3 mol %. Using only hydrogen as the fluidizing gas, the high concentration MTS tests resulted in particles with lower than desired SiC densities. However when hydrogen was partially replaced with argon, high SiC densities were achieved with the high MTS gas fraction.

  17. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect (OSTI)

    R. Wigeland; J. Cahalan

    2009-09-01

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

  18. U.S. Department of Energy Provides Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel from Decommissioned Nuclear Power Reactor Sites

    Broader source: Energy.gov [DOE]

    Washington D.C. - The U.S. Department of Energy (DOE) today released its Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel from Decommissioned Nuclear Power...

  19. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

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

    SciTech Connect (OSTI)

    Santos, N. D.; Blaise, P.; Santamarina, A.

    2013-07-01

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

  1. Demonstration & Deployment

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

    Demonstration & Deployment Plenary Presentation Brian Duff May 20, 2013 2 | Bioenergy Technologies Office Demonstration & Deployment Peer Review Plenary * Introduction to the ...

  2. Case study of the regional manufacturers not participating in the manufactured housing RCDP (Residential Construction Demonstration Project)

    SciTech Connect (OSTI)

    Lee, A.D.; Baechler, M.C.

    1990-03-01

    To develop reliable alternatives for building energy-efficient homes, the Bonneville Power Administration (Bonneville) is conducting the Residential Construction Demonstration Project (RCDP). RCDP Cycle 2 is the subject of this report and involves manufactured housing (commonly called mobile homes) constructed to US Department of Housing and Urban Development (HUD) codes. The primary objectives of the RCDP include: develop conservation techniques and innovations, assess these techniques and innovations, and introduce exhibiting the potential to be cost-effective, reliable, and marketable. 1 ref., 1 tab.

  3. POWER REACTOR

    DOE Patents [OSTI]

    Zinn, W.H.

    1958-07-01

    A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

  4. GASIS demonstration

    SciTech Connect (OSTI)

    Vidas, E.H.

    1995-04-01

    A prototype of the GASIS database and retrieval software has been developed and is the subject of this poster session and computer demonstration. The prototype consists of test or preliminary versions of the GASIS Reservoir Data System and Source Directory datasets and the software for query and retrieval. The prototype reservoir database covers the Rocky Mountain region and contains the full GASIS data matrix (all GASIS data elements) that will eventually be included on the CD-ROM. It is populated for development purposes primarily by the information included in the Rocky Mountain Gas Atlas. The software has been developed specifically for GASIS using Foxpro for Windows. The application is an executable file that does not require Foxpro to run. The reservoir database software includes query and retrieval, screen display, report generation, and data export functions. Basic queries by state, basin, or field name will be assisted by scrolling selection lists. A detailed query screen will allow record selection on the basis of any data field, such as depth, cumulative production, or geological age. Logical operators can be applied to any-numeric data element or combination of elements. Screen display includes a {open_quotes}browse{close_quotes} display with one record per row and a detailed single record display. Datasets can be exported in standard formats for manipulation with other software packages. The Source Directory software will allow record retrieval by database type or subject area.

  5. Industrial Scale Demonstration of Smart Manufacturing Achieving...

    Energy Savers [EERE]

    ... Transformational IT infrastructure demonstration for manufacturing with potential ... furnaces is underway. Platform demonstration use case brings infrared camera data ...

  6. The IAEA Coordinated Research Program on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis: Description of the Benchmark Test Cases and Phases

    SciTech Connect (OSTI)

    Frederik Reitsma; Gerhard Strydom; Bismark Tyobeka; Kostadin Ivanov

    2012-10-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The uncertainties in the HTR analysis tools are today typically assessed with sensitivity analysis and then a few important input uncertainties (typically based on a PIRT process) are varied in the analysis to find a spread in the parameter of importance. However, one wish to apply a more fundamental approach to determine the predictive capability and accuracies of coupled neutronics/thermal-hydraulics and depletion simulations used for reactor design and safety assessment. Today there is a broader acceptance of the use of uncertainty analysis even in safety studies and it has been accepted by regulators in some cases to replace the traditional conservative analysis. Finally, there is also a renewed focus in supplying reliable covariance data (nuclear data uncertainties) that can then be used in uncertainty methods. Uncertainty and sensitivity studies are therefore becoming an essential component of any significant effort in data and simulation improvement. In order to address uncertainty in analysis and methods in the HTGR community the IAEA launched a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modelling early in 2012. The project is built on the experience of the OECD/NEA Light Water Reactor (LWR) Uncertainty Analysis in Best-Estimate Modelling (UAM) benchmark activity, but focuses specifically on the peculiarities of HTGR designs and its simulation requirements. Two benchmark problems were defined with the prismatic type design represented by the MHTGR-350 design from General Atomics (GA) while a 250 MW modular pebble bed design, similar to the INET (China) and indirect-cycle PBMR (South Africa) designs are also included. In the paper more detail on the benchmark cases, the different specific phases and tasks and the latest

  7. Small Modular Reactors Presentation to Secretary of Energy Advisory...

    Energy Savers [EERE]

    DOE Small Modular Reactor Program (SMR) Research, Development & Deployment (RD&D) to ... and demonstration of innovative reactor technologies and concepts ...

  8. NEAC Nuclear Reactor Technology (NRT) Subcommittee On the Planning...

    Energy Savers [EERE]

    On the Planning Study of Future TestDemonstration Reactors March 2, 2015 Final Given ... a planning study for an advanced testdemonstration reactor in the United States. ...

  9. Decline of radionuclides in the nearshore environment following nuclear reactor closure: A U.K. case study

    SciTech Connect (OSTI)

    Cundy, A.B.; Croudace, I.W.; Warwick, P.E.; Bains, M.E.D.

    1999-09-01

    Radioactive discharges from nuclear facilities are frequently made into the marine environment and their fate during and after cessation of discharges is a matter of interest and concern. This study examines the decline of the radionuclides {sup 60} and {sup 65}Zn along the southern UK. coast, over the per 1988--1998, following the closure of the steam-generating heavy water (SGHW) reactor at AEA Winfrith, Dorset, UK. {sup 60}Co and {sup 65}Zn (and other activation products such as {sup 63}Ni and {sup 55}Fe) were widely dispersed in the marine environment off the central south coast of England, due to authorized releases from AEA Winfrith. Significant interaction occurred with clay-rich sediments and biota. A general exponential decline in {sup 60}Co activities (and in {sup 65}Zn activity) is found in intertidal mudflat sediments, seaweed and marine fauna in different areas along the south coast following closure of the reactor in 1990. Effective half-lives are determined which vary from 1 to 4 years in surface sediments ({sup 60}Co only), 1--4 years in seaweed and 0.5--2.5 years in crustaceans, bivalves and molluscs. Physical mixing and bioturbation largely control the rate at which {sup 60}Co declines in surface sediments. Both {sup 60}Co and {sup 65}Zn show a relatively slow rate of decline in seaweed and in marine fauna, showing that even after the virtual cessation of discharge from nuclear facilities, contamination of these organisms may persist for a number of years, albeit at reduced activities. Reasons for this persistence are likely to include absorption of radionuclides from sediment, and release and recycling of radionuclides via breakdown of contaminated organic material.

  10. BOILING REACTORS

    DOE Patents [OSTI]

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  11. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Daniels, F.

    1959-10-27

    A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

  12. Product Demonstrations | Department of Energy

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

    Product Demonstrations Product Demonstrations The Consortium will pursue a number of demonstrations following the general procedure used by DOE's GATEWAY demonstration program. Specific products to be featured in a demonstration may be selected by the host site or may be suggested for a given installation by the Consortium based on the product's anticipated performance in that installation. In the latter case, products will be evaluated for suitability based on performance relative to other

  13. CONVECTION REACTOR

    DOE Patents [OSTI]

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  14. Reactor Safety Research Programs

    SciTech Connect (OSTI)

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  15. The impact of fuel cladding failure events on occupational radiation exposures at nuclear power plants: Case study, PWR (pressurized-water reactor) during an outage

    SciTech Connect (OSTI)

    Moeller, M.P.; Martin, G.F.; Kenoyer, J.L.

    1987-08-01

    This report is the second in a series of case studies designed to evaluate the magnitude of increase in occupational radiation exposures at commercial US nuclear power plants resulting from small incidents or abnormal events. The event evaluated is fuel cladding failure, which can result in elevated primary coolant activity and increased radiation exposure rates within a plant. For this case study, radiation measurements were made at a pressurized-water reactor (PWR) during a maintenance and refueling outage. The PWR had been operating for 22 months with fuel cladding failure characterized as 105 pin-hole leakers, the equivalent of 0.21% failed fuel. Gamma spectroscopy measurements, radiation exposure rate determinations, thermoluminescent dosimeter (TLD) assessments, and air sample analyses were made in the plant's radwaste, pipe penetration, and containment buildings. Based on the data collected, evaluations indicate that the relative contributions of activation products and fission products to the total exposure rates were constant over the duration of the outage. This constancy is due to the significant contribution from the longer-lived isotopes of cesium (a fission product) and cobalt (an activation product). For this reason, fuel cladding failure events remain as significant to occupational radiation exposure during an outage as during routine operations. As documented in the previous case study (NUREG/CR-4485 Vol. 1), fuel cladding failure events increased radiation exposure rates an estimated 540% at some locations of the plant during routine operations. Consequently, such events can result in significantly greater radiation exposure rates in many areas of the plant during the maintenance and refueling outages than would have been present under normal fuel conditions.

  16. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fraas, A.P.; Mills, C.B.

    1961-11-21

    A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

  17. REACTOR COOLING

    DOE Patents [OSTI]

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  18. Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Assessment of High Value Surveillance Materials

    Office of Energy Efficiency and Renewable Energy (EERE)

    The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the...

  19. Navajo Electrification Demonstration Program

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

    Future Plans * Navajo Electrification Demonstration Program -Video OBJECTIVES OBJECTIVES " ... Navajo Electrification Demonstration Navajo Electrification Demonstration Program Program ...

  20. Piqua, Ohio, Decommissioned Reactor Site

    Office of Legacy Management (LM)

    cooled and moderated, thermal reactor was built by the U.S. Atomic Energy Commission (AEC)-predecessor agency to the U.S. Department of Energy (DOE)-as a demonstration project. ...

  1. Novel Catalytic Membrane Reactors

    SciTech Connect (OSTI)

    Stuart Nemser, PhD

    2010-10-01

    There are many industrial catalytic organic reversible reactions with amines or alcohols that have water as one of the products. Many of these reactions are homogeneously catalyzed. In all cases removal of water facilitates the reaction and produces more of the desired chemical product. By shifting the reaction to right we produce more chemical product with little or no additional capital investment. Many of these reactions can also relate to bioprocesses. Given the large number of water-organic compound separations achievable and the ability of the Compact Membrane Systems, Inc. (CMS) perfluoro membranes to withstand these harsh operating conditions, this is an ideal demonstration system for the water-of-reaction removal using a membrane reactor. Enhanced reaction synthesis is consistent with the DOE objective to lower the energy intensity of U.S. industry 25% by 2017 in accord with the Energy Policy Act of 2005 and to improve the United States manufacturing competitiveness. The objective of this program is to develop the platform technology for enhancing homogeneous catalytic chemical syntheses.

