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1

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

SciTech Connect (OSTI)

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

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

2012-08-01T23:59:59.000Z

2

Minimizing the Cost of Innovative Nuclear Technology Through Flexibility: The Case of a Demonstration Accelerator-Driven Subcritical Reactor Park  

E-Print Network [OSTI]

Presented is a methodology to analyze the expected Levelised Cost Of Electricity (LCOE) in the face of technology uncertainty for Accelerator-Driven Subcritical Reactors (ADSRs). It shows that flexibility in the design and deployment strategy...

Cardin, Michel-Alexandre; Steer, Steven J.; Nuttall, William J.; Parks, Geoffrey T.; Gonçalves, Leonardo V.N.; de Neufville, Richard

3

Simulator platform for fast reactor operation and safety technology demonstration  

SciTech Connect (OSTI)

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

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

2012-07-30T23:59:59.000Z

4

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

SciTech Connect (OSTI)

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

Curtis Smith; Steven Prescott; Tony Koonce

2014-04-01T23:59:59.000Z

5

Fundamental demonstration of natural circulation feasibility for an HLMC reactor  

SciTech Connect (OSTI)

Concepts are being developed and evaluated at Argonne National Laboratory for a smaller nuclear steam supply system with proliferation-resistant features targeted for export to developing countries. Specific features of interest here include low reactor power [300 MW(thermal)]; utilization of inert heavy-liquid-metal coolant (HLMC), namely, lead-bismuth eutectic (T{sub mp} = 125 C), eliminating concerns over metal-water reactions; 15-yr core lifetime, enabling access to fissile materials to be restricted by design; and reliance on purely natural-circulation coolant heat transport, eliminating primary system coolant pumps. Evaluation of this concept is being carried out in stages. The stage 1 investigations to which the results presented in this paper belong are directed at establishing the basic feasibility of the concept through the application of first-principles analyses. This approach is warranted while detailed aspects of the core design are yet to be determined. The objective of the present work is to demonstrate at a fundamental level the feasibility of utilizing natural-circulation coolant heat transport with the HLMC.

Sienicki, J.J.; Spencer, B.W.; Farmer, M.T.

1999-07-01T23:59:59.000Z

6

Antineutrino reactor safeguards - a case study  

E-Print Network [OSTI]

Antineutrinos have been proposed as a means of reactor safeguards for more than 30 years and there has been impressive experimental progress in neutrino detection. In this paper we conduct, for the first time, a case study of the application of antineutrino safeguards to a real-world scenario - the North Korean nuclear crisis in 1994. We derive detection limits to a partial or full core discharge in 1989 based on actual IAEA safeguards access and find that two independent methods would have yielded positive evidence for a second core with very high confidence. To generalize our results, we provide detailed estimates for the sensitivity to the plutonium content of various types of reactors, including most types of plutonium production reactors, based on detailed reactor simulations. A key finding of this study is that a wide class of reactors with a thermal power of less than 0.1-1 GWth can be safeguarded achieving IAEA goals for quantitative sensitivity and timeliness with detectors right outside the reactor building. This type of safeguards does not rely on the continuity of knowledge and provides information about core inventory and power status in real-time.

Eric Christensen; Patrick Huber; Patrick Jaffke

2014-02-13T23:59:59.000Z

7

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

SciTech Connect (OSTI)

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.

McAdams, R., E-mail: roy.mcadams@ccfe.ac.uk [EURATOM/CCFE Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom)

2014-02-15T23:59:59.000Z

8

Criticality Safety Evaluation for the Advanced Test Reactor U-Mo Demonstration Elements  

SciTech Connect (OSTI)

The Reduced Enrichment Research Test Reactors (RERTR) fuel development program is developing a high uranium density fuel based on a (LEU) uranium-molybdenum alloy. Testing of prototypic RERTR fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. Two RERTR-Full Size Demonstration fuel elements based on the ATR-Reduced YA elements (all but one plate fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). The two fuel elements will be irradiated in alternating cycles such that only one element is loaded in the reactor at a time. Existing criticality analyses have analyzed Standard (HEU) ATR elements (all plates fueled) from which controls have been derived. This criticality safety evaluation (CSE) documents analysis that determines the reactivity of the Demonstration fuel elements relative to HEU ATR elements and shows that the Demonstration elements are bound by the Standard HEU ATR elements and existing HEU ATR element controls are applicable to the Demonstration elements.

Leland M. Montierth

2010-12-01T23:59:59.000Z

9

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

SciTech Connect (OSTI)

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

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

2013-11-01T23:59:59.000Z

10

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

SciTech Connect (OSTI)

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

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

2011-05-31T23:59:59.000Z

11

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

SciTech Connect (OSTI)

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.

Tokuhiro, Akira; Jones, Byron

2013-09-13T23:59:59.000Z

12

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

SciTech Connect (OSTI)

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

Frank Darmann; Robert Lombaerde; Franco Moriconi; Albert Nelson

2011-10-31T23:59:59.000Z

13

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

SciTech Connect (OSTI)

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

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

2013-05-17T23:59:59.000Z

14

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

SciTech Connect (OSTI)

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.

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

2000-02-01T23:59:59.000Z

15

Preliminary neutronics design of china lead-alloy cooled demonstration reactor (CLEAR-III) for nuclear waste transmutation  

SciTech Connect (OSTI)

China Lead-Alloy cooled Demonstration Reactor (CLEAR-III), which is the concept of lead-bismuth cooled accelerator driven sub-critical reactor for nuclear waste transmutation, was proposed and designed by FDS team in China. In this study, preliminary neutronics design studies have primarily focused on three important performance parameters including Transmutation Support Ratio (TSR), effective multiplication factor and blanket thermal power. The constraint parameters, such as power peaking factor and initial TRU loading, were also considered. In the specific design, uranium-free metallic dispersion fuel of (TRU-Zr)-Zr was used as one of the CLEAR-III fuel types and the ratio between MA and Pu was adjusted to maximize transmutation ratio. In addition, three different fuel zones differing in the TRU fraction of the fuel were respectively employed for this subcritical reactor, and the zone sizes and TRU fractions were determined such that the linear powers of these zones were close to each other. The neutronics calculations and analyses were performed by using Multi-Functional 4D Neutronics Simulation System named VisualBUS and nuclear data library HENDL (Hybrid Evaluated Nuclear Data Library). In the preliminary design, the maximum TSRLLMA was {approx}11 and the blanket thermal power was {approx}1000 MW when the effective multiplication factor was 0.98. The results showed that good performance of transmutation could be achieved based on the subcritical reactor loaded with uranium-free fuel. (authors)

Chen, Z. [Inst. of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Southwest Science and Technology Univ., No.350 Shushanhu Road, Shushan District, Hefei, Anhui, 230031 (China); Chen, Y.; Bai, Y.; Wang, W.; Chen, Z.; Hu, L.; Long, P. [Inst. of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, Univ. of Science and Technology of China, Hefei, Anhui, 230031 (China)

2012-07-01T23:59:59.000Z

16

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

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

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

17

Fort Meade demonstration test LEDS in freezer rooms, fiber optics in display cases  

SciTech Connect (OSTI)

Demonstration projects at Fort George G. Meade, MD, substituted LED lighting for incandescent bulbs in commisary wal-in freezers and fiber optic lighting in reach-in display cases. The goal was to reduce energy consumption and the results were positive. Journal article published in Public Works Digest

Parker, Steven; Parker, Graham B.

2008-10-25T23:59:59.000Z

18

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

SciTech Connect (OSTI)

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.

Gastelum, Zoe N.; Henry, Michael J.

2013-11-13T23:59:59.000Z

19

An Overview of the Safety Case for Small Modular Reactors  

SciTech Connect (OSTI)

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.

Ingersoll, Daniel T [ORNL] [ORNL

2011-01-01T23:59:59.000Z

20

E-Print Network 3.0 - assessing reactor reliability Sample Search...  

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

Technology Through Flexibility: The Case of a Demonstration Accelerator-Driven Subcritical Reactor... technology facing significant technological uncertainty, the...

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


21

6.8 Cylinder Liner Boring Case Study Demonstrating the Process of SPC The following case study demonstrates the application of X and R control charts to the study of  

E-Print Network [OSTI]

6.8 Cylinder Liner Boring Case Study ­ Demonstrating the Process of SPC The following case study the complete process of SPC to bring a process into a state of statistical control. The Situation As part

Di Bucchianico, Alessandro

22

The BEI hydrolysis process and reactor system refined engineering proto-type. BEI pilot-plant improvement and operations demonstrations  

SciTech Connect (OSTI)

This BEI project involves BEI-HP and RS's applications toward potential commercial validity demonstrations for dilute-acid corn-fiber cellulose-hydrolysis processing with an aim toward fuel ethanol production.

Brelsford, Donald L.

1999-10-01T23:59:59.000Z

23

Analytical support for the ORR (Oak Ridge Research Reactor) whole-core LEU U/sub 3/Si/sub 2/-Al fuel demonstration  

SciTech Connect (OSTI)

Analytical methods used to analyze neutronic data from the whole-core LEU fuel demonstration in the Oak Ridge Research Reactor are briefly discussed. Calculated eigenvalues corresponding to measured critical control rod positions are presented for each core used in the gradual transition from an all HEU to an all LEU configuration. Some calculated and measured results, including ..beta../sub eff//l/sub p/, are compared for HEU and LEU fresh fuel criticals. Finally, the perturbing influences of the six voided beam tubes on certain core parameters are examined. For reasons yet to be determined, differential shim rod worths are not well-calculated in partially burned cores.

Bretscher, M.M.

1986-01-01T23:59:59.000Z

24

The case for using weight-of-evidence demonstrations in state implementation planning  

SciTech Connect (OSTI)

In an effort to move beyond the current approach of relying heavily on modeling for air quality analyses, the Clean Air Act Advisory Committee's Air Quality Management Workgroup recently recommended that 'weight-of-evidence' demonstrations be used for state implementation planning purposes. This article discusses the workgroup's recommendation and illustrates how such demonstrations can be used to provide comprehensive air quality assessments. 11 refs., 1 tab.

Michael Koerber; Donna Kenski [Lake Michigan Air Directors Consortium in Des Plaines, IL (US)

2005-04-01T23:59:59.000Z

25

Microchannel Reactor System Design & Demonstration For On-Site H2O2 Production by Controlled H2/O2 Reaction  

SciTech Connect (OSTI)

We successfully demonstrated an innovative hydrogen peroxide (H2O2) production concept which involved the development of flame- and explosion-resistant microchannel reactor system for energy efficient, cost-saving, on-site H2O2 production. We designed, fabricated, evaluated, and optimized a laboratory-scale microchannel reactor system for controlled direct combination of H2 and O2 in all proportions including explosive regime, at a low pressure and a low temperature to produce about 1.5 wt% H2O2 as proposed. In the second phase of the program, as a prelude to full-scale commercialization, we demonstrated our H2O2 production approach by ‘numbering up’ the channels in a multi-channel microreactor-based pilot plant to produce 1 kg/h of H2O2 at 1.5 wt% as demanded by end-users of the developed technology. To our knowledge, we are the first group to accomplish this significant milestone. We identified the reaction pathways that comprise the process, and implemented rigorous mechanistic kinetic studies to obtain the kinetics of the three main dominant reactions. We are not aware of any such comprehensive kinetic studies for the direct combination process, either in a microreactor or any other reactor system. We showed that the mass transfer parameter in our microreactor system is several orders of magnitude higher than what obtains in the macroreactor, attesting to the superior performance of microreactor. A one-dimensional reactor model incorporating the kinetics information enabled us to clarify certain important aspects of the chemistry of the direct combination process as detailed in section 5 of this report. Also, through mathematical modeling and simulation using sophisticated and robust commercial software packages, we were able to elucidate the hydrodynamics of the complex multiphase flows that take place in the microchannel. In conjunction with the kinetics information, we were able to validate the experimental data. If fully implemented across the whole industry as a result of our technology demonstration, our production concept is expected to save >5 trillion Btu/year of steam usage and >3 trillion Btu/year in electric power consumption. Our analysis also indicates >50 % reduction in waste disposal cost and ~10% reduction in feedstock energy. These savings translate to ~30% reduction in overall production and transportation costs for the $1B annual H2O2 market.

Adeniyi Lawal

2008-12-09T23:59:59.000Z

26

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

SciTech Connect (OSTI)

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.

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

2012-12-01T23:59:59.000Z

27

Case histories in the Europe/Africa area demonstrate improved capabilities of fiber-optic video camera technology  

SciTech Connect (OSTI)

Cost constraints in the oilfield restrict the performance of remedial services unless a high probability of success can be assured. Unfortunately, a method that could accurately diagnose wellbore problems in a broad scope of oilfield environments was not available, and until the 90`s, accuracy of problem assessment was compromised to the point that estimates of remedial success could not be determined. During the 90`s, however, a downhole video system that showed promise of providing the sought-after diagnostic accuracy for today`s operational conditions was introduced to the oilfield. With the combined capabilities of fiber-optic signal transmission and downhole video camera technology, the system can now provide the oil and gas industry with a diagnostic tool that is rapidly becoming invaluable in oilfield services and is proving the worth of the old adage, seeing is believing. This paper will present a brief overview of the use of the downhole video (DHV) camera technique from its first documented usage until early 1992. The enhancements that have been instrumental in enlarging the scope of its usage since that time will then be discussed. Case histories recorded by the video camera operation in the Europe/Africa area will be presented and analyzed to demonstrate usage and benefits of the system.

Olsen, J.E.; Kristensen, R.; Taylor, R.W.

1995-10-01T23:59:59.000Z

28

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)

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.

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

2013-09-01T23:59:59.000Z

29

Fast Reactor Fuel Type and Reactor Safety Performance  

SciTech Connect (OSTI)

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

R. Wigeland; J. Cahalan

2009-09-01T23:59:59.000Z

30

Portfolio for fast reactor collaboration  

SciTech Connect (OSTI)

The development of the LMFBR type reactor in the United Kingdom is reviewed. Design characteristics of a commercial demonstration fast reactor are presented and compared with the Super Phenix reactor.

Rippon, S.

1981-12-01T23:59:59.000Z

31

Product Demonstrations  

Broader source: Energy.gov [DOE]

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...

32

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

SciTech Connect (OSTI)

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

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

2013-07-01T23:59:59.000Z

33

Making It Easier To Be Green: A Single Case Demonstration of the Effects of Computer Defaults To Conserve Energy in a University Computer Lab  

E-Print Network [OSTI]

KU ScholarWorks | http://kuscholarworks.ku.edu Making It Easier To Be Green: A Single Case Demonstration of the Effects of Computer Defaults To Conserve Energy in a University Computer Lab by Jason M. Hirst et al. KU ScholarWorks is a service... provided by the KU Libraries’ Office of Scholarly Communication & Copyright. This is the published version of the article, made available with the permission of the publisher. The original published version can be found at the link below. Jason M. Hirst...