  2. Imaging Fukushima Daiichi reactors with muons

    SciTech Connect (OSTI)

    Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.; Milner, Edward C.; Morris, Christopher L.; Lukic, Zarija; Masuda, Koji; Perry, John O.

    2013-05-15

    A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi to make this determination in the near future.

  3. Advanced Reactor Research and Development Funding Opportunity...

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

    Nuclear Energy (NE) sponsors a program of research, development, and demonstration related to advanced non-light water reactor concepts. A goal of the program is to facilitate...

  4. Advanced Reactor Technologies | Department of Energy

    Energy Savers [EERE]

    demonstrate a new generation of gas-cooled, accident-tolerant reactors in the United States. ... on TRISO coated particle fuels, materials, design methods, and user applications. ...

  5. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Moore, R.V.; Bowen, J.H.; Dent, K.H.

    1958-12-01

    A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

  6. Thermionic Reactor Design Studies

    SciTech Connect (OSTI)

    Schock, Alfred

    1994-08-01

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

  7. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.

    1960-04-01

    A nuclear reactor is described consisting of blocks of graphite arranged in layers, natural uranium bodies disposed in holes in alternate layers of graphite blocks, and coolant tubes disposed in the layers of graphite blocks which do not contain uranium.

  8. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

    1959-03-24

    A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

  9. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Wigner, E.P.

    1960-11-22

    A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

  10. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  11. REACTOR SHIELD

    DOE Patents [OSTI]

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  12. Reactor apparatus

    DOE Patents [OSTI]

    Echtler, J. Paul

    1981-01-01

    A reactor apparatus for hydrocracking a polynuclear aromatic hydrocarbonaceous feedstock to produce lighter hydrocarbon fuels by contacting the hydrocarbonaceous feedstock with hydrogen in the presence of a molten metal halide catalyst.

  13. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  14. Tightly Coupled Multiphysics Algorithm for Pebble Bed Reactors

    SciTech Connect (OSTI)

    HyeongKae Park; Dana Knoll; Derek Gaston; Richard Martineau

    2010-10-01

    We have developed a tightly coupled multiphysics simulation tool for the pebble-bed reactor (PBR) concept, a type of Very High-Temperature gas-cooled Reactor (VHTR). The simulation tool, PRONGHORN, takes advantages of the Multiphysics Object-Oriented Simulation Environment library, and is capable of solving multidimensional thermal-fluid and neutronics problems implicitly with a Newton-based approach. Expensive Jacobian matrix formation is alleviated via the Jacobian-free Newton-Krylov method, and physics-based preconditioning is applied to minimize Krylov iterations. Motivation for the work is provided via analysis and numerical experiments on simpler multiphysics reactor models. We then provide detail of the physical models and numerical methods in PRONGHORN. Finally, PRONGHORN's algorithmic capability is demonstrated on a number of PBR test cases.

  15. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Breden, C.R.; Dietrich, J.R.

    1961-06-20

    A water-soluble non-volatile poison may be introduced into a reactor to nullify excess reactivity. The poison is removed by passing a side stream of the water containing the soluble poison to an evaporation chamber. The vapor phase is returned to the reactor to decrease the concentration of soluble poison and the liquid phase is returned to increase the concentration of soluble poison.

  16. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Vernon, H.C.

    1959-01-13

    A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

  17. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  18. Manufacturing Demonstration Facility

    Energy Savers [EERE]

    of Energy Manufacturing Demonstration Facility DOE Advanced Manufacturing Office Merit Review Craig Blue Director, Manufacturing Demonstration Facility Energy and ...

  19. Product Recovery from HTGR Reactor Fuel Processing Salt Official...

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

    Demonstration of Fuel and Fission Product Recovery from HTGR Reactor Fuel Processing Salt ... HTGR, MST, CST Retention: Permanent Demonstration of Fuel and Fission Product Recovery ...

  20. Small Reactor for Deep Space Exploration

    SciTech Connect (OSTI)

    none,

    2012-11-29

    This is the first demonstration of a space nuclear reactor system to produce electricity in the United States since 1965, and an experiment demonstrated the first use of a heat pipe to cool a small nuclear reactor and then harvest the heat to power a Stirling engine at the Nevada National Security Site's Device Assembly Facility confirms basic nuclear reactor physics and heat transfer for a simple, reliable space power system.

  1. Small Reactor for Deep Space Exploration

    ScienceCinema (OSTI)

    none,

    2014-05-30

    This is the first demonstration of a space nuclear reactor system to produce electricity in the United States since 1965, and an experiment demonstrated the first use of a heat pipe to cool a small nuclear reactor and then harvest the heat to power a Stirling engine at the Nevada National Security Site's Device Assembly Facility confirms basic nuclear reactor physics and heat transfer for a simple, reliable space power system.

  2. Demonstration Assessment of Light-Emitting Diode (LED) Freezer...

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

    Demonstration Assessment of Light-Emitting Diode (LED) Freezer Case Lighting Demonstration Assessment of Light-Emitting Diode (LED) Freezer Case Lighting This document is a report ...

  3. Challenges in the Development of Advanced Reactors

    SciTech Connect (OSTI)

    P. Sabharwall; M.C. Teague; S.M. Bragg-Sitton; M.W. Patterson

    2012-08-01

    Past generations of nuclear reactors have been successively developed and the next generation is currently being developed, demonstrating the constant progress and technical and industrial vitality of nuclear energy. In 2000 US Department of Energy launched Generation IV International Forum (GIF) which is one of the main international frameworks for the development of future nuclear systems. The six systems that were selected were: sodium cooled fast reactor, lead cooled fast reactor, supercritical water cooled reactor, very high temperature gas cooled reactor (VHTR), gas cooled fast reactor and molten salt reactor. This paper discusses some of the proposed advanced reactor concepts that are currently being researched to varying degrees in the United States, and highlights some of the major challenges these concepts must overcome to establish their feasibility and to satisfy licensing requirements.

  4. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

    1958-09-01

    This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

  5. Demonstration Assessment of LED Freezer Case Lighting

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

    ... Measurements were taken with a Photo Research Luminance ... The resolution of the loggers was set to a reading level of ... energy road mapping provided by the IES handbook 3 . ...

  6. Smart Grid Demonstration Project

    SciTech Connect (OSTI)

    Miller, Craig; Carroll, Paul; Bell, Abigail

    2015-03-11

    The National Rural Electric Cooperative Association (NRECA) organized the NRECA-U.S. Department of Energy (DOE) Smart Grid Demonstration Project (DE-OE0000222) to install and study a broad range of advanced smart grid technologies in a demonstration that spanned 23 electric cooperatives in 12 states. More than 205,444 pieces of electronic equipment and more than 100,000 minor items (bracket, labels, mounting hardware, fiber optic cable, etc.) were installed to upgrade and enhance the efficiency, reliability, and resiliency of the power networks at the participating co-ops. The objective of this project was to build a path for other electric utilities, and particularly electrical cooperatives, to adopt emerging smart grid technology when it can improve utility operations, thus advancing the co-ops’ familiarity and comfort with such technology. Specifically, the project executed multiple subprojects employing a range of emerging smart grid technologies to test their cost-effectiveness and, where the technology demonstrated value, provided case studies that will enable other electric utilities—particularly electric cooperatives— to use these technologies. NRECA structured the project according to the following three areas: Demonstration of smart grid technology; Advancement of standards to enable the interoperability of components; and Improvement of grid cyber security. We termed these three areas Technology Deployment Study, Interoperability, and Cyber Security. Although the deployment of technology and studying the demonstration projects at coops accounted for the largest portion of the project budget by far, we see our accomplishments in each of the areas as critical to advancing the smart grid. All project deliverables have been published. Technology Deployment Study: The deliverable was a set of 11 single-topic technical reports in areas related to the listed technologies. Each of these reports has already been submitted to DOE, distributed to co-ops, and

  7. CALDERON COKEMAKING PROCESS/DEMONSTRATION PROJECT

    SciTech Connect (OSTI)

    Albert Calderon

    1998-12-23

    This project deals with the demonstration of a coking process using proprietary technology of Calderon, with the following objectives geared to facilitate commercialization: (1) making coke of such quality as to be suitable for use in hard-driving, large blast furnaces; (2) providing proof that such process is continuous and environmentally closed to prevent emissions; (3) demonstrating that high-coking-pressure (non-traditional) coal blends which cannot be safely charged into conventional by-product coke ovens can be used in the Calderon process; and (4) demonstrating that coke can be produced economically, at a level competitive with coke imports. The activities of the past quarter were focused on the following: Conducting bench-scale tests to produce coke and acceptable tar from the process to satisfy Koppers, a prospective stakeholder; Consolidation of the project team players to execute the full size commercial cokemaking reactor demonstration; and Progress made in advancing the design of the full size commercial cokemaking reactor.

  8. Bioconversion reactor

    DOE Patents [OSTI]

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  9. Catalytic reactor

    DOE Patents [OSTI]

    Aaron, Timothy Mark; Shah, Minish Mahendra; Jibb, Richard John

    2009-03-10

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  10. REACTOR CONTROL

    DOE Patents [OSTI]

    Fortescue, P.; Nicoll, D.

    1962-04-24

    A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

  11. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Wigner, E.P.; Young, G.J.

    1958-10-14

    A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

  12. (Reactor dosimetry)

    SciTech Connect (OSTI)

    West, C.D.

    1990-09-13

    The lead in most aspects of research reactor design and use passed from the USA about 15 years ago, soon after the construction of the HFIR and HFBR. The Europeans have consistently upgraded and improved their existing facilities and have built new ones including the HFR at Grenoble and ORPHEE at Saclay. They studied ultra-high flux concepts ({approximately}10{sup 20}/m{sup {minus}2}{center dot}s{sup {minus}1}) about 10 years ago, and are in the design phase of a new, highly efficient medium flux reactor to be built at Garching, near Munich in Germany. A visit was made to Interatom, the firm -- the equivalent of the Architect/Engineer for the ANS project -- responsible, under contract to the Technical University of Munich, for the new Munich reactor design. There are many similarities to the ANS design, and we reviewed and discussed technical and safety aspects of the two reactors. A request was made for some new, hitherto proprietary, experimental data on reactor thermal hydraulics and cooling that will be very valuable to the ANS project. I presented a seminar on the ANS project. A visit was made to Kernforschungszentrum Karlsruhe and knowledge was gained from Dr. Kuchle, a true pioneer of ultra-high flux reactor concepts, of their work. Dr. Kuchle kindly reviewed the ANS reference core and cooling system design (with favorable conclusions). I then talked with researchers working on materials irradiation damage and activation of structural materials by neutron irradiation, both key issues for the ANS. I was shown some new techniques they have developed for testing materials irradiation effects at high fluences, in a short time, using accelerated particle beams.