Hirst, Jason M.; Reed, Derek D.; Kaplan, Brent A.; Miller, Jonathan R.

2013-12-06T23:59:59.000Z

34

GATEWAY Demonstrations  

Broader source: Energy.gov [DOE]

DOE GATEWAY demonstrations showcase high-performance LED products for general illumination in a variety of commercial and residential applications. Demonstration results provide real-world experience and data on state-of-the-art solid-state lighting (SSL) product performance and cost effectiveness. These results connect DOE technology procurement efforts with large-volume purchasers and provide buyers with reliable data on product performance.

35

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

SciTech Connect (OSTI)

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

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

2007-05-02T23:59:59.000Z

36

Antineutrino monitoring for the Iranian heavy water reactor  

E-Print Network [OSTI]

In this note we discuss the potential application of antineutrino monitoring to the Iranian heavy water reactor at Arak, the IR-40, as a non-proliferation measure. We demonstrate that an above ground detector positioned right outside the IR-40 reactor building could meet and in some cases significantly exceed the verification goals identified by IAEA for plutonium production or diversion from declared inventories. In addition to monitoring the reactor during operation, observing antineutrino emissions from long-lived fission products could also allow monitoring the reactor when it is shutdown. Antineutrino monitoring could also be used to distinguish different levels of fuel enrichment. Most importantly, these capabilities would not require a complete reactor operational history and could provide a means to re-establish continuity of knowledge in safeguards conclusions should this become necessary.

Christensen, Eric; Jaffke, Patrick; Shea, Thomas

2014-01-01T23:59:59.000Z

37

Antineutrino monitoring for the Iranian heavy water reactor  

E-Print Network [OSTI]

In this note we discuss the potential application of antineutrino monitoring to the Iranian heavy water reactor at Arak, the IR-40, as a non-proliferation measure. We demonstrate that an above ground detector positioned right outside the IR-40 reactor building could meet and in some cases significantly exceed the verification goals identified by IAEA for plutonium production or diversion from declared inventories. In addition to monitoring the reactor during operation, observing antineutrino emissions from long-lived fission products could also allow monitoring the reactor when it is shutdown. Antineutrino monitoring could also be used to distinguish different levels of fuel enrichment. Most importantly, these capabilities would not require a complete reactor operational history and could provide a means to re-establish continuity of knowledge in safeguards conclusions should this become necessary.

Eric Christensen; Patrick Huber; Patrick Jaffke; Thomas Shea

2014-03-27T23:59:59.000Z

38

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)

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 status and plans are presented.

Frederik Reitsma; Gerhard Strydom; Bismark Tyobeka; Kostadin Ivanov

2012-10-01T23:59:59.000Z

39

Reactor Physics  

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

Reactor Physics Reactor and nuclear physics is a key area of research at INL. Much of the research done in reactor physics can be separated into one of three categories:...

40

Impacts of Static Pressure Reset on VAV System Air Leakage, Fan Power and Thermal Energy - Part 2: Case Demonstration for a Typical Climate System  

E-Print Network [OSTI]

. In this part, a simulated air handling unit (AHU) system in Omaha NE is used to demonstrate the energy savings performance in one typical climate year. This AHU system has a static pressure reset system and two constant static pressure systems, one having...

Liu, M.; Zheng, K.; Wu, L.; Wang, Z.; Johnson, C.

2007-01-01T23:59:59.000Z

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


41

Reactor Safety Research Programs  

SciTech Connect (OSTI)

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.

Edler, S. K.

1981-07-01T23:59:59.000Z

42

Novel Catalytic Membrane Reactors  

SciTech Connect (OSTI)

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.

Stuart Nemser, PhD

2010-10-01T23:59:59.000Z

43

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

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

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...

44

Imaging Fukushima Daiichi reactors with muons  

SciTech Connect (OSTI)

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.

Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.; Milner, Edward C.; Morris, Christopher L. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Lukic, Zarija [Computational Cosmology Center, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Masuda, Koji [University of New Mexico, Albuquerque, NM 87131 (United States); Perry, John O. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); University of New Mexico, Albuquerque, NM 87131 (United States)

2013-05-15T23:59:59.000Z

45

Reactor Safety Research Programs  

SciTech Connect (OSTI)

This document summarizes the work performed by Pacific Northwest laboratory from October 1 through December 31, 1979, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, lspra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

Dotson, CW

1980-08-01T23:59:59.000Z

46

Reactor technology assessment and selection utilizing systems engineering approach  

SciTech Connect (OSTI)

The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

Zolkaffly, Muhammed Zulfakar; Han, Ki-In [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2014-02-12T23:59:59.000Z

47

Fluoride Salt-Cooled High-Temperature Reactor Development Roadmap  

SciTech Connect (OSTI)

Fluoride salt-cooled high-temperature reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics and fully passive safety. This paper provides an overview of a technology development pathway for expeditious commercial deployment of first-generation FHRs. The paper describes the principal remaining FHR technology challenges and the development path needed to address the challenges. First-generation FHRs do not appear to require any technology breakthroughs, but will require significant technology development and demonstration. FHRs are currently entering early phase engineering development. As such, the development roadmap is not as technically detailed or specific as would be the case for a more mature reactor class. The higher cost of fuel and coolant; the lack of an approved licensing framework; the lack of qualified, salt-compatible structural materials; and the potential for tritium release into the environment are the most obvious issues that remain to be resolved.

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

2014-01-01T23:59:59.000Z

48

Demonstration Assessment of LED Freezer Case Lighting  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-UpHeat Pump Models | Department1 Prepared1217 Release Notes

49

Potential of Mirror Systems as Future Fusion Power Reactors  

SciTech Connect (OSTI)

Mirror based fusion reactors - as other fusion reactor concepts - have considerable environmental and safety advantages. They could make available energy resources for many 1000 years. Mirror type fusion reactors have additional technical advantages over other fusion reactor concepts. These are: simple design topology, steady state power generation, decoupling of end plugs from central power producing regions, small power units as demonstration facilities.

Kessler, Guenter; Kulcinski, Gerald L. [University of Madison (United States)

2005-01-15T23:59:59.000Z

50

Small Reactor for Deep Space Exploration  

SciTech Connect (OSTI)

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.

None

2012-11-29T23:59:59.000Z

51

Small Reactor for Deep Space Exploration  

ScienceCinema (OSTI)

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.

None

2014-05-30T23:59:59.000Z

52

Thermionic Reactor Design Studies  

SciTech Connect (OSTI)

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.

Schock, Alfred

1994-08-01T23:59:59.000Z

53

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.

54

Initial Modeling of a Pressurized Water Reactor Completed Using RELAP-7  

Broader source: Energy.gov [DOE]

RELAP-7 is a nuclear reactor system safety analysis code where initial capabilities were demonstrated by simulating a steady-state single-phase pressurized water reactor (PWR) with two parallel loops and multiple reactor core flow channels.

55

Nuclear reactor engineering  

SciTech Connect (OSTI)

Chapters are presented concerning energy from nuclear fission; nuclear reactions and radiations; diffusion and slowing-down of neutrons; principles of reactor analysis; nuclear reactor kinetics and control; energy removal; non-fuel reactor materials; the reactor fuel system; radiation protection and environmental effects; nuclear reactor shielding; nuclear reactor safety; and power reactor systems.

Glasstone, S.; Sesonske, A.

1981-01-01T23:59:59.000Z

56

Automated Test Coverage Measurement for Reactor Protection System Software  

E-Print Network [OSTI]

Automated Test Coverage Measurement for Reactor Protection System Software Implemented in Function- ing a case study using test cases prepared by domain experts for reactor protection system software) are widely used to implement safety- critical systems such as nuclear reactor protection systems, testing

57

Research reactors - an overview  

SciTech Connect (OSTI)

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

West, C.D.

1997-03-01T23:59:59.000Z

58

Prototypical Consolidation Demonstration Project: Final report  

SciTech Connect (OSTI)

This is the final report of the Prototypical Consolidation Demonstration Project, which was funded by the US Department of Energy`s Office of Civilian Radioactive Waste Management. The project had two objectives: (a) to develop and demonstrate a prototype of production-scale equipment for the dry, horizontal consolidation and packaging of spent nuclear fuel rods from commercial boiling water reactor and pressurized water reactor fuel assemblies, and (b) to report the development and demonstration results to the US Department of Energy, Idaho Operations Office. This report summarizes the activities and conclusions of the project management contractor, EG&G Idaho, Inc., and the fabrication and testing contractor, NUS Corporation (NUS). The report also presents EG&G Idaho`s assessments of the equipment and procedures developed by NUS.

Gili, J.A.; Poston, V.K.

1993-11-01T23:59:59.000Z

59

Light Water Reactor Sustainability  

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

3 Light Water Reactor Sustainability Program ACCOMPLISHMENTS REPORT 2013 Accomplishments Report | Light Water Reactor Sustainability 2 T he mission of the Light Water Reactor...

60

Light Water Reactor Sustainability  

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

4 Light Water Reactor Sustainability ACCOMPLISHMENTS REPORT 2014 Accomplishments Report | Light Water Reactor Sustainability 2 T he mission of the Light Water Reactor...

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


61

Catalytic reactor  

DOE Patents [OSTI]

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.

Aaron, Timothy Mark (East Amherst, NY); Shah, Minish Mahendra (East Amherst, NY); Jibb, Richard John (Amherst, NY)

2009-03-10T23:59:59.000Z

62

Bioconversion reactor  

DOE Patents [OSTI]

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.

McCarty, Perry L. (Stanford, CA); Bachmann, Andre (Palo Alto, CA)

1992-01-01T23:59:59.000Z

63

aeew fast reactor: Topics by E-print Network  

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

reactors. Here we investigate the special case of mixtures of lithium fluoride and thorium fluoride, which act both as coolant and fuel in the molten salt fast reactor concept....

64

LIMB Demonstration Project Extension and Coolside Demonstration  

SciTech Connect (OSTI)

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.

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

1992-11-10T23:59:59.000Z

65

Neutronic reactor  

DOE Patents [OSTI]

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.

Wende, Charles W. J. (Augusta, GA); Babcock, Dale F. (Wilmington, DE); Menegus, Robert L. (Wilmington, DE)

1983-01-01T23:59:59.000Z

66

Strategy Guideline: Demonstration Home  

SciTech Connect (OSTI)

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.

Savage, C.; Hunt, A.

2012-12-01T23:59:59.000Z

67

Chevrolet Volt Vehicle Demonstration  

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

Chevrolet Volt Vehicle Demonstration Fleet Summary Report Reporting period: October 2011 through December 2011 Number of vehicles: 135 Number of vehicle days driven: 4,746 All...

68

Spherical torus fusion reactor  

DOE Patents [OSTI]

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.

Peng, Yueng-Kay M. (Oak Ridge, TN)

1989-01-01T23:59:59.000Z

69

CALDERON COKEMAKING PROCESS/DEMONSTRATION PROJECT  

SciTech Connect (OSTI)

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.

ALBERT CALDERON

1998-06-22T23:59:59.000Z

70

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

E-Print Network [OSTI]

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

Rodriguez, Judy N

2013-01-01T23:59:59.000Z

71

Stochastic solution of population balance equations for reactor networks  

SciTech Connect (OSTI)

This work presents a sequential modular approach to solve a generic network of reactors with a population balance model using a stochastic numerical method. Full-coupling to the gas-phase is achieved through operator-splitting. The convergence of the stochastic particle algorithm in test networks is evaluated as a function of network size, recycle fraction and numerical parameters. These test cases are used to identify methods through which systematic and statistical error may be reduced, including by use of stochastic weighted algorithms. The optimal algorithm was subsequently used to solve a one-dimensional example of silicon nanoparticle synthesis using a multivariate particle model. This example demonstrated the power of stochastic methods in resolving particle structure by investigating the transient and spatial evolution of primary polydispersity, degree of sintering and TEM-style images. Highlights: •An algorithm is presented to solve reactor networks with a population balance model. •A stochastic method is used to solve the population balance equations. •The convergence and efficiency of the reported algorithms are evaluated. •The algorithm is applied to simulate silicon nanoparticle synthesis in a 1D reactor. •Particle structure is reported as a function of reactor length and time.

Menz, William J.; Akroyd, Jethro; Kraft, Markus

2014-01-01T23:59:59.000Z

72

Core Drilling Demonstration  

Broader source: Energy.gov [DOE]

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...

73

Chevrolet Volt Vehicle Demonstration  

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

Volt Vehicle Demonstration Fleet Summary Report Reporting period: January 2013 through March 2013 Number of vehicles: 146 Number of vehicle days driven: 6,680 4292013 2:38:13 PM...

74

Molten-Salt Depleted-Uranium Reactor  

E-Print Network [OSTI]

The supercritical, reactor core melting and nuclear fuel leaking accidents have troubled fission reactors for decades, and greatly limit their extensive applications. Now these troubles are still open. Here we first show a possible perfect reactor, Molten-Salt Depleted-Uranium Reactor which is no above accident trouble. We found this reactor could be realized in practical applications in terms of all of the scientific principle, principle of operation, technology, and engineering. Our results demonstrate how these reactors can possess and realize extraordinary excellent characteristics, no prompt critical, long-term safe and stable operation with negative feedback, closed uranium-plutonium cycle chain within the vessel, normal operation only with depleted-uranium, and depleted-uranium high burnup in reality, to realize with fission nuclear energy sufficiently satisfying humanity long-term energy resource needs, as well as thoroughly solve the challenges of nuclear criticality safety, uranium resource insuffic...

Dong, Bao-Guo; Gu, Ji-Yuan

2015-01-01T23:59:59.000Z

75

International Research Reactor Decommissioning Project  

SciTech Connect (OSTI)

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.

Leopando, Leonardo [Philippine Nuclear Research Institute, Quezon City (Philippines); Warnecke, Ernst [International Atomic Energy Agency, Vienna (Austria)

2008-01-15T23:59:59.000Z

76

Hybrid adsorptive membrane reactor  

DOE Patents [OSTI]

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.

Tsotsis, Theodore T. (Huntington Beach, CA); Sahimi, Muhammad (Altadena, CA); Fayyaz-Najafi, Babak (Richmond, CA); Harale, Aadesh (Los Angeles, CA); Park, Byoung-Gi (Yeosu, KR); Liu, Paul K. T. (Lafayette Hill, PA)

2011-03-01T23:59:59.000Z

77

Nuclear reactor engineering  

SciTech Connect (OSTI)

A book is reviewed which emphasizes topics directly related to the light water reactor power plant and the fast reactor power system. Current real-world problems are addressed throughout the text, and a chapter on safety includes much of the postThree Mile Island impact on operating systems. Topics covered include Doppler broadening, neutron resonances, multigroup diffusion theory, reactor kinetics, reactor control, energy removal, nonfuel materials, reactor fuel, radiation protection, environmental effects, and reactor safety.

Glasstone, S.; Sesonske, A.