  13. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.; Szilard, L.

    1957-09-24

    Reactors of the type employing plates of natural uranium in a moderator are discussed wherein the plates are um-formly disposed in parallel relationship to each other thereby separating the moderator material into distinct and individual layers. Each plate has an uninterrupted sunface area substantially equal to the cross-sectional area of the active portion of the reactor, the particular size of the plates and the volume ratio of moderator to uranium required to sustain a chain reaction being determinable from the known purity of these materials and other characteristics such as the predictable neutron losses due to the formation of radioactive elements of extremely high neutron capture cross section.

  14. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  15. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Young, G.

    1963-01-01

    This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

  16. Fueling of tandem mirror reactors

    SciTech Connect (OSTI)

    Gorker, G.E.; Logan, B.G.

    1985-01-01

    This paper summarizes the fueling requirements for experimental and demonstration tandem mirror reactors (TMRs), reviews the status of conventional pellet injectors, and identifies some candidate accelerators that may be needed for fueling tandem mirror reactors. Characteristics and limitations of three types of accelerators are described; neutral beam injectors, electromagnetic rail guns, and laser beam drivers. Based on these characteristics and limitations, a computer module was developed for the Tandem Mirror Reactor Systems Code (TMRSC) to select the pellet injector/accelerator combination which most nearly satisfies the fueling requirements for a given machine design.

  17. Robotic inspection of PWR coolant pump casing welds

    SciTech Connect (OSTI)

    Pratt, W.R.; Alford, J.W.; Davis, J.B.

    1997-12-01

    As of January 1, 1995, the Swedish Nuclear Inspectorate began requiring more thorough inspections of cast stainless-steel components in nuclear power plants, including pressurized water reactor (PWR) reactor coolant pump (RCP) casings. The examination requirements are established by fracture mechanics analyses of component weldments and demonstrated test system detection capabilities. This may include full volumetric inspection or some portion thereof. Ringhals station is a four-unit nuclear power plant, owned and operated by the Swedish State Power Board, Vattenfall. Unit 1 is a boiling water reactor. Units 2, 3, and 4 are Westinghouse-designed PWRs, ranging in size from 795 to 925 MW. The RCP casings at the PWR units are made of cast stainless steel and contain four circumferential welds that require inspection. Due to the thickness of the casings at the weld locations and configuration and surface conditions on the outside diameter of the casings, remote inspection from the inside diameter of the pump casing was mandated.

  18. Technology Demonstration Partnership Policy

    Broader source: Energy.gov [DOE]

    This City Council memorandum establishes a framework for engaging in and evaluating demonstration partnerships with the goal of developing, testing, and demonstrating emerging technologies, product, and service innovations.

  19. Demonstration & Market Transformation

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

    Demonstration & Market Transformation Peer Review Break-Out Presentation Jim Spaeth Program Manager Demonstration & Market Transformation March 23, 2015 2 | Bioenergy Technologies ...

  20. Radiation Emergency Procedure Demonstrations

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

    Dressing To Prevent the Spread of Radioactive Contamination This demonstration shows how ... Preparing The Area This demonstration shows basic steps you can take to gather equipment ...

  1. Fast Reactor Technology Preservation

    SciTech Connect (OSTI)

    Wootan, David W.; Omberg, Ronald P.

    2008-01-11

    There is renewed worldwide interest in developing and implementing a new generation of advanced fast reactors. International cooperative efforts are underway such as the Global Nuclear Energy Partnership (GNEP). Advanced computer modeling and simulation efforts are a key part of these programs. A recognized and validated set of Benchmark Cases are an essential component of such modeling efforts. Testing documentation developed during the operation of the Fast Flux Test Facility (FFTF) provide the information necessary to develop a very useful set of Benchmark Cases.

  2. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

    1983-01-01

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  3. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Anderson, H.L.

    1958-10-01

    The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

  4. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

    1959-10-27

    BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

  5. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J.

    1976-08-17

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

  6. Neutronic reactor

    DOE Patents [OSTI]

    Carleton, John T.

    1977-01-25

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment.

  7. Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point...

    Office of Scientific and Technical Information (OSTI)

    ... Authors: Qualls, A. L. 1 ; Betzler, Benjamin R. 1 ; Carbajo, Juan 1 ; Hale, Richard Edward 1 ; Harrison, Thomas J. 1 ; Powers, Jeffrey J. 1 ; Robb, Kevin R. 1 ; ...

  8. LIMB Demonstration Project Extension and Coolside Demonstration

    SciTech Connect (OSTI)

    Goots, T.R.; DePero, M.J.; Nolan, P.S.

    1992-11-10

    This report presents results from the limestone Injection Multistage Burner (LIMB) Demonstration Project Extension. LIMB is a furnace sorbent injection technology designed for the reduction of sulfur dioxide (SO[sub 2]) and nitrogen oxides (NO[sub x]) emissions from coal-fired utility boilers. The testing was conducted on the 105 Mwe, coal-fired, Unit 4 boiler at Ohio Edison's Edgewater Station in Lorain, Ohio. In addition to the LIMB Extension activities, the overall project included demonstration of the Coolside process for S0[sub 2] removal for which a separate report has been issued. The primary purpose of the DOE LIMB Extension testing, was to demonstrate the generic applicability of LIMB technology. The program sought to characterize the S0[sub 2] emissions that result when various calcium-based sorbents are injected into the furnace, while burning coals having sulfur content ranging from 1.6 to 3.8 weight percent. The four sorbents used included calcitic limestone, dolomitic hydrated lime, calcitic hydrated lime, and calcitic hydrated lime with a small amount of added calcium lignosulfonate. The results include those obtained for the various coal/sorbent combinations and the effects of the LIMB process on boiler and plant operations.

  9. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Grebe, J.J.

    1959-12-15

    A reactor which is particularly adapted tu serve as a heat source for a nuclear powered alrcraft or rocket is described. The core of this reactor consists of a porous refractory modera;or body which is impregnated with fissionable nuclei. The core is designed so that its surface forms tapered inlet and outlet ducts which are separated by the porous moderator body. In operation a gaseous working fluid is circulated through the inlet ducts to the surface of the moderator, enters and passes through the porous body, and is heated therein. The hot gas emerges into the outlet ducts and is available to provide thrust. The principle advantage is that tremendous quantities of gas can be quickly heated without suffering an excessive pressure drop.

  10. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Treshow, M.

    1958-08-19

    A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

  11. REACTOR UNLOADING

    DOE Patents [OSTI]

    Leverett, M.C.

    1958-02-18

    This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

  12. REACTOR MONITORING

    DOE Patents [OSTI]

    Bugbee, S.J.; Hanson, V.F.; Babcock, D.F.

    1959-02-01

    A neutron density inonitoring means for reactors is described. According to this invention a tunnel is provided beneath and spaced from the active portion of the reactor and extends beyond the opposite faces of the activc portion. Neutron beam holes are provided between the active portion and the tunnel and open into the tunnel near the middle thereof. A carriage operates back and forth in the tunnel and is adapted to convey a neutron detector, such as an ion chamber, and position it beneath one of the neutron beam holes. This arrangement affords convenient access of neutron density measuring instruments to a location wherein direct measurement of neutron density within the piles can be made and at the same time affords ample protection to operating personnel.

  13. Nuclear reactor

    DOE Patents [OSTI]

    Wade, Elman E.

    1979-01-01

    A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

  14. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Wade, E.J.

    1958-09-16

    This patent relates to a reflector means for a neutronic reactor. A reflector comprised of a plurality of vertically movable beryllium control members is provided surrounding the sides of the reactor core. An absorber of fast neutrons comprised of natural uramum surrounds the reflector. An absorber of slow neutrons surrounds the absorber of fast neutrons and is formed of a plurality of beryllium blocks having natural uranium members distributcd therethrough. in addition, a movable body is positioned directly below the core and is comprised of a beryllium reflector and an absorbing member attached to the botiom thereof, the absorbing member containing a substance selected from the goup consisting of natural urantum and Th/sup 232/.

  15. NUCLEAR REACTORS

    DOE Patents [OSTI]

    Long, E.; Ashby, J.W.

    1958-09-16

    ABS>A graphite moderator structure is presented for a nuclear reactor compriscd of an assembly of similarly orientated prismatic graphite blocks arranged on spaced longitudinal axes lying in common planes wherein the planes of the walls of the blocks are positioned so as to be twisted reintive to the planes of said axes so thatthe unlmpeded dtrect paths in direction wholly across the walls of the blocks are limited to the width of the blocks plus spacing between the blocks.

  16. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Anderson, J.B.

    1960-01-01

    A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.

  17. Neutronic reactor

    DOE Patents [OSTI]

    Lewis, Warren R.

    1978-05-30

    A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

  18. Newberry EGS Demonstration

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

    Newberry EGS Demonstration Susan Petty, PI AltaRock Energy EGS Demonstration Projects Project Officer: Lauren Boyd Total Project Funding: 43.8 m April 22, 2013 This presentation ...

  19. REACTOR CONTROL

    DOE Patents [OSTI]

    Ruano, W.J.

    1957-12-10

    This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

  20. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Peng, Yueng-Kay M.

    1989-04-04

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  1. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Peng, Yueng-Kay M.

    1989-01-01

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  2. Nuclear Reactor Technology Subcommittee of NEAC

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

    NEAC Mike Corradini (UW), Chair Ashok Bhatnagar (FPL), Doug Chapin (NPR), Tom Cochran (NRDC), Regis Matzie (Consultant) , Harold Ray (Consultant), Joy Rempe (Consultant) Nuclear Energy Advisory Committee Meeting December 11, 2015 1 Subcommittee Scope * Congress appropriated funds for "an advanced test/demonstration reactor planning study by the national laboratories, industry, and relevant stakeholders of such a reactor in the U.S. The study will evaluate advanced reactor technology

  3. Strategy Guideline. Demonstration Home

    SciTech Connect (OSTI)

    Hunt, A.; Savage, C.

    2012-12-01

    This guideline will provide a general overview of the different kinds of demonstration home projects, a basic understanding of the different roles and responsibilities involved in the successful completion of a demonstration home, and an introduction into some of the lessons learned from actual demonstration home projects. Also, this guideline will specifically look at the communication methods employed during demonstration home projects. And lastly, we will focus on how to best create a communication plan for including an energy efficient message in a demonstration home project and carry that message to successful completion.

  4. Strategy Guideline: Demonstration Home

    SciTech Connect (OSTI)

    Savage, C.; Hunt, A.

    2012-12-01

    This guideline will provide a general overview of the different kinds of demonstration home projects, a basic understanding of the different roles and responsibilities involved in the successful completion of a demonstration home, and an introduction into some of the lessons learned from actual demonstration home projects. Also, this guideline will specifically look at the communication methods employed during demonstration home projects. And lastly, we will focus on how to best create a communication plan for including an energy efficient message in a demonstration home project and carry that message to successful completion.

  5. Data surety demonstrations

    SciTech Connect (OSTI)

    Draelos, T.; Harris, M.; Herrington, P.; Kromer, D.

    1998-08-01

    The use of data surety within the International Monitoring System (IMS) is designed to offer increased trust of acquired sensor data at a low cost. The demonstrations discussed in the paper illustrate the feasibility of hardware authentication for sensor data and commands in a retrofit environment and a new system and of the supporting key management system. The individual demonstrations which are summarized in the paper are: (1) demonstration of hardware authentication for communication authentication in a retrofit environment; (2)demonstration of hardware authentication in a new system; and (3) demonstration of key management for sensor data and command authentication.

  6. B Reactor | Department of Energy

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

    Operational Management History Manhattan Project Signature Facilities B Reactor B Reactor B Reactor Completed in September 1944, the B Reactor was the world's first ...

  7. Nuclear reactor

    DOE Patents [OSTI]

    Pennell, William E.; Rowan, William J.

    1977-01-01

    A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

  8. NUCLEAR REACTORS

    DOE Patents [OSTI]

    Long, E.; Ashley, J.W.

    1958-12-16

    A graphite moderator structure is described for a gas-cooled nuclear reactor having a vertical orlentation wherein the structure is physically stable with regard to dlmensional changes due to Wigner growth properties of the graphite, and leakage of coolant gas along spaces in the structure is reduced. The structure is comprised of stacks of unlform right prismatic graphite blocks positioned in layers extending in the direction of the lengths of the blocks, the adjacent end faces of the blocks being separated by pairs of tiles. The blocks and tiles have central bores which are in alignment when assembled and are provided with cooperatlng keys and keyways for physical stability.

  9. Photocatalytic reactor

    DOE Patents [OSTI]

    Bischoff, B.L.; Fain, D.E.; Stockdale, J.A.D.

    1999-01-19

    A photocatalytic reactor is described for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane. 4 figs.

  10. Commercial office daylighting demonstration

    SciTech Connect (OSTI)

    Pike, T.F.; Rizzuto, J.

    1981-01-01

    The results of a commerical office photoelectrically controlled, dimmable lighting demonstration in New York City have shown that daylighting can be used to conserve energy and limit peak electrical demand. In this demonstration, three photo-electrically controlled dimming systems were compared side-by-side to test the concept of independent dimming of banks of luminaires controlled with individual photocells.

  11. CALDERON COKEMAKING PROCESS/DEMONSTRATION PROJECT

    SciTech Connect (OSTI)

    ALBERT CALDERON

    1998-06-22

    This project deals with the demonstration of a coking process using proprietary technology of Calderon, with the following objectives geared to facilitating commercialization: (1) making coke of such quality as to be suitable for use in hard-driving, large blast furnaces; (2) providing proof that such process is continuous and environmentally closed to prevent emissions; (3) demonstrating that high-coking-pressure (non-traditional) coal blends which cannot be safely charged into conventional by-product coke ovens can be used in the Calderon process; and (4) demonstrating that coke can be produced economically, at a level competitive with coke imports. The activities of the past quarter were focused on three main activities: Continuation of design of the coking reactor; Raising funds from the private sector; and Detailed analysis of the tests conducted in Alliance, Ohio. The design of the reactor work centered on the provision for the capability to inspect and maintain the internals of the reactor. The activities relating to raising funds from the steel industry have been fruitful. Bethlehem Steel has agreed to contribute funds. The collected data from the tests at Alliance were analyzed and a detailed report was completed and presented to the International Iron & Steel Institute by invitation.

  12. Path to a commercial fast reactor option in the United States

    SciTech Connect (OSTI)

    Murray, P.; McCullum, R.; Nesbit, S.; Sowder, A.; Stout, D.

    2012-07-01

    Fast reactors represent one technology with the potential to transform the energy sector through increased resource utilization and unique waste management opportunities that are not available through the established light water reactor (LWR) technology. The technology is 'old' by 21 st century standards, as the United States and other nations have funded fast reactor research and development efforts from the birth of nuclear power that have led to deployment at all scales from experimental to commercial. Yet, in spite of great promise, the experience with fast reactors can be characterized as mixed, with both notable successes and failures. Why then, after 50 years of development, is there no commercial fast reactor operating in the United States? Is commercialization of the technology possible? If so, what business case is needed for a commercial utility to operate a fast reactor? This paper will attempt to sketch a possible path for the development and demonstration of technology needed to make fast reactors a real option for commercial operation by mid-century. This path starts with today's proven LWR technology and uranium oxide fuels and takes an evolutionary route to a first commercial fast reactor that complements and supports the existing reliable LWR fleet. This paper is not intended to promote any one technology or approach, but instead seeks to illustrate the value of national level support for an energy technology program oriented toward demonstration and seeks eventual deployment of transformational energy systems on the scale and time-frame that matters. The observations and conclusions presented are derived primarily from the collective experience and expertise of the authors who represent a broad range of commercial perspectives from the utility, vendor, research and development, and regulatory policy communities. (authors)

  13. Hybrid adsorptive membrane reactor

    DOE Patents [OSTI]

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  14. H Reactor - Hanford Site

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

    About Us Projects & Facilities H Reactor About Us About Hanford Cleanup Hanford History ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  15. C Reactor - Hanford Site

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

    C Reactor About Us About Hanford Cleanup Hanford History Hanford Site Wide Programs ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  16. F Reactor - Hanford Site

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

    About Us Projects & Facilities F Reactor About Us About Hanford Cleanup Hanford History ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  17. N Reactor - Hanford Site

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

    Projects & Facilities N Reactor About Us About Hanford Cleanup Hanford History Hanford ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  18. Control Means for Reactor

    DOE Patents [OSTI]

    Manley, J. H.

    1961-06-27

    An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

  19. Power Burst Facility (PBF) Reactor Reactor Decommissioning

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

    Click here to view Click here to view Reactor Decommissioning Click on an image to enlarge A crane removes the reactor vessel from the Power Burst Facility (top), then places it ...

  20. Core Drilling Demonstration

    Office of Energy Efficiency and Renewable Energy (EERE)

    Tank Farms workers demonstrate core drilling capabilities for Hanford single-shell tanks. Core drilling is used to determine the current condition of each tank to assist in the overall assessment...

  1. International Research Reactor Decommissioning Project

    SciTech Connect (OSTI)

    Leopando, Leonardo; Warnecke, Ernst

    2008-01-15

    Many research reactors have been or will be shut down and are candidates for decommissioning. Most of the respective countries neither have a decommissioning policy nor the required expertise and funds to effectively implement a decommissioning project. The IAEA established the Research Reactor Decommissioning Demonstration Project (R{sup 2}D{sup 2}P) to help answer this need. It was agreed to involve the Philippine Research Reactor (PRR-1) as model reactor to demonstrate 'hands-on' experience as it is just starting the decommissioning process. Other facilities may be included in the project as they fit into the scope of R{sup 2}D{sup 2}P and complement to the PRR-1 decommissioning activities. The key outcome of the R{sup 2}D{sup 2}P will be the decommissioning of the PRR-1 reactor. On the way to this final goal the preparation of safety related documents (i.e., decommissioning plan, environmental impact assessment, safety analysis report, health and safety plan, cost estimate, etc.) and the licensing process as well as the actual dismantling activities could provide a model to other countries involved in the project. It is expected that the R{sup 2}D{sup 2}P would initiate activities related to planning and funding of decommissioning activities in the participating countries if that has not yet been done.

  2. Response Resources Demonstration

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

    Interoperability of Demand Response Resources Demonstration in NY Final Technical Report Award Number: DE-FC26-08NT02869 Project Type: Regional Demonstration Principal Investigator: Andre Wellington, Project Manager, Smart Grid Implementation Group Recipient: Consolidated Edison Company of New York, Inc. Team members: Innoventive Power and Verizon Communications Consolidated Edison Company of New York, Inc. Taxpayer ID Number: 13-5009340 Organizational DUNS: 00-698-2359 4 Irving Place New York,

  3. Demonstration & Market Transformation

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

    Demonstration & Market Transformation Peer Review Break-Out Presentation Jim Spaeth Program Manager Demonstration & Market Transformation March 23, 2015 2 | Bioenergy Technologies Office DMT Portfolio Peer Review * Introduction of the DMT Peer Review Team * Peer Review Process - Ground rules for review process * DMT Approach to Project Management - Budget Periods * Changes Made in Response to the 2013 Peer Review - Lessons Learned / Best Practices * Portfolio Overview - FOA Status and

  4. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Stewart, H.B.

    1958-12-23

    A nuclear reactor of the type speclfically designed for the irradiation of materials is discussed. In this design a central cyllndrical core of moderating material ls surrounded by an active portlon comprlsed of an annular tank contalning fissionable material immersed ln a liquid moderator. The active portion ls ln turn surrounded by a reflector, and a well ls provided in the center of the core to accommodate the materlals to be irradiated. The over-all dimensions of the core ln at least one plane are equal to or greater than twice the effective slowing down length and equal to or less than twlce the effective diffuslon length for neutrons in the core materials.

  5. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Daniels, F.

    1962-12-18

    A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

  6. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Wigner, E.P.

    1957-09-17

    A reactor of the type having coolant liquid circulated through clad fuel elements geometrically arranged in a solid moderator, such as graphite, is described. The core is enclosed in a pressure vessel and suitable shielding, wherein means is provided for circulating vapor through the core to superheat the same. This is accomplished by drawing off the liquid which has been heated in the core due to the fission of the fuel, passing it to a nozzle within a chamber where it flashes into a vapor, and then passing the vapor through separate tubes extending through the moderator to pick up more heat developed in the core due to the fission of the fuel, thereby producing superheated vapor.

  7. Energy Department to Invest in Advanced Reactor Concept Development

    Broader source: Energy.gov [DOE]

    Today the Energy Department released a funding opportunity announcement to support the research, development and demonstration of advanced nuclear reactor concepts.

  8. Nuclear reactor

    DOE Patents [OSTI]

    Yant, Howard W.; Stinebiser, Karl W.; Anzur, Gregory C.

    1977-01-01

    A nuclear reactor, particularly a liquid-metal breeder reactor, whose upper internals include outlet modules for channeling the liquid-metal coolant from selected areas of the outlet of the core vertically to the outlet plenum. The modules are composed of a highly-refractory, high corrosion-resistant alloy, for example, INCONEL-718. Each module is disposed to confine and channel generally vertically the coolant emitted from a subplurality of core-component assemblies. Each module has a grid with openings, each opening disposed to receive the coolant from an assembly of the subplurality. The grid in addition serves as a holdown for the assemblies of the corresponding subplurality preventing their excessive ejection upwardly from the core. In the region directly over the core the outlet modules are of such peripheral form that they nest forming a continuum over the core-component assemblies whose outlet coolant they confine. Each subassembly includes a chimney which confines the coolant emitted by its corresponding subassemblies to generally vertical flow between the outlet of the core and the outlet plenum. Each subplurality of assemblies whose emitted coolant is confined by an outlet module includes assemblies which emit lower-temperature coolant, for example, a control-rod assembly, or fertile assemblies, and assemblies which emit coolant of substantially higher temperature, for example, fuel-rod assemblies. The coolants of different temperatures are mixed in the chimneys reducing the effect of stripping (hot-cold temperature fluctuations) on the remainder of the upper internals which are composed typically of AISI-304 or AISI-316 stainless steel.