1982-07-01T23:59:59.000Z

78

New Technology Demonstration Program  

E-Print Network [OSTI]

New Technology Demonstration Program Technical Brief FEMPFederal Energy Management Program Tom for saving energy in refrigerated walk-in coolers, and to evaluate the potential for this technology in Federal facilities. The focus of this study was on a single manufacturer of the technology, Nevada Energy

79

MAJORANA Demonstrator Motivation  

E-Print Network [OSTI]

1 #12;OVERVIEW MAJORANA Demonstrator Motivation Neutrinoless double beta decay Search for axions: MAJORANA Collaboration #12;NEUTRINOLESS DOUBLE BETA DECAY Emission of 2 electrons from Ge-76 and application to neutrinoless double beta decay search in Ge- 76." Journal of Instrumentation 6 (2011).13 #12

Washington at Seattle, University of - Department of Physics, Electroweak Interaction Research Group

80

Nucla CFB Demonstration Project  

SciTech Connect (OSTI)

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)

Not Available

1990-12-01T23:59:59.000Z

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


81

LIMB Demonstration Project Extension  

SciTech Connect (OSTI)

The basic goal of the Limestone Injection Multistage Burner (LIMB) demonstration is to extend LIMB technology development to a full- scale application on a representative wall-fired utility boiler. The successful retrofit of LIMB to an existing boiler is expected to demonstrate that (a) reductions of 50 percent or greater in SO{sub x} and NO{sub x} emissions can be achieved at a fraction of the cost of add-on FGD systems, (b) boiler reliability, operability, and steam production can be maintained at levels existing prior to LIMB retrofit, and (c) technical difficulties attributable to LIMB operation, such as additional slagging and fouling, changes in ash disposal requirements, and an increased particulate load, can be resolved in a cost-effective manner. The primary fuel to be used will be an Ohio bituminous coal having a nominal sulfur content of 3 percent or greater.

Not Available

1989-06-15T23:59:59.000Z

82

Reactor safety method  

DOE Patents [OSTI]

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

Vachon, Lawrence J. (Clairton, PA)

1980-03-11T23:59:59.000Z

83

SRS Small Modular Reactors  

SciTech Connect (OSTI)

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

None

2012-04-27T23:59:59.000Z

84

SRS Small Modular Reactors  

ScienceCinema (OSTI)

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

None

2014-05-21T23:59:59.000Z

85

Residential Transactive Control Demonstration  

SciTech Connect (OSTI)

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.

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

2014-02-19T23:59:59.000Z

86

PFBC Utility Demonstration Project  

SciTech Connect (OSTI)

This report provides a summary of activities by American Electric Power Service Corporation during the first budget period of the PFBC Utility Demonstration Project. In April 1990, AEP signed a Cooperative Agreement with the US Department of Energy to repower the Philip Sporn Plant, Units 3 4 in New Haven, West Virginia, with a 330 KW PFBC plant. The purpose of the program was to demonstrate and verify PFBC in a full-scale commercial plant. The technical and cost baselines of the Cooperative Agreement were based on a preliminary engineering and design and a cost estimate developed by AEP subsequent to AEP's proposal submittal in May 1988, and prior to the signing of the Cooperative Agreement. The Statement of Work in the first budget period of the Cooperative Agreement included a task to develop a preliminary design and cost estimate for erecting a Greenfield plant and to conduct a comparison with the repowering option. The comparative assessment of the options concluded that erecting a Greenfield plant rather than repowering the existing Sporn Plant could be the technically and economically superior alternative. The Greenfield plant would have a capacity of 340 MW. The ten additional MW output is due to the ability to better match the steam cycle to the PFBC system with a new balance of plant design. In addition to this study, the conceptual design of the Sporn Repowering led to several items which warranted optimization studies with the goal to develop a more cost effective design.

Not Available

1992-11-01T23:59:59.000Z

87

DOE`s annealing prototype demonstration projects  

SciTech Connect (OSTI)

One of the challenges U.S. utilities face in addressing technical issues associated with the aging of nuclear power plants is the long-term effect of plant operation on reactor pressure vessels (RPVs). As a nuclear plant operates, its RPV is exposed to neutrons. For certain plants, this neutron exposure can cause embrittlement of some of the RPV welds which can shorten the useful life of the RPV. This RPV embrittlement issue has the potential to affect the continued operation of a number of operating U.S. pressurized water reactor (PWR) plants. However, RPV material properties affected by long-term irradiation are recoverable through a thermal annealing treatment of the RPV. Although a dozen Russian-designed RPVs and several U.S. military vessels have been successfully annealed, U.S. utilities have stated that a successful annealing demonstration of a U.S. RPV is a prerequisite for annealing a licensed U.S. nuclear power plant. In May 1995, the Department of Energy`s Sandia National Laboratories awarded two cost-shared contracts to evaluate the feasibility of annealing U.S. licensed plants by conducting an anneal of an installed RPV using two different heating technologies. The contracts were awarded to the American Society of Mechanical Engineers (ASME) Center for Research and Technology Development (CRTD) and MPR Associates (MPR). The ASME team completed its annealing prototype demonstration in July 1996, using an indirect gas furnace at the uncompleted Public Service of Indiana`s Marble Hill nuclear power plant. The MPR team`s annealing prototype demonstration was scheduled to be completed in early 1997, using a direct heat electrical furnace at the uncompleted Consumers Power Company`s nuclear power plant at Midland, Michigan. This paper describes the Department`s annealing prototype demonstration goals and objectives; the tasks, deliverables, and results to date for each annealing prototype demonstration; and the remaining annealing technology challenges.

Warren, J.; Nakos, J.; Rochau, G.

1997-02-01T23:59:59.000Z

88

Nuclear reactor  

DOE Patents [OSTI]

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.

Thomson, Wallace B. (Severna Park, MD)

2004-03-16T23:59:59.000Z

89

Thermionic Reactor Design Studies  

SciTech Connect (OSTI)

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-core-height diodes. Moreover, placing the fuel on the outside of the diode makes possible reactors with much higher fuel volume fractions, which enable power-flattened fast reactors scalable to very low power levels without the need for life-limiting hydride moderators or the use of efficiency-limiting driver fuel. In addition, with the fuel on the outside its swelling does not increase the emitter diameter or reduce the interelectrode gap. This should permit long lifetimes even with closer spacings, which can significantly improve the system efficiences. This was confirmed by coupled neutronic, thermal, thermionic, and electrical system analyses - some of which are presented in this paper - and by subsequent experiments. A companion paper presented next describes the fabrication and testing of full-scale converter elements, both fueled and unfueled, and summarizes the test results obtained. There is a duplicate copy in the file.

Schock, Alfred

1994-06-01T23:59:59.000Z

90

Operational control of boiling water reactor stability  

SciTech Connect (OSTI)

Boiling water reactor cores are susceptible to instabilities, which generate power oscillations. Specific reactor operating practices can provide a mechanism for control of the instability phenomenon. An axial separation of the core into a single-phase region and a two-phase region resolves the influence of axial flux shapes on core stability. This separation provides the means to derive a core stability control that ensures significant reactor stability margin. The control is achieved by maintaining the core average bulk coolant saturation elevation above a predetermined axial plane. The control can be reliably and efficiently implemented during reactor operations. Analysis demonstrates that variations in parameters important to stability have only secondary influences on stability margin when the control is in effect. Actual plant experience with a large commercial boiling water reactor confirms the capabilities of this stability control in an operational setting.

Mowry, C.M. [PECO Energy, Wayne, PA (United States); Nir, I. [Entergy Operations, Jackson, MS (United States); Newkirk, D.W. [GE Nuclear Energy, San Jose, CA (United States)

1995-03-01T23:59:59.000Z

91

Jennings Demonstration PLant  

SciTech Connect (OSTI)

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.

Russ Heissner

2010-08-31T23:59:59.000Z

92

Undergraduate reactor control experiment  

SciTech Connect (OSTI)

A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise.

Edwards, R.M.; Power, M.A.; Bryan, M. (Pennsylvania State Univ., University Park (United States))

1992-01-01T23:59:59.000Z

93

Attrition reactor system  

DOE Patents [OSTI]

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.

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

1993-01-01T23:59:59.000Z

94

Attrition reactor system  

DOE Patents [OSTI]

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.

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

1993-09-28T23:59:59.000Z

95

Reactor Sharing Program  

SciTech Connect (OSTI)

Progress achieved at the University of Florida Training Reactor (UFTR) facility through the US Department of Energy's University Reactor Sharing Program is reported for the period of 1991--1992.

Vernetson, W.G.

1993-01-01T23:59:59.000Z

96

High solids fermentation reactor  

DOE Patents [OSTI]

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.

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

1993-03-02T23:59:59.000Z

97

Improved vortex reactor system  

DOE Patents [OSTI]

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.

Diebold, James P. (Lakewood, CO); Scahill, John W. (Evergreen, CO)

1995-01-01T23:59:59.000Z

98

Advanced Test Reactor Tour  

SciTech Connect (OSTI)

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

Miley, Don

2011-01-01T23:59:59.000Z

99

Advanced Test Reactor Tour  

ScienceCinema (OSTI)

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

Miley, Don

2013-05-28T23:59:59.000Z

100

High solids fermentation reactor  

DOE Patents [OSTI]

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.

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

1993-01-01T23:59:59.000Z

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


101

Hypothetical Reactor Accident Study  

E-Print Network [OSTI]

- W 4 DfcSkoollo Rise-R-427 CARNSORE: Hypothetical Reactor Accident Study O. Walmod-Larsen, N. O: HYPOTHETICAL REACTOR ACCIDENT STUDY O. Walmod-Larsen, N.O. Jensen, L. Kristensen, A. Heide, K.L. NedergĂĄrd, P-basis accident and a series of hypothetical core-melt accidents to a 600 MWe reactor are de- scribed

102

The Integral Fast Reactor: A practical approach to waste management  

SciTech Connect (OSTI)

This report discusses development of the method for pyroprocessing of spent fuel from the Integral Fast Reactor (or Advanced Liquid Metal Reactor). The technology demonstration phase, in which recycle will be demonstrated with irradiated fuel from the EBR-II reactor has been reached. Methods for recovering actinides from spent LWR fuel are at an earlier stage of development but appear to be technically feasible at this time, and a large-scale demonstration of this process has begun. The utilization of fully compatible processes for recycling valuable spent fuel materials promises to provide substantial economic incentives for future applications of the pyroprocessing technology.

Laidler, J.J.

1993-12-31T23:59:59.000Z

103

Neutron behavior, reactor control, and reactor heat transfer. Volume four  

SciTech Connect (OSTI)

Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant).

Not Available

1986-01-01T23:59:59.000Z

104

Reactor vessel support system  

DOE Patents [OSTI]

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.

Golden, Martin P. (Trafford, PA); Holley, John C. (McKeesport, PA)

1982-01-01T23:59:59.000Z

105

Spinning fluids reactor  

DOE Patents [OSTI]

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.

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

2012-11-20T23:59:59.000Z

106

Determining Reactor Neutrino Flux  

E-Print Network [OSTI]

Flux is an important source of uncertainties for a reactor neutrino experiment. It is determined from thermal power measurements, reactor core simulation, and knowledge of neutrino spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle $\\theta_{13}$ by reactor neutrino experiments in the coming years will use near-far detector configurations. Most uncertainties from reactor will be canceled out. Understanding of the correlation of uncertainties is required for $\\theta_{13}$ experiments. Precise determination of reactor neutrino flux will also improve the sensitivity of the non-proliferation monitoring and future reactor experiments. We will discuss the flux calculation and recent progresses.

Jun Cao

2012-03-08T23:59:59.000Z

107

Reactor water cleanup system  

DOE Patents [OSTI]

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.

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

1994-12-20T23:59:59.000Z

108

Reactor water cleanup system  

DOE Patents [OSTI]

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.

Gluntz, Douglas M. (San Jose, CA); Taft, William E. (Los Gatos, CA)

1994-01-01T23:59:59.000Z

109

CALDERON COKEMAKING PROCESS/DEMONSTRATION PROJECT  

SciTech Connect (OSTI)

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.

ALBERT CALDERON

1998-09-22T23:59:59.000Z

110

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

SciTech Connect (OSTI)

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.

Vinson, Dennis

2010-06-01T23:59:59.000Z

111

POTENTIAL BENCHMARKS FOR ACTINIDE PRODUCTION IN HANFORD REACTORS  

SciTech Connect (OSTI)

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.

PUIGH RJ; TOFFER H

2011-10-19T23:59:59.000Z

112

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

SciTech Connect (OSTI)

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 increases along with the LCOE, the projects may become unprofitable. This is the case at the LCOE of $103 $109/MW, in which the NPV became negative. The IRR increased with increasing electricity price. Three cases, electric only base, storage—compressed air energy storage or pumped hydro, and hydrogen production, were performed incorporating SMRs into a nuclear wind natural gas hybrid energy system for the New York West Central region. The operational costs for three cases were calculated as $27/MWh, $25/MWh, and $28/MWh, respectively. A 3% increase in profits was demonstrated for the storage case over the electric only base case.

Lauren M. Boldon; Piyush Sabharwall

2014-08-01T23:59:59.000Z

113

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

SciTech Connect (OSTI)

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.

Vinson, D.; Webb, R.

2010-09-30T23:59:59.000Z

114

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

SciTech Connect (OSTI)

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.

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

1997-04-01T23:59:59.000Z

115

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

SciTech Connect (OSTI)

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 cost of decommissioning a PWR is lowest for ENTOMB and highest for SAFSTOR • the estimated cost of decommissioning a BWR is lowest for OECON and highest for SAFSTOR. In all cases, SAFSTOR has the lowest occupational radiation dose and the highest cost.

Wittenbrock, N. G.

1982-01-01T23:59:59.000Z

116

Livestock Odor Reduction Demonstration Project  

E-Print Network [OSTI]

Livestock Odor Reduction Demonstration Project Objectives The 1996 and 1997 Iowa General Assembly-share basis to livestock producers and operators selected to carry out various demonstration projects. Organization The Livestock Odor Reduction Demonstration Project was administered by ISU Extension. Stewart

Lin, Zhiqun

117

Adaptive Nodal Transport Methods for Reactor Transient Analysis  

SciTech Connect (OSTI)

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.

Thomas Downar; E. Lewis

2005-08-31T23:59:59.000Z

118

Dynamic behavior of chemical exchange column in a water detritiation system for a fusion reactor  

SciTech Connect (OSTI)

The dynamic behavior of a CECE column used for a demonstration reactor (DEMO) plant has been studied. In the case where the column was filled with natural water, the time required to achieve steady state was almost the same as that for the column operated under the total reflux mode. The manipulated variables were flow rate of the bottom stream for the control of the bottom tritium concentration, and flow rate of the hydrogen stream for the control of the top tritium concentration. For both the variables, the response curve was expressed by the first-order lag system, and a PID controller could be applied. (authors)

Yamanishi, T.; Iwai, Y. [Tritium Engineering Group, JAEA, Tokai, Ibaraki, 319-1195 (Japan)

2008-07-15T23:59:59.000Z

119

Improved vortex reactor system  

DOE Patents [OSTI]

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.

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

1995-05-09T23:59:59.000Z

120

Role of research reactors in training of NPP personnel with special focus on training reactor VR-1  

SciTech Connect (OSTI)

Research reactors play an important role in providing key personnel of nuclear power plants a hands-on experience from operation and experiments at nuclear facilities. Training of NPP (Nuclear Power Plant) staff is usually deeply theoretical with an extensive utilisation of simulators and computer visualisation. But a direct sensing of the reactor response to various actions can only improve the personnel awareness of important aspects of reactor operation. Training Reactor VR-1 and its utilization for training of NPP operators and other professionals from Czech Republic and Slovakia is described. Typical experimental exercises and good practices in organization of a training program are demonstrated. (authors)

Sklenka, L.; Rataj, J.; Frybort, J.; Huml, O. [Dept. of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical Univ. in Prague, V Holesovickach 2, Prague 8, 180 00 (Czech Republic)

2012-07-01T23:59:59.000Z

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


121

The economics of fuel depletion in fast breeder reactor blankets  

E-Print Network [OSTI]

A fast breeder reactor fuel depletion-economics model was developed and applied to a number of 1000 MWe UMBR case studies, involving radial blanket-radial reflector design, radial blanket fuel management, and sensitivity ...

Brewer, Shelby Templeton

1972-01-01T23:59:59.000Z

122

Pressurized fluidized bed reactor  

DOE Patents [OSTI]

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.

Isaksson, J.

1996-03-19T23:59:59.000Z

123

Pressurized fluidized bed reactor  

DOE Patents [OSTI]

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.

Isaksson, Juhani (Karhula, FI)

1996-01-01T23:59:59.000Z

124

Tokamak reactor first wall  

DOE Patents [OSTI]

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.

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

1984-11-20T23:59:59.000Z

125

Next Generation Reactors  

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

Nuclear Advances We are coordinating the Generation IV Nuclear Systems Initiative - an international effort to develop the next generation of nuclear power reactors. Skip...

126

Simultaneous Cyclic Scheduling and Control of Tubular Reactors: Parallel Production Lines  

E-Print Network [OSTI]

Simultaneous Cyclic Scheduling and Control of Tubular Reactors: Parallel Production Lines Antonio scheduling and control formulation is tested using three multiproduct continuous tubular reactors featuring on the simultaneous scheduling and control (SC) of tubular reactors in a single production line [1] to the case

Grossmann, Ignacio E.

127

Boundary control for an industrial under-actuated tubular chemical reactor  

E-Print Network [OSTI]

Ltd. All rights reserved. Keywords: Polystyrene; Tubular reactor; Control; Optimization; IndustrialBoundary control for an industrial under-actuated tubular chemical reactor D. Del Vecchio a , N and studied for an industrial under-actuated tubular chemical reactor. This work presents a case

128

Kinetic Modeling Of Solid-Gas Reactions At Reactor Scale: A General Approach Loc Favergeon1  

E-Print Network [OSTI]

the knowledge of the kinetic model for the calculation of the speed of reaction in one part of the reactorKinetic Modeling Of Solid-Gas Reactions At Reactor Scale: A General Approach LoĂŻc Favergeon1 favergeon@emse.fr ABSTRACT A rigorous simulation of industrial reactors in the case of solid-gas reacting

Paris-Sud XI, Université de

129

Nuclear Reactor Safeguards and Monitoring with Antineutrino Detectors  

E-Print Network [OSTI]

Cubic-meter-sized antineutrino detectors can be used to non-intrusively, robustly and automatically monitor and safeguard a wide variety of nuclear reactor types, including power reactors, research reactors, and plutonium production reactors. Since the antineutrino spectra and relative yields of fissioning isotopes depend on the isotopic composition of the core, changes in composition can be observed without ever directly accessing the core itself. Information from a modest-sized antineutrino detector, coupled with the well-understood principles that govern the core's evolution in time, can be used to determine whether the reactor is being operated in an illegitimate way. A group at Sandia is currently constructing a one cubic meter antineutrino detector at the San Onofre reactor site in California to demonstrate these principles.

Adam Bernstein; Yifang Wang; Giorgio Gratta; Todd West

2001-08-01T23:59:59.000Z

130

Reactor-specific spent fuel discharge projections: 1985 to 2020  

SciTech Connect (OSTI)

The creation of four spent-fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No New Orders with Extended Burnup, (2) No New Orders with Constant Burnup, (3) Middle Case with Extended Burnup, and (4) Middle Case with Constant Burnup. Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel.

Heeb, C.M.; Libby, R.A.; Walling, R.C.; Purcell, W.L.

1986-09-01T23:59:59.000Z

131

A TEN MEGAWATT BOILING HETEROGENEOUS PACKAGE POWER REACTOR. Reactor...  

Office of Scientific and Technical Information (OSTI)

A TEN MEGAWATT BOILING HETEROGENEOUS PACKAGE POWER REACTOR. Reactor Design and Feasibility Problem Re-direct Destination: Temp Data Fields Rosen, M. A.; Coburn, D. B.; Flynn, T....

132

Tandem mirror technology demonstration facility  

SciTech Connect (OSTI)

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.

Not Available

1983-10-01T23:59:59.000Z

133

Demand Response Spinning Reserve Demonstration  

E-Print Network [OSTI]

F) Enhanced ACP Date RAA ACP Demand Response – SpinningReserve Demonstration Demand Response – Spinning Reservesupply spinning reserve. Demand Response – Spinning Reserve

2007-01-01T23:59:59.000Z

134

MDF | Manufacturing Demonstration Facility | ORNL  

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

MDF Working with MDF NTRC OLCF SNS Titanium robotic hand holding sphere fabricated using additive manufacturing Home | User Facilities | MDF MDF | Manufacturing Demonstration...

135

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...

136

Advanced burner test reactor preconceptual design report.  

SciTech Connect (OSTI)

The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an advanced fuel cycle; (2) To qualify the transuranics-containing fuels and advanced structural materials needed for a full-scale ABR; and (3) To support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. The ABTR should also address the following additional objectives: (1) To incorporate and demonstrate innovative design concepts and features that may lead to significant improvements in cost, safety, efficiency, reliability, or other favorable characteristics that could promote public acceptance and future private sector investment in ABRs; (2) To demonstrate improved technologies for safeguards and security; and (3) To support development of the U.S. infrastructure for design, fabrication and construction, testing and deployment of systems, structures and components for the ABRs. Based on these objectives, a pre-conceptual design of a 250 MWt ABTR has been developed; it is documented in this report. In addition to meeting the primary and additional objectives listed above, the lessons learned from fast reactor programs in the U.S. and worldwide and the operating experience of more than a dozen fast reactors around the world, in particular the Experimental Breeder Reactor-II have been incorporated into the design of the ABTR to the extent possible.

Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

2008-12-16T23:59:59.000Z

137

REACTOR OPERATIONS AND CONTROL  

E-Print Network [OSTI]

REACTOR OPERATIONS AND CONTROL KEYWORDS: core calculations, neural networks, control rod elevation of a control rod, or a group of control rods, is an important parameter from the viewpoint of reactor control DETERMINATION OF PWR CONTROL ROD POSITION BY CORE PHYSICS AND NEURAL NETWORK METHODS NINOS S. GARIS* and IMRE

Pázsit, Imre

138

Distributed computing and nuclear reactor analysis  

SciTech Connect (OSTI)

Large-scale scientific and engineering calculations for nuclear reactor analysis can now be carried out effectively in a distributed computing environment, at costs far lower than for traditional mainframes. The distributed computing environment must include support for traditional system services, such as a queuing system for batch work, reliable filesystem backups, and parallel processing capabilities for large jobs. All ANL computer codes for reactor analysis have been adapted successfully to a distributed system based on workstations and X-terminals. Distributed parallel processing has been demonstrated to be effective for long-running Monte Carlo calculations.

Brown, F.B.; Derstine, K.L.; Blomquist, R.N.

1994-03-01T23:59:59.000Z

139

Reactor & Nuclear Systems Publications | ORNL  

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

Nuclear Science Home | Science & Discovery | Nuclear Science | Publications and Reports | Reactor and Nuclear Systems Publications SHARE Reactor and Nuclear Systems Publications...

140

Reed Reactor Facility. Final report  

SciTech Connect (OSTI)

This report discusses the operation and maintenance of the Reed Reactor Facility. The Reed reactor is mostly used for education and train purposes.

Frantz, S.G.

1994-12-31T23:59:59.000Z

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


141

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

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in Review: TopEnergy DOEDealing With the Issues of NuclearHighStatement| Department

142

Irradiation behavior of metallic fast reactor fuels  

SciTech Connect (OSTI)

Metallic fuels were the first fuels chosen for liquid metal cooled fast reactors (LMR's). In the late 1960's world-wide interest turned toward ceramic LMR fuels before the full potential of metallic fuel was realized. However, during the 1970's the performance limitations of metallic fuel were resolved in order to achieve a high plant factor at the Argonne National Laboratory's Experimental Breeder Reactor II. The 1980's spawned renewed interest in metallic fuel when the Integral Fast Reactor (IFR) concept emerged at Argonne National Laboratory. A fuel performance demonstration program was put into place to obtain the data needed for the eventual licensing of metallic fuel. This paper will summarize the results of the irradiation program carried out since 1985.

Pahl, R.G.; Porter, D.L.; Crawford, D.C.; Walters, L.C.

1991-01-01T23:59:59.000Z

143

Nuclear reactor control column  

DOE Patents [OSTI]

The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

Bachovchin, Dennis M. (Plum Borough, PA)

1982-01-01T23:59:59.000Z

144

Nuclear reactor control column  

SciTech Connect (OSTI)

The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest crosssectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

Bachovchin, D.M.

1982-08-10T23:59:59.000Z

145

Decathletes Demonstrate Affordable Solar Housing  

Broader source: Energy.gov [DOE]

Two teams tied for first place in the first-ever Affordability Contest at the 2011 Solar Decathlon -- demonstrating that innovative technology doesn’t have to be expensive.

146

Nuclear reactor reflector  

DOE Patents [OSTI]

A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

Hopkins, Ronald J. (Pensacola, FL); Land, John T. (Pensacola, FL); Misvel, Michael C. (Pensacola, FL)

1994-01-01T23:59:59.000Z

147

Nuclear reactor reflector  

DOE Patents [OSTI]

A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

Hopkins, R.J.; Land, J.T.; Misvel, M.C.

1994-06-07T23:59:59.000Z

148

Fast Breeder Reactor studies  

SciTech Connect (OSTI)

This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

1980-07-01T23:59:59.000Z

149

Spherical torus fusion reactor  

DOE Patents [OSTI]

The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

Martin Peng, Y.K.M.

1985-10-03T23:59:59.000Z

150

Microfluidic electrochemical reactors  

DOE Patents [OSTI]

A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

Nuzzo, Ralph G. (Champaign, IL); Mitrovski, Svetlana M. (Urbana, IL)

2011-03-22T23:59:59.000Z

151

Reactor-specific spent fuel discharge projections, 1987-2020  

SciTech Connect (OSTI)

The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs.

Walling, R.C.; Heeb, C.M.; Purcell, W.L.

1988-03-01T23:59:59.000Z

152

Toms Creek IGCC Demonstration Project  

SciTech Connect (OSTI)

The Toms Creek Integrated Gasification Combined Cycle (IGCC) Demonstration Project was selected by DOE in September 1991 to participate in Round Four of the Clean Coal Technology Demonstration Program. The project will demonstrate a simplified IGCC process consisting of an air-blown, fluidized-bed gasifier (Tampella U-Gas), a gas cooler/steam generator, and a hot gas cleanup system in combination with a gas turbine modified for use with a low-Btu content fuel and a conventional steam bottoming cycle. The demonstration plant will be located at the Toms Creek coal mine near Coeburn, Wise County, Virginia. Participants in the project are Tampella Power Corporation and Coastal Power Production Company. The plant will use 430 tons per day of locally mined bituminous coal to produce 55 MW of power from the gasification section of the project. A modern pulverized coal fired unit will be located adjacent to the Demonstration Project producing an additional 150 MW. A total 190 MW of power will be delivered to the electric grid at the completion of the project. In addition, 50,000 pounds per hour of steam will be exported to be used in the nearby coal preparation plant. Dolomite is used for in-bed gasifier sulfur capture and downs cleanup is accomplished in a fluidized-bed of regenerative zinc titanate. Particulate clean-up, before the gas turbine, will be performed by high temperature candle filters (1020[degree]F). The demonstration plant heat rate is estimated to be 8,700 Btu/kWh. The design of the project goes through mid 1995, with site construction activities commencing late in 1995 and leading to commissioning and start-up by the end of 1997. This is followed by a three year demonstration period.

Virr, M.J.

1992-01-01T23:59:59.000Z

153

Toms Creek IGCC Demonstration Project  

SciTech Connect (OSTI)

The Toms Creek Integrated Gasification Combined Cycle (IGCC) Demonstration Project was selected by DOE in September 1991 to participate in Round Four of the Clean Coal Technology Demonstration Program. The project will demonstrate a simplified IGCC process consisting of an air-blown, fluidized-bed gasifier (Tampella U-Gas), a gas cooler/steam generator, and a hot gas cleanup system in combination with a gas turbine modified for use with a low-Btu content fuel and a conventional steam bottoming cycle. The demonstration plant will be located at the Toms Creek coal mine near Coeburn, Wise County, Virginia. Participants in the project are Tampella Power Corporation and Coastal Power Production Company. The plant will use 430 tons per day of locally mined bituminous coal to produce 55 MW of power from the gasification section of the project. A modern pulverized coal fired unit will be located adjacent to the Demonstration Project producing an additional 150 MW. A total 190 MW of power will be delivered to the electric grid at the completion of the project. In addition, 50,000 pounds per hour of steam will be exported to be used in the nearby coal preparation plant. Dolomite is used for in-bed gasifier sulfur capture and downs cleanup is accomplished in a fluidized-bed of regenerative zinc titanate. Particulate clean-up, before the gas turbine, will be performed by high temperature candle filters (1020{degree}F). The demonstration plant heat rate is estimated to be 8,700 Btu/kWh. The design of the project goes through mid 1995, with site construction activities commencing late in 1995 and leading to commissioning and start-up by the end of 1997. This is followed by a three year demonstration period.

Virr, M.J.