  9. Reactor and method of operation

    DOE Patents [OSTI]

    Wheeler, John A.

    1976-08-10

    A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

  10. DOE - Office of Legacy Management -- Elk River Reactor - MN 01

    Office of Legacy Management (LM)

    Elk River Reactor - MN 01 FUSRAP Considered Sites Site: Elk River Reactor (MN.01 ) Eliminated from consideration under FUSRAP - Reactor was dismantled and decommissioned by 1974 Designated Name: Not Designated Alternate Name: None Location: Elk River , Minnesota MN.01-1 Evaluation Year: 1985 MN.01-1 Site Operations: Boiling water reactor demonstration, research and development program MN.01-1 Site Disposition: Eliminated MN.01-1 Radioactive Materials Handled: None Indicated Primary Radioactive

  11. Innovative Demonstration Platform: PEC

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

    JOINT CENTER FOR ARTIFICIAL PHOTOSYNTHESIS SOLAR PEC H 2 DEVICES FRANCES HOULE AWSM workshop April 14-15, 2016 LAWRENCE BERKELEY NATIONAL LABORATORY THE STATE OF THE ART Fraunhofer/JCAP JCAP JCAP JCAP JCAP 2016 >15% JCAP Updated with 2015-16 demonstrations DEVICE TYPES FOR DEMONSTRATIONS 10% PLANAR DEVICE: THE LOUVERED DESIGN WITH FULL PRODUCT SEPARATION Potential (mV) Cell parameters: PV width = 1.43 cm Nafion height = 3.10 mm Channel height = 3.25 mm Catalysts: IrO 2 and Pt Solution: 1 M H

  12. Gigashot Optical Laser Demonstrator

    SciTech Connect (OSTI)

    Deri, R. J.

    2015-10-13

    The Gigashot Optical Laser Demonstrator (GOLD) project has demonstrated a novel optical amplifier for high energy pulsed lasers operating at high repetition rates. The amplifier stores enough pump energy to support >10 J of laser output, and employs conduction cooling for thermal management to avoid the need for expensive and bulky high-pressure helium subsystems. A prototype amplifier was fabricated, pumped with diode light at 885 nm, and characterized. Experimental results show that the amplifier provides sufficient small-signal gain and sufficiently low wavefront and birefringence impairments to prove useful in laser systems, at repetition rates up to 60 Hz.

  13. Demonstrating carbon capture

    SciTech Connect (OSTI)

    Qader, A.; Hooper, B.; Stevens, G.

    2009-11-15

    Australia is at the forefront of advancing CCS technology. The CO2CRC's H3 (Post-combustion) and Mulgrave (pre-combustion) capture projects are outlined. The capture technologies for these 2 demonstration projects are described. 1 map., 2 photos.

  14. Light Water Reactor Sustainability (LWRS) Program | Department...

    Energy Savers [EERE]

    Nuclear Reactor Technologies Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) ...

  15. SRS Small Modular Reactors

    ScienceCinema (OSTI)

    None

    2014-05-21

    The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

  16. Reactor safety method

    DOE Patents [OSTI]

    Vachon, Lawrence J.

    1980-03-11

    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.

  17. Nuclear reactor

    DOE Patents [OSTI]

    Thomson, Wallace B.

    2004-03-16

    A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

  18. Business Opportunities for Small Reactors

    SciTech Connect (OSTI)

    Minato, Akio; Nishimura, Satoshi; Brown, Neil W.

    2007-07-01

    This report assesses the market potential and identifies a number of potential paths for developing the small nuclear reactor business. There are several potential opportunities identified and evaluated. Selecting a specific approach for the business development requires additional information related to a specific market and sources of capital to support the investment. If and how a market for small nuclear plants may develop is difficult to predict because of the complexity of the economic and institutional factors that will influence such development. Key factors are; economics, safety, proliferation resistance and investment risk. The economic and political interest of any of the identified markets is also dependent on successful demonstration of the safety and reliability of small nuclear reactor. Obtaining a US-NRC Standard Design approval would be an important development step toward establishing a market for small reactors. (authors)

  19. AVNG system demonstration

    SciTech Connect (OSTI)

    Thron, Jonathan Louis; Mac Arthur, Duncan W; Kondratov, Sergey; Livke, Alexander; Razinkov, Sergey

    2010-01-01

    An attribute measurement system (AMS) measures a number of unclassified attributes of potentially classified material. By only displaying these unclassified results as red or green lights, the AMS protects potentially classified information while still generating confidence in the measurement result. The AVNG implementation that we describe is an AMS built by RFNC - VNIIEF in Sarov, Russia. To provide additional confidence, the AVNG was designed with two modes of operation. In the secure mode, potentially classified measurements can be made with only the simple red light/green light display. In the open mode, known unclassified material can be measured with complete display of the information collected from the radiation detectors. The AVNG demonstration, which occurred in Sarov, Russia in June 2009 for a joint US/Russian audience, included exercising both modes of AVNG operation using a number of multi-kg plutonium sources. In addition to describing the demonstration, we will show photographs and/or video taken of AVNG operation.

  20. Nucla CFB Demonstration Project

    SciTech Connect (OSTI)

    Not Available

    1990-12-01

    This report documents Colorado-Ute Electric Association's Nucla Circulating Atmospheric Fluidized-Bed Combustion (AFBC) demonstration project. It describes the plant equipment and system design for the first US utility-size circulating AFBC boiler and its support systems. Included are equipment and system descriptions, design/background information and appendices with an equipment list and selected information plus process flow and instrumentation drawings. The purpose of this report is to share the information gathered during the Nucla circulating AFBC demonstration project and present it so that the general public can evaluate the technical feasibility and cost effectiveness of replacing pulverized or stoker-fired boiler units with circulating fluidized-bed boiler units. (VC)

  1. National Hydrogen Learning Demonstration

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

    Keith Wipke, Sam Sprik, Jennifer Kurtz, Todd Ramsden, Chris Ainscough, Genevieve Saur February 6, 2012 DOE's Informational Webinar Series National Hydrogen Learning Demonstration Status This presentation does not contain any proprietary, confidential, or otherwise restricted information NREL is a national laboratory of the U.S. Department of Energy Office of Energy Efficiency and Renewable Energy operated by the Alliance for Sustainable Energy, LLC v8 National Renewable Energy Laboratory 2

  2. Thermionic Reactor Design Studies

    SciTech Connect (OSTI)

    Schock, Alfred

    1994-06-01

    During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic losses in the electrodes of full

  3. Demonstration of fully coupled simplified extended station black-out accident simulation with RELAP-7

    SciTech Connect (OSTI)

    Zhao, Haihua; Zhang, Hongbin; Zou, Ling; Anders, David; Martineau, Richard

    2014-10-01

    The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The RELAP-7 code develop-ment effort started in October of 2011 and by the end of the second development year, a number of physical components with simplified two phase flow capability have been de-veloped to support the simplified boiling water reactor (BWR) extended station blackout (SBO) analyses. The demonstration case includes the major components for the primary system of a BWR, as well as the safety system components for the safety relief valve (SRV), the reactor core isolation cooling (RCIC) system, and the wet well. Three scenar-ios for the SBO simulations have been considered. Since RELAP-7 is not a severe acci-dent analysis code, the simulation stops when fuel clad temperature reaches damage point. Scenario I represents an extreme station blackout accident without any external cooling and cooling water injection. The system pressure is controlled by automatically releasing steam through SRVs. Scenario II includes the RCIC system but without SRV. The RCIC system is fully coupled with the reactor primary system and all the major components are dynamically simulated. The third scenario includes both the RCIC system and the SRV to provide a more realistic simulation. This paper will describe the major models and dis-cuss the results for the three scenarios. The RELAP-7 simulations for the three simplified SBO scenarios show the importance of dynamically simulating the SRVs, the RCIC sys-tem, and the wet well system to the reactor safety during extended SBO accidents.

  4. Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment

    Broader source: Energy.gov [DOE]

    The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations, which govern the...

  5. Manufacturing Demonstration Facility: Low Temperature Materials Synthesis

    SciTech Connect (OSTI)

    Graham, David E.; Moon, Ji-Won; Armstrong, Beth L.; Datskos, Panos G.; Duty, Chad E.; Gresback, Ryan; Ivanov, Ilia N.; Jacobs, Christopher B.; Jellison, Gerald Earle; Jang, Gyoung Gug; Joshi, Pooran C.; Jung, Hyunsung; Meyer, III, Harry M.; Phelps, Tommy

    2015-06-30

    The Manufacturing Demonstration Facility (MDF) low temperature materials synthesis project was established to demonstrate a scalable and sustainable process to produce nanoparticles (NPs) for advanced manufacturing. Previous methods to chemically synthesize NPs typically required expensive, high-purity inorganic chemical reagents, organic solvents and high temperatures. These processes were typically applied at small laboratory scales at yields sufficient for NP characterization, but insufficient to support roll-to-roll processing efforts or device fabrication. The new NanoFermentation processes described here operated at a low temperature (~60 C) in low-cost, aqueous media using bacteria that produce extracellular NPs with controlled size and elemental stoichiometry. Up-scaling activities successfully demonstrated high NP yields and quality in a 900-L pilot-scale reactor, establishing this NanoFermentation process as a competitive biomanufacturing strategy to produce NPs for advanced manufacturing of power electronics, solid-state lighting and sensors.

  6. CCUS Demonstrations Making Progress

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

    9, First Quarter, 2013 www.fossil.energy.gov/news/energytoday.html HigHligHts inside 2 CCUS Demonstrations Making Progress A Column from the Director of Clean Energy Sys- tems, Office of Clean Coal 4 LNG Exports DOE Releases Third Party Study on Impact of Natural Gas Exports 5 Providing Emergency Relief Petroleum Reservers Helps Out with Hurricane Relief Efforts 7 Game-Changing Membranes FE-Funded Project Develops Novel Membranes for CCUS 8 Shale Gas Projects Selected 15 Projects Will Research

  7. NAVAJO ELECTRIFICATION DEMONSTRATION PROJECT

    SciTech Connect (OSTI)

    Terry W. Battiest

    2008-06-11

    The Navajo Electrification Demonstration Project (NEDP) is a multi-year project which addresses the electricity needs of the unserved and underserved Navajo Nation, the largest American Indian tribe in the United States. The program serves to cumulatively provide off-grid electricty for families living away from the electricty infrastructure, line extensions for unserved families living nearby (less than 1/2 mile away from) the electricity, and, under the current project called NEDP-4, the construction of a substation to increase the capacity and improve the quality of service into the central core region of the Navajo Nation.