1992-11-01T23:59:59.000Z

154

Results and implications of the EBR-II inherent safety demonstration tests  

SciTech Connect (OSTI)

On April 3, 1986 two milestone tests were conducted in Experimental Breeder Reactor-2 (EBR-II). The first test was a loss of flow without scram and the second was a loss of heat sink without scram. Both tests were initiated from 100% power and in both tests the reactor was shut down by natural processes, principally thermal expansion, without automatic scram, operator intervention or the help of special in-core devices. The temperature transients during the tests were mild, as predicted, and there was no damage to the core or reactor plant structures. In a general sense, therefore, the tests plus supporting analysis demonstrated the feasibility of inherent passive shutdown for undercooling accidents in metal-fueled LMRs. The results provide a technical basis for future experiments in EBR-II to demonstrate inherent safety for overpower accidents and provide data for validation of computer codes used for design and safety analysis of inherently safe reactor plants.

Planchon, H.P.; Golden, G.H.; Sackett, J.I.; Mohr, D.; Chang, L.K.; Feldman, E.E.; Betten, P.R.

1987-01-01T23:59:59.000Z

155

Better Buildings Residential Program Solution Center Demonstration...  

Energy Savers [EERE]

Better Buildings Residential Program Solution Center Demonstration Webinar Better Buildings Residential Program Solution Center Demonstration Webinar Demonstration webinar slides...

156

Reactor hot spot analysis  

SciTech Connect (OSTI)

The principle methods for performing reactor hot spot analysis are reviewed and examined for potential use in the Applied Physics Division. The semistatistical horizontal method is recommended for future work and is now available as an option in the SE2-ANL core thermal hydraulic code. The semistatistical horizontal method is applied to a small LMR to illustrate the calculation of cladding midwall and fuel centerline hot spot temperatures. The example includes a listing of uncertainties, estimates for their magnitudes, computation of hot spot subfactor values and calculation of two sigma temperatures. A review of the uncertainties that affect liquid metal fast reactors is also presented. It was found that hot spot subfactor magnitudes are strongly dependent on the reactor design and therefore reactor specific details must be carefully studied. 13 refs., 1 fig., 5 tabs.

Vilim, R.B.

1985-08-01T23:59:59.000Z

157

P Reactor Grouting  

SciTech Connect (OSTI)

Filling the P Reactor with grout. This seals the radioactive material and reduces the environmental footprint left from the Cold War. Project sponsored by the Recovery Act at the Savannah River Site.

None

2010-01-01T23:59:59.000Z

158

Buried Waste Integrated Demonstration Plan  

SciTech Connect (OSTI)

This document presents the plan of activities for the Buried Waste Integrated Demonstration (BWID) program which supports the environmental restoration (ER) objectives of the Department of Energy (DOE) Complex. Discussed in this plan are the objectives, organization, roles and responsibilities, and the process for implementing and managing BWID. BWID is hosted at the Idaho National Engineering Laboratory (INEL), but involves participants from throughout the DOE Complex, private industry, universities, and the international community. These participants will support, demonstrate, and evaluate a suite of advanced technologies representing a comprehensive remediation system for the effective and efficient remediation of buried waste. The processes for identifying technological needs, screening candidate technologies for applicability and maturity, selecting appropriate technologies for demonstration, field demonstrating, evaluation of results and transferring technologies to environmental restoration programs are also presented. This document further describes the elements of project planning and control that apply to BWID. It addresses the management processes, operating procedures, programmatic and technical objectives, and schedules. Key functions in support of each demonstration such as regulatory coordination, safety analyses, risk evaluations, facility requirements, and data management are presented.

Kostelnik, K.M.

1991-12-01T23:59:59.000Z

159

Nuclear reactor control  

SciTech Connect (OSTI)

A liquid metal cooled fast breeder nuclear reactor has power setback means for use in an emergency. On initiation of a trip-signal a control rod is injected into the core in two stages, firstly, by free fall to effect an immediate power-set back to a safe level and, secondly, by controlled insertion. Total shut-down of the reactor under all emergencies is avoided. 4 claims.

Ingham, R.V.

1980-01-01T23:59:59.000Z

160

Polymerization reactor control  

SciTech Connect (OSTI)

The principal difficulties in achieving good control of polymerization reactors are related to inadequate on-line measurement, a lack of understanding of the dynamics of the process, the highly sensitive and nonlinear behavior of these reactors, and the lack of well-developed techniques for the control of nonlinear processes. Some illustrations of these problems and a discussion of potential techniques for overcoming some of these difficulties is provided.

Ray, W.H.

1985-01-01T23:59:59.000Z

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


161

Molten metal reactors  

DOE Patents [OSTI]

A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

Bingham, Dennis N; Klingler, Kerry M; Turner, Terry D; Wilding, Bruce M

2013-11-05T23:59:59.000Z

162

Propane Vehicle Demonstration Grant Program  

SciTech Connect (OSTI)

Project Description: Propane Vehicle Demonstration Grants The Propane Vehicle Demonstration Grants was established to demonstrate the benefits of new propane equipment. The US Department of Energy, the Propane Education & Research Council (PERC) and the Propane Vehicle Council (PVC) partnered in this program. The project impacted ten different states, 179 vehicles, and 15 new propane fueling facilities. Based on estimates provided, this project generated a minimum of 1,441,000 new gallons of propane sold for the vehicle market annually. Additionally, two new off-road engines were brought to the market. Projects originally funded under this project were the City of Portland, Colorado, Kansas City, Impco Technologies, Jasper Engines, Maricopa County, New Jersey State, Port of Houston, Salt Lake City Newspaper, Suburban Propane, Mutual Liquid Propane and Ted Johnson.

Jack Mallinger

2004-08-27T23:59:59.000Z

163

F Reactor Inspection  

SciTech Connect (OSTI)

Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-10-29T23:59:59.000Z

164

F Reactor Inspection  

ScienceCinema (OSTI)

Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-11-24T23:59:59.000Z

165

Energy Research, Development and Demonstration  

E-Print Network [OSTI]

80-L-l i I Study of low Btu fixed-bed gasification of li~nite pellets. SPI NUMBER 80-L-2 Research and/or development of advattced technologies for the use of lignite as an industrial fU~l. SPI NUMBER 80-L-3 I Demonstration of advanced... Nelson, (July, 1979, EDF-017 Demonstration of Solar Ener Conversion of A ricultural or Industrial Wastes of Fuels, Dow Chemical Co., May, 1979, Project #B-0-2 EDF-018 Alternative Energy Sources for Agricultural Applications Including Gasification...

Ray, R. R., Jr.

1980-01-01T23:59:59.000Z

166

Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)  

SciTech Connect (OSTI)

High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next Generation Safeguards Initiative (NGSI).

Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

2009-10-01T23:59:59.000Z

167

LIMB Demonstration Project Extension and Coolside Demonstration. [Final report  

SciTech Connect (OSTI)

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.

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

1992-11-10T23:59:59.000Z

168

Leak detection capability in CANDU reactors  

SciTech Connect (OSTI)

This paper addresses the moisture leak detection capability of Ontario Hydro CANDU reactors which has been demonstrated by performing tests on the reactor. The tests confirmed the response of the annulus gas system (AGS) to the presence of moisture injected to simulate a pressure tube leak and also confirmed the dew point response assumed in leak before break assessments. The tests were performed on Bruce A Unit 4 by injecting known and controlled rates of heavy water vapor. To avoid condensation during test conditions, the amount of moisture which could be injected was small (2-3.5 g/hr). The test response demonstrated that the AGS is capable of detecting and annunciating small leaks. Thus confidence is provided that it would alarm for a growing pressure tube leak where the leak rate is expected to increase to kg/hr rapidly. The measured dew point response was close to that predicted by analysis.

Azer, N.; Barber, D.H.; Boucher, P.J. [and others

1997-04-01T23:59:59.000Z

169

International Conference on Microwave and High Frequency Heating Nottingham, UK, September 2013 Scaling Up Reactors for Microwave-Assisted  

E-Print Network [OSTI]

2013 316 Scaling Up Reactors for Microwave-Assisted Chemistry via ANN Optimization Andrew O. Holmesa by demonstrating that a particular microwave reactor can be designed with the help of computer optimization optimization technique to a microwave reactor. This technique is used here to optimize the geometry of a system

Yakovlev, Vadim

170

EA-1148: Electrometallurgical Treatment Research and Demonstration Project in the Fuel Conditioning Facility at Argonne National Laboratory- West  

Broader source: Energy.gov [DOE]

DOE prepared an EA that evaluated the potential environmental impacts associated with the research and demonstration of electrometallurgical technology for treating Experimental Breeder Reactor-II Spent Nuclear Fuel in the Fuel Conditioning Facility at Argonne National Laboratory-West.

171

REACTOR GROUT THERMAL PROPERTIES  

SciTech Connect (OSTI)

Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

2011-01-28T23:59:59.000Z

172

Multiplicity features of nonadiabatic, autothermal tubular reactors  

SciTech Connect (OSTI)

Singularity theory is combined with asymptotic analysis to determine the exact uniqueness-multiplicity boundary and ignition and extinction locus for the non-adiabatic, autothermal tubular reactor model. It is found that the steady-state behavior of the nonadiabatic reactor is described by the two limiting cases of adiabatic and strongly cooled models. The adiabatic case has been examined in a previous study. In this paper, the authors develop limiting models to describe the strongly cooled asymptotes. The authors also classify the different types of bifurcation diagrams of conversion vs. residence time using the results of singularity theory with a distinguished parameter. Analytical criteria are developed for predicting the conditions under which autothermal operation is feasible when heat losses are significant.

Lovo, M.; Balakotaiah, V. (Houston Univ., TX (United States). Dept. of Chemical Engineering)

1992-01-01T23:59:59.000Z

173

SRS reactor control rod cooling without normal forced convection cooling  

SciTech Connect (OSTI)

This paper describes an analytical study of the coolability of the control rods in the Savannah River site (SRS) K production reactor under conditions of loss of normal forced convection cooling. The study was performed as part of the overall safety analysis of the reactor supporting its restart. The analysis addresses the buoyancy-driven boiling flow over the control rods that occurs when forced cooling is lost. The objective of the study was to demonstrate that the control rods will remain cooled (i.e., no melting) at powers representative of those anticipated for restart of the reactor.

Smith, D.C. (SAIC, Albuquerque, NM (United States)); Easterling, T.C. (Westinghouse Savannah River Co., Aiken, SC (United States))

1993-01-01T23:59:59.000Z

174

Microchannel Reactor System for Catalytic Hydrogenation  

SciTech Connect (OSTI)

We successfully demonstrated a novel process intensification concept enabled by the development of microchannel reactors, for energy efficient catalytic hydrogenation reactions at moderate temperature, and pressure, and low solvent levels. We designed, fabricated, evaluated, and optimized a laboratory-scale microchannel reactor system for hydrogenation of onitroanisole and a proprietary BMS molecule. In the second phase of the program, as a prelude to full-scale commercialization, we designed and developed a fully-automated skid-mounted multichannel microreactor pilot plant system for multiphase reactions. The system is capable of processing 1 – 10 kg/h of liquid substrate, and an industrially relevant immiscible liquid-liquid was successfully demonstrated on the system. Our microreactor-based pilot plant is one-of-akind. We anticipate that this process intensification concept, if successfully demonstrated, will provide a paradigm-changing basis for replacing existing energy inefficient, cost ineffective, environmentally detrimental slurry semi-batch reactor-based manufacturing practiced in the pharmaceutical and fine chemicals industries.

Adeniyi Lawal; Woo Lee; Ron Besser; Donald Kientzler; Luke Achenie

2010-12-22T23:59:59.000Z

175

Transient thermal analysis of a space reactor power system  

E-Print Network [OSTI]

Thermoelectric Power Conversion Module Heat Pipe Radiator Module . Auxiliary Modules . Flow of Calculation . Transient Test Cases Studied Summary . 10 10 CHAPTER II. ENERGY EQUATION FINITE DIFFERENCING . . 12 Energy Equation for a Solid Finite..., but this stud~ uses a generic liquid metal cooled fast reactor concept as the model to test the code. The space power svstem to be modeled consists of a liquid lithium cooled fast reactor, primarv and secondary loops svith a sell-induced thermoelectric...

Gaeta, Michael J.

1988-01-01T23:59:59.000Z

176

Core follow calculation with the nTRACER numerical reactor and verification using power reactor measurement data  

SciTech Connect (OSTI)

The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)

Jung, Y. S.; Joo, H. G. [Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul (Korea, Republic of); Yoon, J. I. [KEPCO Nuclear Fuel, 1047 Daedukdae-ro, Yuseong-gu, Daejeon (Korea, Republic of)

2013-07-01T23:59:59.000Z

177

Methanation assembly using multiple reactors  

DOE Patents [OSTI]

A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

Jahnke, Fred C.; Parab, Sanjay C.

2007-07-24T23:59:59.000Z

178

Power Burst Facility (PBF) Reactor Reactor Decommissioning  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 - September 2006PhotovoltaicSeptember 22,Reactor Decommissioning Click here to view

179

A TEN MEGAWATT BOILING HETEROGENEOUS PACKAGE POWER REACTOR. Reactor...  

Office of Scientific and Technical Information (OSTI)

A reactor and associated power plant designed to produce 1.05 Mwh and 3.535 Mwh of steam for heating purposes are described. The total thermal output of the reactor is 10 Mwh....

180

Heat dissipating nuclear reactor  

DOE Patents [OSTI]

Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extends from the metal base plate downwardly and outwardly into the earth.

Hunsbedt, Anstein (Los Gatos, CA); Lazarus, Jonathan D. (Sunnyvale, CA)

1987-01-01T23:59:59.000Z

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


181

Inexpensive Mini Thermonuclear Reactor  

E-Print Network [OSTI]

This proposed design for a mini thermonuclear reactor uses a method based upon a series of important innovations. A cumulative explosion presses a capsule with nuclear fuel up to 100 thousands of atmospheres, the explosive electric generator heats the capsule/pellet up to 100 million degrees and a special capsule and a special cover which keeps these pressure and temperature in capsule up to 0.001 sec. which is sufficient for Lawson criteria for ignition of thermonuclear fuel. Major advantages of these reactors/bombs is its very low cost, dimension, weight and easy production, which does not require a complex industry. The mini thermonuclear bomb can be delivered as a shell by conventional gun (from 155 mm), small civil aircraft, boat or even by an individual. The same method may be used for thermonuclear engine for electric energy plants, ships, aircrafts, tracks and rockets. Key words: Thermonuclear mini bomb, thermonuclear reactor, nuclear energy, nuclear engine,

Alexander Bolonkin; Alexander Bolonkin

182

Nuclear reactor safety device  

DOE Patents [OSTI]

A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

Hutter, Ernest (Wilmette, IL)

1986-01-01T23:59:59.000Z

183

Fusion reactor control  

SciTech Connect (OSTI)

The plasma kinetic temperature and density changes, each per an injected fuel density rate increment, control the energy supplied by a thermonuclear fusion reactor in a power production cycle. This could include simultaneously coupled control objectives for plasma current, horizontal and vertical position, shape and burn control. The minimum number of measurements required, use of indirect (not plasma parameters) system measurements, and distributed control procedures for burn control are to be verifiable in a time dependent systems code. The International Thermonuclear Experimental Reactor (ITER) has the need to feedback control both the fusion output power and the driven plasma current, while avoiding damage to diverter plates. The system engineering of fusion reactors must be performed to assure their development expeditiously and effectively by considering reliability, availability, maintainability, environmental impact, health and safety, and cost.