  8. Residential Transactive Control Demonstration

    SciTech Connect (OSTI)

    Widergren, Steven E.; Fuller, Jason C.; Marinovici, Maria C.; Somani, Abhishek

    2014-02-19

    Arguably the most exciting aspect of the smart grid vision is the full participation of end-use resources with all forms of generation and energy storage in the reliable and efficient operation of an electric power system. Engaging all of these resources in a collaborative manner that respects the objectives of each resource, is sensitive to the system and local constraints of electricity flow, and scales to the large number of devices and systems participating is a grand challenge. Distributed decision-making system approaches have been presented and experimentation is underway. This paper reports on the preliminary findings of a residential demand response demonstration that uses the bidding transactions of supply and end-use air conditioning resources communicating with a real-time, 5 minute market to balance the various needs of the participants on a distribution feeder. The nature of the demonstration, the value streams being explored, and the operational scenarios implemented to characterize the system response are summarized along with preliminary findings.

  9. Fuel Cell Demonstration Program

    SciTech Connect (OSTI)

    Gerald Brun

    2006-09-15

    In an effort to promote clean energy projects and aid in the commercialization of new fuel cell technologies the Long Island Power Authority (LIPA) initiated a Fuel Cell Demonstration Program in 1999 with six month deployments of Proton Exchange Membrane (PEM) non-commercial Beta model systems at partnering sites throughout Long Island. These projects facilitated significant developments in the technology, providing operating experience that allowed the manufacturer to produce fuel cells that were half the size of the Beta units and suitable for outdoor installations. In 2001, LIPA embarked on a large-scale effort to identify and develop measures that could improve the reliability and performance of future fuel cell technologies for electric utility applications and the concept to establish a fuel cell farm (Farm) of 75 units was developed. By the end of October of 2001, 75 Lorax 2.0 fuel cells had been installed at the West Babylon substation on Long Island, making it the first fuel cell demonstration of its kind and size anywhere in the world at the time. Designed to help LIPA study the feasibility of using fuel cells to operate in parallel with LIPA's electric grid system, the Farm operated 120 fuel cells over its lifetime of over 3 years including 3 generations of Plug Power fuel cells (Lorax 2.0, Lorax 3.0, Lorax 4.5). Of these 120 fuel cells, 20 Lorax 3.0 units operated under this Award from June 2002 to September 2004. In parallel with the operation of the Farm, LIPA recruited government and commercial/industrial customers to demonstrate fuel cells as on-site distributed generation. From December 2002 to February 2005, 17 fuel cells were tested and monitored at various customer sites throughout Long Island. The 37 fuel cells operated under this Award produced a total of 712,635 kWh. As fuel cell technology became more mature, performance improvements included a 1% increase in system efficiency. Including equipment, design, fuel, maintenance, installation

  10. Fusion Power Demonstration III

    SciTech Connect (OSTI)

    Lee, J.D.

    1985-07-01

    This is the third in the series of reports covering the Fusion Power Demonstration (FPD) design study. This volume considers the FPD-III configuration that incorporates an octopole end plug. As compared with the quadrupole end-plugged designs of FPD-I and FPD-II, this octopole configuration reduces the number of end cell magnets and shortens the minimum ignition length of the central cell. The end-cell plasma length is also reduced, which in turn reduces the size and cost of the end cell magnets and shielding. As a contiuation in the series of documents covering the FPD, this report does not stand alone as a design description of FPD-III. Design details of FPD-III subsystems that do not differ significantly from those of the FPD-II configuration are not duplicated in this report.

  11. Jennings Demonstration PLant

    SciTech Connect (OSTI)

    Russ Heissner

    2010-08-31

    Verenium operated a demonstration plant with a capacity to produce 1.4 million gallons of cellulosic ethanol from agricultural resiues for about two years. During this time, the plant was able to evaluate the technical issues in producing ethanol from three different cellulosic feedstocks, sugar cane bagasse, energy cane, and sorghum. The project was intended to develop a better understanding of the operating parameters that would inform a commercial sized operation. Issues related to feedstock variability, use of hydrolytic enzymes, and the viability of fermentative organisms were evaluated. Considerable success was achieved with pretreatment processes and use of enzymes but challenges were encountered with feedstock variability and fermentation systems. Limited amounts of cellulosic ethanol were produced.

  12. Tidd PFBC demonstration project

    SciTech Connect (OSTI)

    Marrocco, M.

    1997-12-31

    The Tidd project was one of the first joint government-industry ventures to be approved by the US Department of Energy (DOE) in its Clean Coal Technology Program. In March 1987, DOE signed an agreement with the Ohio Power Company, a subsidiary of American Electric Power, to refurbish the then-idle Tidd plant on the banks of the Ohio River with advanced pressurized fluidized bed technology. Testing ended after 49 months of operation, 100 individual tests, and the generation of more than 500,000 megawatt-hours of electricity. The demonstration plant has met its objectives. The project showed that more than 95 percent of sulfur dioxide pollutants could be removed inside the advanced boiler using the advanced combustion technology, giving future power plants an attractive alternative to expensive, add-on scrubber technology. In addition to its sulfur removal effectiveness, the plant`s sustained periods of steady-state operation boosted its availability significantly above design projections, heightening confidence that pressurized fluidized bed technology will be a reliable, baseload technology for future power plants. The technology also controlled the release of nitrogen oxides to levels well below the allowable limits set by federal air quality standards. It also produced a dry waste product that is much easier to handle than wastes from conventional power plants and will likely have commercial value when produced by future power plants.

  13. Nuclear Reactor Technology Subcommittee of NEAC

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

    NEAC: Advanced Test/Demo Reactor Options Study Mike Corradini (UW), Chair Ashok Bhatnagar (Consultant), Doug Chapin (MPR), Tom Cochran (NRDC emeritus), Regis Matzie (Consultant) , Harold Ray (Consultant), Joy Rempe (Consultant), John Sackett (Consultant) Nuclear Energy Advisory Committee Meeting June 17, 2016 1 NRT Subcommittee Scope * Congress appropriated funds for "an advanced test/demonstration reactor planning study by the national laboratories, industry, and relevant stakeholders of

  14. Attrition reactor system

    DOE Patents [OSTI]

    Scott, C.D.; Davison, B.H.

    1993-09-28

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

  15. Hybrid plasmachemical reactor

    SciTech Connect (OSTI)

    Lelevkin, V. M. Smirnova, Yu. G.; Tokarev, A. V.

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  16. Attrition reactor system

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN); Davison, Brian H. (Knoxvile, TN)

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  17. Prospects for Tokamak Fusion Reactors

    SciTech Connect (OSTI)

    Sheffield, J.; Galambos, J.

    1995-04-01

    This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant.

  18. NEUTRONIC REACTOR POWER PLANT

    DOE Patents [OSTI]

    Metcalf, H.E.

    1962-12-25

    This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

  19. Reactor System Transient Code.

    Energy Science and Technology Software Center (OSTI)

    1999-07-14

    RELAP3B describes the behavior of water-cooled nuclear reactors during postulated accidents or power transients, such as large reactivity excursions, coolant losses or pump failures. The program calculates flows, mass and energy inventories, pressures, temperatures, and steam qualities along with variables associated with reactor power, reactor heat transfer, or control systems. Its versatility allows one to describe simple hydraulic systems as well as complex reactor systems.

  20. Period meter for reactors

    DOE Patents [OSTI]

    Rusch, Gordon K.

    1976-01-06

    An improved log N amplifier type nuclear reactor period meter with reduced probability for noise-induced scrams is provided. With the reactor at low power levels a sampling circuit is provided to determine the reactor period by measuring the finite change in the amplitude of the log N amplifier output signal for a predetermined time period, while at high power levels, differentiation of the log N amplifier output signal provides an additional measure of the reactor period.

  1. NUCLEAR REACTOR CONTROL SYSTEM

    DOE Patents [OSTI]

    Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

    1959-11-01

    A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

  2. NEUTRONIC REACTOR SHIELDING

    DOE Patents [OSTI]

    Borst, L.B.

    1961-07-11

    A special hydrogenous concrete shielding for reactors is described. In addition to Portland cement and water, the concrete essentially comprises 30 to 60% by weight barytes aggregate for enhanced attenuation of fast neutrons. The biological shields of AEC's Oak Ridge Graphite Reactor and Materials Testing Reactor are particular embodiments.

  3. Advanced Test Reactor Tour

    ScienceCinema (OSTI)

    Miley, Don

    2013-05-28

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

  4. Advanced Test Reactor Tour

    SciTech Connect (OSTI)

    Miley, Don

    2011-01-01

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

  5. High solids fermentation reactor

    DOE Patents [OSTI]

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

    1993-01-01

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

  6. High solids fermentation reactor

    DOE Patents [OSTI]

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

    1993-03-02

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

  7. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, James P.; Scahill, John W.

    1995-01-01

    An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

  8. Reactor vessel support system

    DOE Patents [OSTI]

    Golden, Martin P.; Holley, John C.

    1982-01-01

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  9. B Reactor - Hanford Site

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

    Reactor About Us About Hanford Cleanup Hanford History Hanford Site Wide Programs Contact Us 100 Area 118-K-1 Burial Ground 200 Area 222-S Laboratory 242-A Evaporator 300 Area 324 Building 325 Building 400 Area/Fast Flux Test Facility 618-10 and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and Interim Storage Area Canyon Facilities Cold Test Facility D and DR Reactors Effluent Treatment Facility Environmental Restoration Disposal Facility F Reactor H

  10. Nuclear reactor overflow line

    DOE Patents [OSTI]

    Severson, Wayne J.

    1976-01-01

    The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

  11. Fission reactors and materials

    SciTech Connect (OSTI)

    Frost, B.R.T.

    1981-12-01

    The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions.

  12. Spinning fluids reactor

    DOE Patents [OSTI]

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  13. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

  14. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, D.M.; Taft, W.E.

    1994-12-20

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  15. Use of acceptable knowledge to demonstrate TRAMPAC compliance

    SciTech Connect (OSTI)

    Whitworth, J. (Julia); Becker, B. (Blair); Guerin, D. (David); Shokes, T. (Tamara)

    2004-01-01

    Recently, Los Alamos National Laboratory-Carlsbad Operations (LANL-CO) has supported the Central Characterization Project (CCP) managed by the U.S. Department of Energy (DOE) in the shipment of transuranic (TRU) waste from various small-quantity TRU waste generators to hub sites or other DOE sites in TRUPACT-II shipping containers. This support has involved using acceptable knowledge (AK) to demonstrate compliance with various requirements of Revision 19 of the TRUPACT-II Authorized Methods of Payload Compliance (TRAMPAC). LANL-CO has worked to facilitate TRUPACT-II shipments from the University of Missouri Research Reactor (MURR) and Lovelace Respiratory Research Institute (LRRI) to Argonne National Laboratory-East (ANL-E) and Los Alamos National Laboratory (LANL), respectively. The latter two sites have TRU waste certification programs approved to ship waste to the Waste Isolation Pilot Plant (WIPP) for disposal. In each case, AK was used to satisfy the necessary information to ship the waste to other DOE facilities. For the purposes of intersite shipment, AK provided data to WIPP Waste Information System (WWIS) transportation modules to ensure that required information was obtained prior to TRUPACT-II shipments. The WWIS modules were used for the intersite shipments, not to enter certification data into WWIS, but rather to take advantage of a validated system to ensure that the containers to be shipped were compliant with TRAMPAC requirements, particularly in the evaluation of quantitative criteria. LANL-CO also assisted with a TRAMPAC compliance demonstration for homogeneous waste containers shipped in TRUPACT-II containers from ANL-E to Idaho National Engineering and Environmental Laboratory (INEEL) for the purpose of core sampling. The basis for the TRAMPAC compliance determinations was AK regarding radiological composition, chemical composition, TRU waste container packaging, and absence of prohibited items. Also, even in the case where AK is not used to

  16. RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (1) Reactor concept and plant dynamics analyses

    SciTech Connect (OSTI)

    Kambe, Mitsuru [Central Research Institute of Electric Power Industry (CRIEPI), 2-11-1, Iwado Kita, Komae-shi, Tokyo, 201-8511 (Japan); Tsunoda, Hirokazu [Mitsubishi Research Institute, Inc. 3-6, Otemachi 2-chome, Chiyoda-ku, Tokyo, 100-8141 (Japan); Mishima, Kaichiro [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka, 590-20494 (Japan); Iwamura, Takamichi [Japan Atomic Energy Research Institute, 2-4, Shirakata-shirane, Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195 (Japan)

    2002-07-01

    The 200 kWe uranium-nitride fueled lithium cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for Lunar base power system. It is one of the variants of RAPID (Refueling by All Pins Integrated Design), fast reactor concept, which enable quick and simplified refueling. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small size reactor core, 2700 fuel pins are integrated altogether and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years. Unique challenges in reactivity control systems design have been attempted in RAPID-L concept. The reactor has no control rod, but involves the following innovative reactivity control systems: Lithium Expansion Modules (LEM) for inherent reactivity feedback, Lithium Injection Modules (LIM) for inherent ultimate shutdown, and Lithium Release Modules (LRM) for automated reactor startup. All these systems adopt lithium-6 as a liquid poison instead of B{sub 4}C rods. In combination with LEMs, LIMs and LRMs, RAPID-L can be operated without operator. This is the first reactor concept ever established in the world. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, RAPID-L reactor concept and its transient characteristics are presented. (authors)

  17. Vanadium recycling for fusion reactors

    SciTech Connect (OSTI)

    Dolan, T.J.; Butterworth, G.J.

    1994-04-01

    Very stringent purity specifications must be applied to low activation vanadium alloys, in order to meet recycling goals requiring low residual dose rates after 50--100 years. Methods of vanadium production and purification which might meet these limits are described. Following a suitable cooling period after their use, the vanadium alloy components can be melted in a controlled atmosphere to remove volatile radioisotopes. The aim of the melting and decontamination process will be the achievement of dose rates low enough for ``hands-on`` refabrication of new reactor components from the reclaimed metal. The processes required to permit hands-on recycling appear to be technically feasible, and demonstration experiments are recommended. Background information relevant to the use of vanadium alloys in fusion reactors, including health hazards, resources, and economics, is provided.

  18. CALDERON COKEMAKING PROCESS/DEMONSTRATION PROJECT

    SciTech Connect (OSTI)

    Albert Calderon

    2000-09-19

    This project deals with the demonstration of a coking process using proprietary technology of Calderon, with the following objectives geared to facilitate commercialization: (i) making coke of such quality as to be suitable for use in hard-driving, large blast furnaces; (ii) providing proof that such process is continuous and environmentally closed to prevent emissions; (iii) demonstrating that high-coking-pressure (non-traditional) coal blends which cannot be safely charged into conventional by-product coke ovens can be used in the Calderon process; (iv) conducting a blast furnace test to demonstrate the compatibility of the coke produced; (v) demonstrating that coke can be produced economically, at a level competitive with coke imports; and (vi) applying the Calderon technology to making additional iron units. The activities of the past quarter were focused on the following: (1) Bethlehem Steel's withdrawal and efforts expended to substitute U.S. Steel for Bethlehem; (2) Assessment work performed with U.S. Steel to show that the Calderon Technology has merit and would add to U.S. Steel's economic benefit by being involved in it, including for making additional iron units; (3) Addressing material selection and heat input capacity to increase heat input into the processing reactor by actual modeling of such approach; (4) Construction of two full size courses of heating tiles to verify the manufacturing and the fitting of the tiles with one another; (5) Making available equipment to test carbon deposition on sorbent; and (6) Permitting issues.

  19. Commercial Building Demonstration and Deployment

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

    Commercial Building Demonstration and Deployment 2014 Building Technologies Office Peer ... April 23 rd 11:15-11:30 Commercial DemonstrationDeployment Overview Kristen Taddonio, ...

  20. CHARACTERIZATION OF RADIOACTIVITY IN THE REACTOR VESSEL OF THE HEAVY WATER COMPONENT TEST REACTOR

    SciTech Connect (OSTI)

    Vinson, Dennis

    2010-06-01

    The Heavy Water Component Test Reactor (HWCTR) facility is a pressurized heavy water reactor that was used to test candidate fuel designs for heavy water power reactors. The reactor operated at nominal power of 50 MW{sub th}. The reactor coolant loop operated at 1200 psig and 250 C. Two isolated test loop were designed into the reactor to provide special test conditions. Fig. 1 shows a cut-away view of the reactor. The two loops are contained in four inch diameter stainless steel piping. The HWCTR was operated for only a short duration, from March 1962 to December 1964 in order to test the viability of test fuel elements and other reactor components for use in a heavy water power reactor. The reactor achieved 13,882 MWd of total power while testing 36 different fuel assemblies. In the course of operation, HWCTR experienced the cladding failures of 10 separate test fuel assemblies. In each case, the cladding was breached with some release of fuel core material into the isolated test loop, causing fission product and actinide contamination in the main coolant loop and the liquid and boiling test loops. Despite the contribution of the contamination from the failed fuel, the primary source of radioactivity in the HWCTR vessel and internals is the activation products in the thermal shields, and to a lesser degree, activation products in the reactor vessel walls and liner. A detailed facility characterization report of the HWCTR facility was completed in 1996. Many of the inputs and assumptions in the 1996 characterization report were derived from the HWCTR decommissioning plan published in 1975. The current paper provides an updated assessment of the radioisotopic characteristics of the HWCTR vessel and internals to support decommissioning activities on the facility.

  1. POTENTIAL BENCHMARKS FOR ACTINIDE PRODUCTION IN HANFORD REACTORS

    SciTech Connect (OSTI)

    PUIGH RJ; TOFFER H

    2011-10-19

    A significant experimental program was conducted in the early Hanford reactors to understand the reactor production of actinides. These experiments were conducted with sufficient rigor, in some cases, to provide useful information that can be utilized today in development of benchmark experiments that may be used for the validation of present computer codes for the production of these actinides in low enriched uranium fuel.

  2. Enhanced Geothermal Systems Demonstration Projects

    SciTech Connect (OSTI)

    Geothermal Technologies Office

    2013-08-06

    Several Enhanced Geothermal Systems (EGS) demonstration projects are highlighted on this Geothermal Technologies Office Web page.

  3. THERMAL NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Spinrad, B.I.

    1960-01-12

    A novel thermal reactor was designed in which a first reflector formed from a high atomic weight, nonmoderating material is disposed immediately adjacent to the reactor core. A second reflector composed of a moderating material is disposed outwardly of the first reflector. The advantage of this novel reflector arrangement is that the first reflector provides a high slow neutron flux in the second reflector, where irradiation experiments may be conducted with a small effect on reactor reactivity.

  4. NEUTRONIC REACTOR SYSTEM

    DOE Patents [OSTI]

    Goett, J.J.

    1961-01-24

    A system is described which includes a neutronic reactor containing a dispersion of fissionable material in a liquid moderator as fuel and a conveyor to which a portion of the dispersion may be passed and wherein the self heat of the slurry evaporates the moderator. Means are provided for condensing the liquid moderator and returning it to the reactor and for conveying the dried fissionable material away from the reactor.

  5. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, J.P.; Scahill, J.W.

    1995-05-09

    An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

  6. Small Modular Reactors (SMRs) | Department of Energy

    Energy Savers [EERE]

    Reactor Technologies Small Modular Reactors (SMRs) Small Modular Reactors (SMRs) ... to the NRC by late-2016 Complete reactor module final design by mid-2019 For more ...

  7. Technology, Safety and Costs of Decommissioning Nuclear Reactors At Multiple-Reactor Stations

    SciTech Connect (OSTI)

    Wittenbrock, N. G.

    1982-01-01

    Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWRs) and large (1155-MWe) boiling water reactors {BWRs) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite nuclear waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services. Five scenarios for decommissioning reactors at a multiple-reactor station are investigated. The number of reactors on a site is assumed to be either four or ten; nuclear waste disposal is varied between immediate offsite disposal, interim onsite storage, and immediate onsite disposal. It is assumed that the decommissioned reactors are not replaced in one scenario but are replaced in the other scenarios. Centralized service facilities are provided in two scenarios but are not provided in the other three. Decommissioning of a PWR or a BWR at a multiple-reactor station probably will be less costly and result in lower radiation doses than decommissioning an identical reactor at a single-reactor station. Regardless of whether the light water reactor being decommissioned is at a single- or multiple-reactor station: • the estimated occupational radiation dose for decommissioning an LWR is lowest for SAFSTOR and highest for DECON • the estimated

  8. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, Juhani

    1996-01-01

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

  9. Tokamak reactor first wall

    DOE Patents [OSTI]

    Creedon, R.L.; Levine, H.E.; Wong, C.; Battaglia, J.

    1984-11-20

    This invention relates to an improved first wall construction for a tokamak fusion reactor vessel, or other vessels subjected to similar pressure and thermal stresses.

  10. Small Modular Reactors - SRSCRO

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

    smr Small Modular Reactors The Savannah River National Laboratory (SRNL) has announced several partnerships to bring refrigerator-sized modular nuclear reactors, known as Small Modular Reactors or SMRs, to the Savannah River Site facility and jump start development of the U.S. Energy Freedom CenterTM. Currently, all large commercial power reactors in the United States and most in the rest of the world are based on "light water" designs - that is, they use uranium fuel and ordinary

  11. HOMOGENEOUS NUCLEAR POWER REACTOR

    DOE Patents [OSTI]

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  12. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, J.

    1996-03-19

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

  13. Level 1 transient model for a molybdenum-99 producing aqueous homogeneous reactor and its applicability to the tracy reactor

    SciTech Connect (OSTI)

    Nygaard, E. T.; Williams, M. M. R.; Angelo, P. L.