Plummer, D.A.

1995-12-31T23:59:59.000Z

184

Reactor for exothermic reactions  

DOE Patents [OSTI]

A liquid phase process for oligomerization of C.sub.4 and C.sub.5 isoolefins or the etherification thereof with C.sub.1 to C.sub.6 alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120.degree. to 300.degree. F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

Smith, Jr., Lawrence A. (Bellaire, TX); Hearn, Dennis (Houston, TX); Jones, Jr., Edward M. (Friendswood, TX)

1993-01-01T23:59:59.000Z

185

Heat dissipating nuclear reactor  

DOE Patents [OSTI]

Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

Hunsbedt, A.; Lazarus, J.D.

1985-11-21T23:59:59.000Z

186

Reactor for exothermic reactions  

DOE Patents [OSTI]

A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

1993-03-02T23:59:59.000Z

187

EGS Demonstration | Open Energy Information  

Open Energy Info (EERE)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home5b9fcbce19 NoPublic Utilities Address:011-DNA Jump37. It is classified as ASHRAEDuval County,EEnergy ProjectDemonstration Jump

188

Fusion reactor pumped laser  

DOE Patents [OSTI]

A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam.

Jassby, Daniel L. (Princeton, NJ)

1988-01-01T23:59:59.000Z

189

Fast quench reactor method  

DOE Patents [OSTI]

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream.

Detering, Brent A. (Idaho Falls, ID); Donaldson, Alan D. (Idaho Falls, ID); Fincke, James R. (Idaho Falls, ID); Kong, Peter C. (Idaho Falls, ID); Berry, Ray A. (Idaho Falls, ID)

1999-01-01T23:59:59.000Z

190

Fast quench reactor method  

DOE Patents [OSTI]

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream. 8 figs.

Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.; Berry, R.A.

1999-08-10T23:59:59.000Z

191

Diagnostics for hybrid reactors  

SciTech Connect (OSTI)

The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

Orsitto, Francesco Paolo [ENEA Unita' Tecnica Fusione , Associazione ENEA-EURATOM sulla Fusione C R Frascati v E Fermi 45 00044 Frascati (Italy)

2012-06-19T23:59:59.000Z

192

Perspectives on reactor safety  

SciTech Connect (OSTI)

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

1994-03-01T23:59:59.000Z

193

Bentonite mat demonstration. Final report  

SciTech Connect (OSTI)

The Bentonite Mat Demonstration was developed to provide the Environmental Restoration Department with field performance characteristics and engineering data for an alternative closure cover system configuration. The demonstration was initiated in response to regulatory concerns regarding the use of an alternative cover system for future design configurations. These design considerations are in lieu of the US Environmental Protection Agency (EPA) Recommended Design for Closure Cover Systems and specifically a single compacted kaolin clay layer with a hydraulic conductivity of 1 {times} 10{sup {minus}7} cm/sec. This alternative configuration is a composite geosynthetic material hydraulic barrier consisting from bottom to top: 2 ft compacted sandy clay layer (typical local Savannah River Site soil type) that is covered by a bentonite mat--geosynthetic clay liner (GCL) and is overlaid by a 40 mil High Density Polyethylene (HDPE) geomembrane--flexible membrane liner. This effort was undertaken to obtain and document the necessary field performance/engineering data for future designs and meet regulatory technical requirements for an alternative cover system configuration. The composite geosynthetic materials hydraulic barrier is the recommended alternative cover system configuration for containment of hazardous and low level radiological waste layers that have a high potential of subsidence to be used at the Savannah River Site (SRS). This alternative configuration mitigates subsidence effects in providing a flexible, lightweight cover system to maintain the integrity of the closure. The composite geosynthetic materials hydraulic barrier is recommended for the Sanitary Landfill and Low Level Radiological Waste Disposal Facility (LLRWDF) Closures.

Serrato, M.G.

1994-12-30T23:59:59.000Z

194

Innovative design of uranium startup fast reactors  

E-Print Network [OSTI]

Sodium Fast Reactors are one of the three candidates of GEN-IV fast reactors. Fast reactors play an important role in saving uranium resources and reducing nuclear wastes. Conventional fast reactors rely on transuranic ...

Fei, Tingzhou

2012-01-01T23:59:59.000Z

195

Reactor operation environmental information document  

SciTech Connect (OSTI)

The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

1989-12-01T23:59:59.000Z

196

Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor  

SciTech Connect (OSTI)

The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.

S. Blaine Grover

2005-10-01T23:59:59.000Z

197

Reactor operation safety information document  

SciTech Connect (OSTI)

The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

Not Available

1990-01-01T23:59:59.000Z

198

Reed Reactor Facility Annual Report  

SciTech Connect (OSTI)

This is the report of the operations, experiments, modifications, and other aspects of the Reed Reactor Facility for the year.

Frantz, Stephen G.

2000-09-01T23:59:59.000Z

199

Review of tokamak power reactor and blanket designs in the United States  

SciTech Connect (OSTI)

The last major conceptual design study of a tokamak power reactor in the United States was STARFIRE which was carried out in 1979-1980. Since that time US studies have concentrated on engineering test reactors, demonstration reactors, parametric systems studies, scoping studies, and studies of selected critical issues such as pulsed vs. steady-state operation and blanket requirements. During this period, there have been many advancements in tokamak physics and reactor technology, and there has also been a recognition that it is desirable to improve the tokamak concept as a commercial power reactor candidate. During 1984-1985 several organizations participated in the Tokamak Power Systems Study (TPSS) with the objective of developing ideas for improving the tokamak as a power reactor. Also, the US completed a comprehensive Blanket Comparison and Selection Study which formed the basis for further studies on improved blankets for fusion reactors.

Baker, C.; Brooks, J.; Ehst, D.; Gohar, Y.; Smith, D.; Sze, D.

1986-01-01T23:59:59.000Z

200

UDC Demonstrates Phosphorescent OLED Systems  

Broader source: Energy.gov [DOE]

Universal Display Corporation (UDC), along with project partners Armstrong World Industries and the universities of Michigan and Southern California, have successfully demonstrated two phosphorescent OLED (PHOLED™) luminaire systems, the first of their kind in the U.S. This achievement marks a critical step in the development of practical OLED lighting in a complete luminaire system, including decorative housing, power supply, mounting, and maintenance provisions. Each luminaire has overall dimensions of approximately 15x60 cm and is comprised of four 15x15 cm phosphorescent OLED panels. With a combined power supply and lamp efficacy of 51 lm/W, the prototype luminaire is about twice as efficient as the market-leading halogen-based systems. In addition, the prototype OLED lighting system snaps into Armstrong's TechZone™ Ceiling System, which is commercially available in the U.S.x

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


201

Navy fuel cell demonstration project.  

SciTech Connect (OSTI)

This is the final report on a field evaluation by the Department of the Navy of twenty 5-kW PEM fuel cells carried out during 2004 and 2005 at five Navy sites located in New York, California, and Hawaii. The key objective of the effort was to obtain an engineering assessment of their military applications. Particular issues of interest were fuel cell cost, performance, reliability, and the readiness of commercial fuel cells for use as a standalone (grid-independent) power option. Two corollary objectives of the demonstration were to promote technological advances and to improve fuel performance and reliability. From a cost perspective, the capital cost of PEM fuel cells at this stage of their development is high compared to other power generation technologies. Sandia National Laboratories technical recommendation to the Navy is to remain involved in evaluating successive generations of this technology, particularly in locations with greater environmental extremes, and it encourages their increased use by the Navy.

Black, Billy D.; Akhil, Abbas Ali

2008-08-01T23:59:59.000Z

202

Demonstration of integrated optimization software  

SciTech Connect (OSTI)

NeuCO has designed and demonstrated the integration of five system control modules using its proprietary ProcessLink{reg_sign} technology of neural networks, advanced algorithms and fuzzy logic to maximize performance of coal-fired plants. The separate modules control cyclone combustion, sootblowing, SCR operations, performance and equipment maintenance. ProcessLink{reg_sign} provides overall plant-level integration of controls responsive to plant operator and corporate criteria. Benefits of an integrated approach include NOx reduction improvement in heat rate, availability, efficiency and reliability; extension of SCR catalyst life; and reduced consumption of ammonia. All translate into cost savings. As plant complexity increases through retrofit, repowering or other plant modifications, this integrated process optimization approach will be an important tool for plant operators. 1 fig., 1 photo.

NONE

2008-01-01T23:59:59.000Z

203

Clean Coal Diesel Demonstration Project  

SciTech Connect (OSTI)

A Clean Coal Diesel project was undertaken to demonstrate a new Clean Coal Technology that offers technical, economic and environmental advantages over conventional power generating methods. This innovative technology (developed to the prototype stage in an earlier DOE project completed in 1992) enables utilization of pre-processed clean coal fuel in large-bore, medium-speed, diesel engines. The diesel engines are conventional modern engines in many respects, except they are specially fitted with hardened parts to be compatible with the traces of abrasive ash in the coal-slurry fuel. Industrial and Municipal power generating applications in the 10 to 100 megawatt size range are the target applications. There are hundreds of such reciprocating engine power-plants operating throughout the world today on natural gas and/or heavy fuel oil.

Robert Wilson

2006-10-31T23:59:59.000Z

204

Thermal Reactor Safety  

SciTech Connect (OSTI)

Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

Not Available

1980-06-01T23:59:59.000Z

205

Nuclear reactor building  

DOE Patents [OSTI]

A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

Gou, Perng-Fei (Saratoga, CA); Townsend, Harold E. (Campbell, CA); Barbanti, Giancarlo (Sirtori, IT)

1994-01-01T23:59:59.000Z

206

Nuclear reactor building  

DOE Patents [OSTI]

A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.

Gou, P.F.; Townsend, H.E.; Barbanti, G.

1994-04-05T23:59:59.000Z

207

NETL - Chemical Looping Reactor  

ScienceCinema (OSTI)

NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

None

2014-06-26T23:59:59.000Z

208

Nuclear Reactors and Technology  

SciTech Connect (OSTI)

This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

Cason, D.L.; Hicks, S.C. [eds.

1992-01-01T23:59:59.000Z

209

Fossil fuel furnace reactor  

DOE Patents [OSTI]

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

Parkinson, William J. (Los Alamos, NM)

1987-01-01T23:59:59.000Z

210

Performance Demonstration Program Management Plan  

SciTech Connect (OSTI)

To demonstrate compliance with the Waste Isolation Pilot Plant (WIPP) waste characterization program, each testing and analytical facility performing waste characterization activities participates in the Performance Demonstration Program (PDP). The PDP serves as a quality control check against expected results and provides information about the quality of data generated in the characterization of waste destined for WIPP. Single blind audit samples are prepared and distributed by an independent organization to each of the facilities participating in the PDP. There are three elements within the PDP: analysis of simulated headspace gases, analysis of solids for Resource Conservation and Recovery Act (RCRA) constituents, and analysis for transuranic (TRU) radionuclides using nondestructive assay (NDA) techniques. Because the analysis for TRU radionuclides using NDA techniques involves both the counting of drums and standard waste boxes, four PDP plans are required to describe the activities of the three PDP elements. In accordance with these PDP plans, the reviewing and approving authority for PDP results and for the overall program is the CBFO PDP Appointee. The CBFO PDP Appointee is responsible for ensuring the implementation of each of these plans by concurring with the designation of the Program Coordinator and by providing technical oversight and coordination for the program. The Program Coordinator will designate the PDP Manager, who will coordinate the three elements of the PDP. The purpose of this management plan is to identify how the requirements applicable to the PDP are implemented during the management and coordination of PDP activities. The other participants in the program (organizations that perform site implementation and activities under CBFO contracts or interoffice work orders) are not covered under this management plan. Those activities are governed by the organization’s quality assurance (QA) program and procedures or as otherwise directed by CBFO.

Carlsbad Field Office

2005-07-01T23:59:59.000Z

211

Controlled Hydrogen Fleet and Infrastructure Demonstration and...  

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

Controlled Hydrogen Fleet and Infrastructure Demonstration and Validation Project Solicitation Controlled Hydrogen Fleet and Infrastructure Demonstration and Validation Project...

212

Industrial Scale Demonstration of Smart Manufacturing Achieving...  

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

Scale Demonstration of Smart Manufacturing Achieving Transformational Energy Productivity Gains Industrial Scale Demonstration of Smart Manufacturing Achieving...

213

Webinar: Hydrogen Storage Materials Database Demonstration |...  

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

Storage Materials Database Demonstration Webinar: Hydrogen Storage Materials Database Demonstration Presentation slides from the Fuel Cell Technologies Office webinar "Hydrogen...

214

Electric Drive Vehicle Demonstration and Vehicle Infrastructure...  

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

Electric Drive Vehicle Demonstration and Vehicle Infrastructure Evaluation Electric Drive Vehicle Demonstration and Vehicle Infrastructure Evaluation 2010 DOE Vehicle Technologies...

215

Honeywell Demonstrates Automated Demand Response Benefits for...  

Office of Environmental Management (EM)

Honeywell Demonstrates Automated Demand Response Benefits for Utility, Commercial, and Industrial Customers Honeywell Demonstrates Automated Demand Response Benefits for Utility,...

216

Grid Connectivity Research, Development & Demonstration Projects...  

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

Connectivity Research, Development & Demonstration Projects Grid Connectivity Research, Development & Demonstration Projects 2013 DOE Hydrogen and Fuel Cells Program and Vehicle...

217

Better Buildings Residential Program Solution Center Demonstration...  

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

Program Solution Center Demonstration Webinar Transcript Better Buildings Residential Program Solution Center Demonstration Webinar Transcript The Better Buildings Residential...

218

Better Buildings Residential Program Solution Center Demonstration...  

Energy Savers [EERE]

Better Buildings Residential Program Solution Center Demonstration Better Buildings Residential Program Solution Center Demonstration Better Buildings Residential Program Solution...

219

Distributed Energy Technology Simulator: Microturbine Demonstration...  

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

Simulator: Microturbine Demonstration, October 2001 Distributed Energy Technology Simulator: Microturbine Demonstration, October 2001 This 2001 paper discusses the National Rural...