    2012-07-01

    Babcock and Wilcox Technical Services Group (B and W) has identified aqueous homogeneous reactors (AHRs) as a technology well suited to produce the medical isotope molybdenum 99 (Mo-99). AHRs have never been specifically designed or built for this specialized purpose. However, AHRs have a proven history of being safe research reactors. In fact, in 1958, AHRs had 'a longer history of operation than any other type of research reactor using enriched fuel' and had 'experimentally demonstrated to be among the safest of all various type of research reactor now in use [1].' A 'Level 1' model representing B and W's proposed Medical Isotope Production System (MIPS) reactor has been developed. The Level 1 model couples a series of differential equations representing neutronics, temperature, and voiding. Neutronics are represented by point reactor kinetics while temperature and voiding terms are axially varying (one-dimensional). While this model was developed specifically for the MIPS reactor, its applicability to the Japanese TRACY reactor was assessed. The results from the Level 1 model were in good agreement with TRACY experimental data and found to be conservative over most of the time domains considered. The Level 1 model was used to study the MIPS reactor. An analysis showed the Level 1 model agreed well with a more complex computational model of the MIPS reactor (a FETCH model). Finally, a significant reactivity insertion was simulated with the Level 1 model to study the MIPS reactor's time-dependent response. (authors)

  14. Criteria for the evaluation of a dilute decontamination demonstration

    SciTech Connect (OSTI)

    FitzPatrick, V.F.; Divine, J.R.; Hoenes, G.R.; Munson, L.F.; Card, C.J.

    1981-12-01

    This document provides the prerequisite technical information required to evaluate and/or develop a project to demonstrate the dilute chemical decontamination of the primary coolant system of light water reactors. The document focuses on five key areas: the basis for establishing programmatic prerequisites and the key decision points that are required for proposal evaluation and/or RFP (Request for Proposal) issuance; a technical review of the state-of-the-art to identify the potential impacts of a reactor's primary-system decontamination on typical BWR and PWR plants; a discussion of the licensing, recertification, fuel warranty, and institutional considerations and processes; a preliminary identification and development of the selection criteria for the reactor and the decontamination process; and a preliminary identification of further research and development that might be required.

  15. EVALUATION OF ACTIVATION PRODUCTS IN REMAINING IN REMAINING K-, L- AND C-REACTOR STRUCTURES

    SciTech Connect (OSTI)

    Vinson, D.; Webb, R.

    2010-09-30

    An analytic model and calculational methodology was previously developed for P-reactor and R-reactor to quantify the radioisotopes present in Savannah River Site (SRS) reactor tanks and the surrounding structural materials as a result of neutron activation of the materials during reactor operation. That methodology has been extended to K-reactor, L-reactor, and C-reactor. The analysis was performed to provide a best-estimate source term input to the Performance Assessment for an in-situ disposition strategy by Site Decommissioning and Demolition (SDD). The reactor structure model developed earlier for the P-reactor and R-reactor analyses was also used for the K-reactor and L-reactor. The model was suitably modified to handle the larger Creactor tank and associated structures. For all reactors, the structure model consisted of 3 annular zones, homogenized by the amount of structural materials in the zone, and 5 horizontal layers. The curie content on an individual radioisotope basis and total basis for each of the regions was determined. A summary of these results are provided herein. The efficacy of this methodology to accurately predict the radioisotopic content of the reactor systems in question has been demonstrated and is documented in Reference 1. As noted in that report, results for one reactor facility cannot be directly extrapolated to other SRS reactors.

  16. CALDERON COKEMAKING PROCESS/DEMONSTRATION PROJECT

    SciTech Connect (OSTI)

    ALBERT CALDERON

    1998-09-22

    This project deals with the demonstration of a coking process using proprietary technology of Calderon, with the following objectives geared to facilitate commercialization: (i) making coke of such quality as to be suitable for use in hard-driving, large blast furnaces; (ii) providing proof that such process is continuous and environmentally closed to prevent emissions; (iii) demonstrating that high-coking-pressure (non-traditional) coal blends which cannot be safely charged into conventional by-product coke ovens can be used in the Calderon process; and (iv) demonstrating that coke can be produced economically, at a level competitive with coke imports. The activities of the past quarter were focused on the following: ? Consolidation of the project team-players; ? Recruiting Koppers Industries as an additional stakeholder; ? Developing a closed system for the production of binder pitch from tar in the Calderon coking process as the incentive for Koppers to join the team; ? Gathering appropriate equipment for conducting a set of experiments at bench scale to simulate tar quality produced from the Calderon coking process for the production of binder pitch; and ? Further progress made in the design of the commercial coking reactor.

  17. Small Modular Reactor: First of a Kind (FOAK) and Nth of a Kind (NOAK) Economic Analysis

    SciTech Connect (OSTI)

    Lauren M. Boldon; Piyush Sabharwall

    2014-08-01

    Small modular reactors (SMRs) refer to any reactor design in which the electricity generated is less than 300 MWe. Often medium sized reactors with power less than 700 MWe are also grouped into this category. Internationally, the development of a variety of designs for SMRs is booming with many designs approaching maturity and even in or nearing the licensing stage. It is for this reason that a generalized yet comprehensive economic model for first of a kind (FOAK) through nth of a kind (NOAK) SMRs based upon rated power, plant configuration, and the fiscal environment was developed. In the model, a particular project’s feasibility is assessed with regards to market conditions and by commonly utilized capital budgeting techniques, such as the net present value (NPV), internal rate of return (IRR), Payback, and more importantly, the levelized cost of energy (LCOE) for comparison to other energy production technologies. Finally, a sensitivity analysis was performed to determine the effects of changing debt, equity, interest rate, and conditions on the LCOE. The economic model is primarily applied to the near future water cooled SMR designs in the United States. Other gas cooled and liquid metal cooled SMR designs have been briefly outlined in terms of how the economic model would change. FOAK and NOAK SMR costs were determined for a site containing seven 180 MWe water cooled SMRs and compared to a site containing one 1260 MWe reactor. With an equal share of debt and equity and a 10% cost of debt and equity, the LCOE was determined to be $79 $84/MWh and $80/MWh for the SMR and large reactor sites, respectively. With a cost of equity of 15%, the SMR LCOE increased substantially to $103 $109/MWh. Finally, an increase in the equity share to 70% at the 15% cost of equity resulted in an even higher LCOE, demonstrating the large variation in results due to financial and market factors. The NPV and IRR both decreased with increasing LCOE. Unless the price of electricity

  18. Brookhaven Graphite Research Reactor Workshop

    Broader source: Energy.gov [DOE]

    The Brookhaven Graphite Research Reactor (BGRR) was the first reactor built in the U.S. for peacetime atomic research following World War II.  Construction began in 1947 and the reactor started...

  19. FIRST LEGO League Robotics Demonstration

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

    FIRST LEGO League Robotics Demonstration FIRST LEGO League Robotics Demonstration WHEN: Mar 14, 2015 11:00 AM - 1:00 PM WHERE: Bradbury Science Museum 1350 Central Ave, Los Alamos, ...

  20. QuickPEP Tool Demonstration

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

    QuickPEP Tool Demonstration Riyaz Papar, PE, CEM Director, Energy Assets & Optimization ... * Plant Energy Profiling * QuickPEP Demonstration * New features in Quick 2.0 * Wrap Up ...

  1. NEUTRONIC REACTOR SYSTEM

    DOE Patents [OSTI]

    Treshow, M.

    1959-02-10

    A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

  2. NEUTRONIC REACTOR BURIAL ASSEMBLY

    DOE Patents [OSTI]

    Treshow, M.

    1961-05-01

    A burial assembly is shown whereby an entire reactor core may be encased with lead shielding, withdrawn from the reactor site and buried. This is made possible by a five-piece interlocking arrangement that may be easily put together by remote control with no aligning of bolt holes or other such close adjustments being necessary.

  3. Light water reactor program

    SciTech Connect (OSTI)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  4. REFLECTOR FOR NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Fraas, A.P.

    1963-08-01

    A reflector for nuclear reactors that comprises an assembly of closely packed graphite rods disposed with their major axes substantially perpendicular to the interface between the reactor core and the reflector is described. Each graphite rod is round in transverse cross section at (at least) its interface end and is provided, at that end, with a coaxial, inwardly tapering hole. (AEC)

  5. GATEWAY Demonstrations | Department of Energy

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

    GATEWAY Demonstrations GATEWAY Demonstrations DOE GATEWAY demonstrations enable detailed LED product evaluation and hands-on experience that cannot be replicated in a lab. High-performance LED products are selected and installed in real-world applications, providing valuable data and experience on product performance and cost effectiveness. The results often reveal important issues related to installation, interface, and control. Results DOE shares the results of completed GATEWAY demonstration

  6. Hydrogen Storage Materials Database Demonstration

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

    Hydrogen Storage Materials Database Demonstration FUEL CELL TECHNOLOGIES ... 12132011 Hydrogen Storage Materials Database Marni Lenahan December 13, 2011 Database ...

  7. Application of the LBB regulatory approach to the steamlines of advanced WWER 1000 reactor

    SciTech Connect (OSTI)

    Kiselyov, V.A.; Sokov, L.M.

    1997-04-01

    The LBB regulatory approach adopted in Russia in 1993 as an extra safety barrier is described for advanced WWER 1000 reactor steamline. The application of LBB concept requires the following additional protections. First, the steamline should be a highly qualified piping, performed in accordance with the applicable regulations and guidelines, carefully screened to verify that it is not subjected to any disqualifying failure mechanism. Second, a deterministic fracture mechanics analysis and leak rate evaluation have been performed to demonstrate that postulated through-wall crack that yields 95 1/min at normal operation conditions is stable even under seismic loads. Finally, it has been verified that the leak detection systems are sufficiently reliable, diverse and sensitive, and that adequate margins exist to detect a through wall crack smaller than the critical size. The obtained results are encouraging and show the possibility of the application of the LBB case to the steamline of advanced WWER 1000 reactor.

  8. Advanced Reactor Research and Development Funding Opportunity Announcement

    Office of Energy Efficiency and Renewable Energy (EERE)

    The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) sponsors a program of research, development, and demonstration related to advanced non-light water reactor concepts. A goal of the...

  9. Adaptive Nodal Transport Methods for Reactor Transient Analysis

    SciTech Connect (OSTI)

    Thomas Downar; E. Lewis

    2005-08-31

    Develop methods for adaptively treating the angular, spatial, and time dependence of the neutron flux in reactor transient analysis. These methods were demonstrated in the DOE transport nodal code VARIANT and the US NRC spatial kinetics code, PARCS.

  10. Tandem mirror technology demonstration facility

    SciTech Connect (OSTI)

    Not Available

    1983-10-01

    This report describes a facility for generating engineering data on the nuclear technologies needed to build an engineering test reactor (ETR). The facility, based on a tandem mirror operating in the Kelley mode, could be used to produce a high neutron flux (1.4 MW/M/sup 2/) on an 8-m/sup 2/ test area for testing fusion blankets. Runs of more than 100 h, with an average availability of 30%, would produce a fluence of 5 mW/yr/m/sup 2/ and give the necessary experience for successful operation of an ETR.