220

Reduced-Enrichment Research and Test Reactor Program: Environmental assessment  

SciTech Connect (OSTI)

The principal program objective and principal part of the proposed action is to improve the proliferation resistance of nuclear fuels used in research and test reactors by providing the technical means (through technical development, design, and testing) for reducing the uranium enrichment requirements of these fuels to substantially less than the 90 to 93% enrichment currently used. Operator acceptance of the reduced-enrichment-uranium (REU) fuel alternative will require minimizing of reactor performance reduction, fuel cycle cost increases, the number of new safety and licensing issues raised, and reactor and facility modifications. The other part of the proposed action is to assure the capability for commercial production and supply of REU fuel for use both in the US and abroad. The RERTR Program scope is limited to generic design studies, technical support to reactor operating organizations in preparing for conversions to REU fuels, fuel development, fuel demonstrations, and technical support for commercialization of REU fuels. This environmental assessment addresses the environmental consequences of RERTR Program activities and of specific conversions of typical reactors (the Ford Nuclear Reactor and one or two other to-be-designated demonstrations) to REU-fuel cycles, including domestic and international shipments of enriched uranium pertinent to the conduct of RERTR Program activities.

Not Available

1980-05-01T23:59:59.000Z

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


221

Reactor safety research programs. Quarterly report Apr-Jun 81  

SciTech Connect (OSTI)

This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 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 pipe-to-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-of-coolant 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.

Edler, S.K.

1981-09-01T23:59:59.000Z

222

Reactor Safety Research Programs Quarterly Report April- June 1981  

SciTech Connect (OSTI)

This document summarizes the work performed by Pacific Northwest laboratory (PNL} from April1 through June 30, 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 pipe-to-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-of-coolant 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, lspra, 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.

Edler, S. K.

1981-09-01T23:59:59.000Z

223

Direct conversion nuclear reactor space power systems  

SciTech Connect (OSTI)

This paper presents the results of a study of space nuclear reactor power systems using either thermoelectric or thermionic energy converters. An in-core reactor design and two heat pipe cooled out-of-core reactor designs were considered. One of the out-of-core cases utilized, long heat pipes (LHP) directly coupled to the energy converter. The second utilized a larger number of smaller heat pipes (mini-pipe) radiatively coupled to the energy converter. In all cases the entire system, including power conditioning, was constrained to be launched in a single shuttle flight. Assuming presently available performance, both the LHP thermoelectric system and minipipe thermionic system, designed to produce 100 kWe for seven years, would have a specific mass near 22kg/kWe. The specific mass of the thermionic minipipe system designed for a one year mission is 165 kg/kWe due to less fuel swelling. Shuttle imposed growth limits are near 300 kWe and 1.2 MWe for the thermoelectric and thermionic systems, respectively. Converter performance improvements could double this potential, and over 10 MWe may be possible for very short missions.

Britt, E.J.; Fitzpatrick, G.O.

1982-08-01T23:59:59.000Z

224

Reactor vessel support system. [LMFBR  

DOE Patents [OSTI]

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.

Golden, M.P.; Holley, J.C.

1980-05-09T23:59:59.000Z

225

Design study of the deep-sea reactor X  

SciTech Connect (OSTI)

The deep-sea reactor X (DRX) is a small nuclear plant designed to provide undersea power sources. It has the full advantages of nuclear reactors and can provide large power capacity and does not require oxygen for power production. An application conceivable in the near future is that for a submersible. The Japan Atomic Energy Research Institute is conducting a design study of a 150-kW(electric) DRX plant for a deep-sea research vessel. It has a so-called integrated pressurized water reactor,'' having a steam generator inside the reactor vessel. A pressure shell includes a turbine and a generator as well as a reactor vessel, resulting in a very compact electricity producing plant. It should be easy to operate and have high passive safety characteristics; namely, a short startup time, good reactor response to power demand changes, and passive core flooding and decay heat removal in case of an accident. Transient analyses including those for load follow-up, reactor startup, and accidents have been conducted. The results show that the DRX has excellent inherent characteristics satisfying those requirements.

Iida, Hiromasa (Japan Atomic Energy Research Inst., Ibaraki (Japan)); Ishizaka, Yuichi (Mitsubishi Atomic Power Industries, Inc., Tokyo (Japan)); Kim, Y.C.; Yamaguchi, Chouichi (Japan Research Inst., Ltd., Tokyo (Japan))

1994-07-01T23:59:59.000Z

226

Neutronic analysis of a fusion hybrid reactor  

SciTech Connect (OSTI)

In a PHYSOR 2010 paper(1) we introduced a fusion hybrid reactor whose fusion component is the gasdynamic mirror (GDM), and whose blanket was made of thorium - 232. The thrust of that study was to demonstrate the performance of such a reactor by establishing the breeding of uranium - 233 in the blanket, and the burning thereof to produce power. In that analysis, we utilized the diffusion equation for one-energy neutron group, namely, those produced by the fusion reactions, to establish the power distribution and density in the system. Those results should be viewed as a first approximation since the high energy neutrons are not effective in inducing fission, but contribute primarily to the production of actinides. In the presence of a coolant, however, such as water, these neutrons tend to thermalize rather quickly, hence a better assessment of the reactor performance would require at least a two group analysis, namely the fast and thermal groups. We follow that approach and write an approximate set of equations for the fluxes of these groups. From these relations we deduce the all-important quantity, k{sub eff}, which we utilize to compute the multiplication factor, and subsequently, the power density in the reactor. We show that k{sub eff} can be made to have a value of 0.99, thus indicating that 100 thermal neutrons are generated per fusion neutron, while allowing the system to function as 'subcritical.' Moreover, we show that such a hybrid reactor can generate hundreds of megawatts of thermal power per cm of length depending on the flux of the fusion neutrons impinging on the blanket. (authors)

Kammash, T. [Univ. of Michigan, NERS, 2355 Bonisteel Blvd., Ann Arbor, MI 48109 (United States)

2012-07-01T23:59:59.000Z

227

Technology Options for a Fast Spectrum Test Reactor  

SciTech Connect (OSTI)

Idaho National Laboratory in collaboration with Argonne National Laboratory has evaluated technology options for a new fast spectrum reactor to meet the fast-spectrum irradiation requirements for the USDOE Generation IV (Gen IV) and Advanced Fuel Cycle Initiative (AFCI) programs. The US currently has no capability for irradiation testing of large volumes of fuels or materials in a fast-spectrum reactor required to support the development of Gen IV fast reactor systems or to demonstrate actinide burning, a key element of the AFCI program. The technologies evaluated and the process used to select options for a fast irradiation test reactor (FITR) for further evaluation to support these programmatic objectives are outlined in this paper.

D. M. Wachs; R. W. King; I. Y. Glagolenko; Y. Shatilla

2006-06-01T23:59:59.000Z

228

Nuclear reactor construction with bottom supported reactor vessel  

DOE Patents [OSTI]

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

Sharbaugh, John E. (Bullskin Township, Fayette County, PA)

1987-01-01T23:59:59.000Z

229

Foundational development of an advanced nuclear reactor integrated safety code.  

SciTech Connect (OSTI)

This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

2010-02-01T23:59:59.000Z

230

Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor  

SciTech Connect (OSTI)

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

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

2009-10-01T23:59:59.000Z

231

Nuclear divisional reactor  

SciTech Connect (OSTI)

A nuclear divisional reactor including a reactor core having side and top walls, a heat exchanger substantially surrounding the core, the heat exchanger including a plurality of separate fluid holding and circulating chambers each in contact with a portion of the core, control rod means associated with the core and external of the heat exchanger including control rods and means for moving said control rods, each of the chambers having separate means for delivering and removing fluid therefrom, separate means associated with each of the delivering and removing means for producing useable energy external of the chambers, each of the means for producing useable energy having separate variable capacity energy outputs thereby making available a plurality of individual sources of useable energy of varying degrees.

Administratrix, A.P.; Rugh, J.L.

1982-11-02T23:59:59.000Z

232

Nuclear reactor safety device  

DOE Patents [OSTI]

A safety device is described for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of a thermal excursion. It comprises a laminated strip helically configured to form a tube, said tube being in operative relation to said control rod. The laminated strip is formed of at least two materials having different thermal coefficients of expansion, and is helically configured such that the material forming the outer lamina of the tube has a greater thermal coefficient of expansion than the material forming the inner lamina of said tube. In the event of a thermal excursion the laminated strip will tend to curl inwardly so that said tube will increase in length, whereby as said tube increases in length it exerts a force on said control rod to axially reposition said control rod with respect to said core.

Hutter, E.

1983-08-15T23:59:59.000Z

233

Fusion reactor pumped laser  

DOE Patents [OSTI]

A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam. 10 figs.

Jassby, D.L.

1987-09-04T23:59:59.000Z

234

Guidelines for the Use of Function Block Diagram in Reactor Protection Systems  

E-Print Network [OSTI]

) in reactor protection system (RPS) developed in the Korea Nuclear Instrumentation and Control System RGuidelines for the Use of Function Block Diagram in Reactor Protection Systems Dong-Ah Lee, Junbeom for programmable logic controllers (PLC) widely used in indus- try. The guidelines show that what cases cause

235

Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two  

SciTech Connect (OSTI)

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future.

Glasstone, S.; Sesonske, A.

1994-12-31T23:59:59.000Z

236

Basis for NGNP Reactor Design Down-Selection  

SciTech Connect (OSTI)

The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.

L.E. Demick

2010-08-01T23:59:59.000Z

237

Basis for NGNP Reactor Design Down-Selection  

SciTech Connect (OSTI)

The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.

L.E. Demick

2011-11-01T23:59:59.000Z

238

The development and demonstration of a thermal neutron radiography facility utilizing the TAMU NSC TRIGA reactor  

E-Print Network [OSTI]

an object without permanently changing or destroying the object being inspected. X ? rays and gamma rays are probably the most widely used radiations in the above applications; however, the use of beams of neutrons is becoming more prevalent. Radiography... useful results, the attenuation of radiation per unit thickness of material must depend upon the physical or atomic characteristics of the material. X-rays and gamma rays are absorbed within the objects by various interactions with the elec- tronic...

Lorenz, Robert Wayne

1972-01-01T23:59:59.000Z

239

Louisiana: Verenium Cellulosic Ethanol Demonstration Facility...  

Energy Savers [EERE]

Louisiana: Verenium Cellulosic Ethanol Demonstration Facility Louisiana: Verenium Cellulosic Ethanol Demonstration Facility April 9, 2013 - 12:00am Addthis In 2010, Verenium...

240

Hydrogen Storage Materials Database Demonstration Webinar (Text...  

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

Database Demonstration Webinar (Text Version) Hydrogen Storage Materials Database Demonstration Webinar (Text Version) Below is the text version of the webinar titled "Hydrogen...

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


241

GROWDERS Demonstration of Grid Connected Electricity Systems...  

Open Energy Info (EERE)

GROWDERS Demonstration of Grid Connected Electricity Systems (Smart Grid Project) (Spain) Jump to: navigation, search Project Name GROWDERS Demonstration of Grid Connected...

242

Technical Demonstration and Economic Validation of Geothermal...  

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

Technical Demonstration and Economic Validation of Geothermal-Produced Electricity from Coproduced Water at Existing OilGas Wells in Texas Technical Demonstration and Economic...

243

Daemen Alternative Energy/Geothermal Technologies Demonstration...  

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

Daemen Alternative EnergyGeothermal Technologies Demonstration Program Erie County Daemen Alternative EnergyGeothermal Technologies Demonstration Program Erie County Project...

244

Thanksgiving Goodwill: West Valley Demonstration Project Food...  

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

Goodwill: West Valley Demonstration Project Food Drive Provides 640 Turkeys to People in Need Thanksgiving Goodwill: West Valley Demonstration Project Food Drive Provides...

245

Hydrogen Vehicle and Infrastructure Demonstration and Validation...  

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

Vehicle and Infrastructure Demonstration and Validation Hydrogen Vehicle and Infrastructure Demonstration and Validation 2009 DOE Hydrogen Program and Vehicle Technologies Program...

246

Joint Capability Technology Demonstration (JCTD) Industry Day...  

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

Joint Capability Technology Demonstration (JCTD) Industry Day Agenda Joint Capability Technology Demonstration (JCTD) Industry Day Agenda Agenda outlines the activities of the 2014...

247

Designing Reactors to Facilitate Decommissioning  

SciTech Connect (OSTI)

Critics of nuclear power often cite issues with tail-end-of-the-fuel-cycle activities as reasons to oppose the building of new reactors. In fact, waste disposal and the decommissioning of large nuclear reactors have proven more challenging than anticipated. In the early days of the nuclear power industry the design and operation of various reactor systems was given a great deal of attention. Little effort, however, was expended on end-of-the-cycle activities, such as decommissioning and disposal of wastes. As early power and test reactors have been decommissioned difficulties with end-of-the-fuel-cycle activities have become evident. Even the small test reactors common at the INEEL were not designed to facilitate their eventual decontamination, decommissioning, and dismantlement. The results are that decommissioning of these facilities is expensive, time consuming, relatively hazardous, and generates large volumes of waste. This situation clearly supports critics concerns about building a new generation of power reactors.

Richard H. Meservey

2006-06-01T23:59:59.000Z

248

Post-doc: Modelling & Control of Continuous Reactors  

E-Print Network [OSTI]

development in mechatronics and microsystems, sustainable industrial processes, transportation systems scope is a demonstration of continuous reactors with in-line analytics for fine chemical production for improvement of operation and actuation and will provide data for sensor development and control guidelines. 4

Langendoen, Koen

249

Integral Fast Reactor Program annual progress report, FY 1991  

SciTech Connect (OSTI)

This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1991. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R&D.

Not Available

1992-06-01T23:59:59.000Z

250

Integral Fast Reactor Program annual progress report, FY 1991  

SciTech Connect (OSTI)

This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1991. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R D.

Not Available

1992-06-01T23:59:59.000Z

251

Integral Fast Reactor Program annual progress report, FY 1994  

SciTech Connect (OSTI)

This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1994. Technical accomplishments are presented in the following areas of the IFR technology development activities: metal fuel performance; pyroprocess development; safety experiments and analyses; core design development; fuel cycle demonstration; and LMR technology R&D.

Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, J.J.

1994-12-01T23:59:59.000Z

252

Integral Fast Reactor Program. Annual progress report, FY 1992  

SciTech Connect (OSTI)

This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1992. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R&D.

Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

1993-06-01T23:59:59.000Z

253

Multilevel transport solution of LWR reactor cores  

SciTech Connect (OSTI)

This work presents a multilevel approach for the solution of the transport equation in typical LWR assemblies and core configurations. It is based on the second-order, even-parity formulation of the transport equation, which is solved within the framework provided by the finite element-spherical harmonics code EVENT. The performance of the new solver has been compared with that of the standard conjugate gradient solver for diffusion and transport problems on structured and unstruc-tured grids. Numerical results demonstrate the potential of the multilevel scheme for realistic reactor calculations.

Jose Ignacio Marquez Damian; Cassiano R.E. de Oliveira; HyeonKae Park

2008-09-01T23:59:59.000Z

254

Prompt Neutron Lifetime for the NBSR Reactor  

SciTech Connect (OSTI)

In preparation for the proposed conversion of the National Institute of Standards and Technology (NIST) research reactor (NBSR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, certain point kinetics parameters must be calculated. We report here values of the prompt neutron lifetime that have been calculated using three independent methods. All three sets of calculations demonstrate that the prompt neutron lifetime is shorter for the LEU fuel when compared to the HEU fuel and longer for the equilibrium end-of-cycle (EOC) condition when compared to the equilibrium startup (SU) condition for both the HEU and LEU fuels.

Hanson, A.L.; Diamond, D.

2012-06-24T23:59:59.000Z

255

Demonstrating the Safety of Long-Term Dry Storage - 13468  

SciTech Connect (OSTI)

Commercial nuclear plants in the United States were originally designed with the expectation that used nuclear fuel would be moved directly from the reactor pools and transported off site for either reprocessing or direct geologic disposal. However, Federal programs intended to meet this expectation were never able to develop the capability to remove used fuel from reactor sites - and these programs remain stalled to this day. Therefore, in the 1980's, with reactor pools reaching capacity limits, industry began developing dry cask storage technology to provide for additional on-site storage. Use of this technology has expanded significantly since then, and has today become a standard part of plant operations at most US nuclear sites. As this expansion was underway, Federal programs remained stalled, and it became evident that dry cask systems would be in use longer than originally envisioned. In response to this challenge, a strong technical basis supporting the long term dry storage safety has been developed. However, this is not a static situation. The technical basis must be able to address future challenges. Industry is responding to one such challenge - the increasing prevalence of high burnup (HBU) used fuel and the need to provide long term storage assurance for these fuels equivalent to that which has existed for lower burnup fuels over the past 25 years. This response includes a confirmatory demonstration program designed to address the aging characteristics of HBU fuel and set a precedent for a learning approach to aging management that will have broad applicability across the used fuel storage landscape. (authors)

McCullum, Rod [Nuclear Energy Institute, 1201 F St. NW, Washington, DC, 20004 (United States)] [Nuclear Energy Institute, 1201 F St. NW, Washington, DC, 20004 (United States); Brookmire, Tom [Dominion Energy, 5000 Dominion Boulevard Glen Allen, VA 23060 (United States)] [Dominion Energy, 5000 Dominion Boulevard Glen Allen, VA 23060 (United States); Kessler, John [Electric Power Research Institute, 1300 West W.T. Harris Boulevard, Charlotte, NC 28262 (United States)] [Electric Power Research Institute, 1300 West W.T. Harris Boulevard, Charlotte, NC 28262 (United States); Leblang, Suzanne [Entergy, 1340 Echelon Parkway, Jackson, MS 39211 (United States)] [Entergy, 1340 Echelon Parkway, Jackson, MS 39211 (United States); Levin, Adam [Exelon, 4300 Winfield Road, Warrenville, IL 60555 (United States)] [Exelon, 4300 Winfield Road, Warrenville, IL 60555 (United States); Martin, Zita [Tennessee Valley Authority, 1101 Market Street, Chattanooga, TN 37402 (United States)] [Tennessee Valley Authority, 1101 Market Street, Chattanooga, TN 37402 (United States); Nesbit, Steve [Duke Energy, 550 South Tryon Street, Charlotte, NC 28202 (United States)] [Duke Energy, 550 South Tryon Street, Charlotte, NC 28202 (United States); Nichol, Marc [Nuclear Energy Institute, 1201 F St. NW Washington DC, 2004 (United States)] [Nuclear Energy Institute, 1201 F St. NW Washington DC, 2004 (United States); Pickens, Terry [Xcel Energy, 414 Nicollet Mall, Minneapolis, MN 55401 (United States)] [Xcel Energy, 414 Nicollet Mall, Minneapolis, MN 55401 (United States)

2013-07-01T23:59:59.000Z

256

Progress Update: Reactor Disassembly Grouting  

SciTech Connect (OSTI)

Grouting the P&R reactors in order to remove these basins as an environmental threat. This will end the Cold War legacy and end the environmental footprint.

Cody, Tom

2010-01-01T23:59:59.000Z

257

Neutrino Oscillation Studies with Reactors  

E-Print Network [OSTI]

Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

Petr Vogel; Liangjian Wen; Chao Zhang

2015-03-03T23:59:59.000Z

258

Neutrino Oscillation Studies with Reactors  

E-Print Network [OSTI]

Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

Vogel, Petr; Zhang, Chao

2015-01-01T23:59:59.000Z

259

Thermonuclear Reflect AB-Reactor  

E-Print Network [OSTI]

The author offers a new kind of thermonuclear reflect reactor. The remarkable feature of this new reactor is a three net AB reflector, which confines the high temperature plasma. The plasma loses part of its energy when it contacts with the net but this loss can be compensated by an additional permanent plasma heating. When the plasma is rarefied (has a small density), the heat flow to the AB reflector is not large and the temperature in the triple reflector net is lower than 2000 - 3000 K. This offered AB-reactor has significantly less power then the currently contemplated power reactors with magnetic or inertial confinement (hundreds-thousands of kW, not millions of kW). But it is enough for many vehicles and ships and particularly valuable for tunnelers, subs and space apparatus, where air to burn chemical fuel is at a premium or simply not available. The author has made a number of innovations in this reactor, researched its theory, developed methods of computation, made a sample computation of typical project. The main point of preference for the offered reactor is its likely cheapness as a power source. Key words: Micro-thermonuclear reactor, Multi-reflex AB-thermonuclear reactor, Self-magnetic AB-thermonuclear reactor, aerospace thermonuclear engine.

Alexander Bolonkin

2008-03-26T23:59:59.000Z

260

Light Water Reactor Sustainability Newsletter  

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

hydraulics software RELAP-7 (which is under development in the Light Water Reactor Sustainability LWRS Program). A novel interaction between the probabilistic part (i.e., RAVEN)...

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


261

Light Water Reactor Sustainability Newsletter  

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

30-35, August 2012. Clayton, D. A. and M. S. Hileman, 2012, Light Water Reactor Sustainability Non-Destructive Evaluation for Concrete Research and Development Roadmap, ORNLTM-...

262

Progress Update: Reactor Disassembly Grouting  

ScienceCinema (OSTI)

Grouting the P&R reactors in order to remove these basins as an environmental threat. This will end the Cold War legacy and end the environmental footprint.

Cody, Tom

2012-06-14T23:59:59.000Z

263

TREAT (Transient Reactor Test Facility) reactor control rod scram system simulations and testing  

SciTech Connect (OSTI)

Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs.

Solbrig, C.W.; Stevens, W.W.

1990-01-01T23:59:59.000Z

264

Zircaloy performance in light water reactors  

SciTech Connect (OSTI)

Zircaloy has been successfully used as the primary light water reactor (LWR) core structural material since its introduction in the early days of the US naval nuclear program. Its unique combination of low neutron absorption cross section, fabricability, mechanical strength, and corrosion resistance in water and steam near 300{degrees}C has resulted in remarkable reliability of operation of pressurized and boiling water reactor (PWR, BWR) fuel through the years. At present time, BWRs use Zircaloy-2 and PWRs use Zircaloy-4 for fuel cladding. In BWRs, both Zircaloy-2 and -4 have been successfully used for spacer grids and channels, and in PWRs Zircaloy-4 is used for spacer grids and control rod guide tubes. Performance of fuel rods has been excellent thus far. The current trend for utilities worldwide is to expect both higher fuel reliability in the future. Fuel suppliers have already achieved extended exposures in lead use assemblies, and have demonstrated excellent performance in all areas; therefore unsuspected problems are not likely to arise. However, as exposure and expectations continue to increase, Zircaloy is being taken toward the limits of its known capabilities. This paper reviews Zircaloy performance capabilities in areas related to environmentally affected microstructure, mechanical properties, corrosion resistance, and dimensional stability. The effects of radiation and reactor environment on each property is illustrated with data, micrographs, and analysis.

Adamson, R.B.; Cheng, B.C.; Kruger, R.M. [GE Nuclear Energy, Pleasanton, CA (United States)

1992-12-31T23:59:59.000Z

265

Dynamic Response Testing in an Electrically Heated Reactor Test Facility  

SciTech Connect (OSTI)

Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG testing will utilize a higher fidelity point kinetics model to control core power transients, and reactivity feedback will be based on localized feedback coefficients and several independent temperature measurements taken within the core block. This paper presents preliminary test results and discusses the methodology that will be implemented in follow-on DDG testing and the additional instrumentation required to implement high fidelity dynamic testing.

Bragg-Sitton, Shannon M. [NASA Marshall Space Flight Center, Nuclear and Advanced Propulsion Branch, ER-11, MSFC, AL 35812 (United States); Morton, T. J. [Department of Chemical and Nuclear Engineering, University of New Mexico, Albuquerque, NM 87131 (United States)

2006-01-20T23:59:59.000Z

266

Experimental Results from an Antineutrino Detector for Cooperative Monitoring of Nuclear Reactors  

SciTech Connect (OSTI)

Our collaboration has designed, installed, and operated a compact antineutrino detector at a nuclear power station, for the purpose of monitoring the power and plutonium content of the reactor core. This paper focuses on the basic properties and performance of the detector. We describe the site, the reactor source, and the detector, and provide data that clearly show the expected antineutrino signal. Our data and experience demonstrate that it is possible to operate a simple, relatively small, antineutrino detector near a reactor, in a non-intrusive and unattended mode for months to years at a time, from outside the reactor containment, with no disruption of day-to-day operations at the reactor site. This unique real-time cooperative monitoring capability may be of interest for the International Atomic Energy Agency (IAEA) reactor safeguards program and similar regimes.

Bowden, N S; Bernstein, A; Allen, M; Brennan, J S; Cunningham, M; Estrada, J K; Greaves, C R; Hagmann, C; Lund, J; Mengesha, W; Weinbeck, T D; Winant, C D

2006-09-18T23:59:59.000Z

267

Advanced Reactors Thermal Energy Transport for Process Industries  

SciTech Connect (OSTI)

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

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

2014-07-01T23:59:59.000Z

268

Reactor Safety Research Programs Quarterly Report January - March 1980  

SciTech Connect (OSTI)

This document summarizes the work performed by Pacific Northwest Laboratory from January 1 through March 31, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where serviceinduced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

Hagen, C. M

1980-10-01T23:59:59.000Z

269

Reactor Safety Research Programs Quarterly Report July- September 1980  

SciTech Connect (OSTI)

This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission {NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; blowdown and reflood tests in the test facility at Cadarache, France; 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.

Edler, S. K.

1980-12-01T23:59:59.000Z

270

Reactor Safety Research Programs Quarterly Report April -June 1980  

SciTech Connect (OSTI)

This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission {NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; blowdown and reflood tests in the test facility at Cadarache, France; 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.

Edler, S. K.

1980-11-01T23:59:59.000Z

271

Reactor Safety Research Programs Quarterly Report July - September 1981  

SciTech Connect (OSTI)

This document summarizes the work performed by Pacific Northwest laboratory (PNL) from July 1 through September 30, 1981, for the Division of Accident Evaluation, 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 pipe-to-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-of-coolant 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, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

Edler, S. K.

1982-01-01T23:59:59.000Z

272

Reactor Safety Research Programs Quarterly Report October - December 1980  

SciTech Connect (OSTI)

This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from October 1 through December 31, 1980, 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 structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NOE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor 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.

Edler, S K

1981-04-01T23:59:59.000Z

273

Reactor Safety Research Programs Quarterly Report October - December 1981  

SciTech Connect (OSTI)

This document summarizes the work performed by Pacific Northwest laboratory (PNL) from October 1 through December 31, 1981, for the Division of Accident Evaluation, 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 serviceinduced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-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-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and post accident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

Edler, S. K.

1982-03-01T23:59:59.000Z

274

Reactor coolant pump flywheel  

DOE Patents [OSTI]

A flywheel for a pump, and in particular a flywheel having a number of high density segments for use in a nuclear reactor coolant pump. The flywheel includes an inner member and an outer member. A number of high density segments are provided between the inner and outer members. The high density segments may be formed from a tungsten based alloy. A preselected gap is provided between each of the number of high density segments. The gap accommodates thermal expansion of each of the number of segments and resists the hoop stress effect/keystoning of the segments.

Finegan, John Raymond; Kreke, Francis Joseph; Casamassa, John Joseph

2013-11-26T23:59:59.000Z

275

Reactor refueling containment system  

DOE Patents [OSTI]

A method of refueling a nuclear reactor is disclosed whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced. 2 figs.

Gillett, J.E.; Meuschke, R.E.

1995-05-02T23:59:59.000Z

276

Reactor refueling containment system  

DOE Patents [OSTI]

A method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.

Gillett, James E. (Greensburg, PA); Meuschke, Robert E. (Pittsburgh, PA)

1995-01-01T23:59:59.000Z

277

Nuclear reactor control assembly  

SciTech Connect (OSTI)

This patent describes an assembly for providing global power control in a nuclear reactor having the core split into two halves. It comprises a disk assembly formed from at least two disks each machined with an identical surface hole pattern such that rotation of one disk relative to the other causes the hole pattern to open or close, the disk assembly being positioned substantially at the longitudinal center of and coaxial with the core halves; and means for rotating at least one of the disks relative to the other.

Negron, S.B.

1991-06-11T23:59:59.000Z

278

Nuclear reactor control apparatus  

SciTech Connect (OSTI)

Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additonal magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

Sridhar, B.N.

1981-08-28T23:59:59.000Z

279

Biparticle fluidized bed reactor  

DOE Patents [OSTI]

A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase is described. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figures.

Scott, C.D.

1993-12-14T23:59:59.000Z

280

Biparticle fluidized bed reactor  

DOE Patents [OSTI]

A fluidized bed reactor system utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figs.

Scott, C.D.; Marasco, J.A.

1995-04-25T23:59:59.000Z

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


281

Nuclear reactor control apparatus  

DOE Patents [OSTI]

Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additional magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

Sridhar, Bettadapur N. (Cupertino, CA)

1983-11-01T23:59:59.000Z

282

ABPDU - Advanced Biofuels Process Demonstration Unit  

SciTech Connect (OSTI)

Lawrence Berkeley National Lab opened its Advanced Biofuels Process Demonstration Unit on Aug. 18, 2011.

None

2011-01-01T23:59:59.000Z

283

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

SciTech Connect (OSTI)

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

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

1993-06-01T23:59:59.000Z

284

Hybrid Plasma Reactor/Filter for Transportable Collective Protection Systems  

SciTech Connect (OSTI)

Pacific Northwest National Laboratory (PNNL) has performed an assessment of a Hybrid Plasma/Filter system as an alternative to conventional methods for collective protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader envelope of protection than can be provided through a single-solution approach. The first step uses highly reactive species (e.g. oxygen radicals, hydroxyl radicals, etc.) created in a nonthermal plasma (NTP) reactor to destroy the majority (~75% - 90%) of an incoming threat. Following the NTP reactor an O3 reactor/filter uses the O3 created in the NTP reactor to further destroy the remaining organic materials. This report summarizes the laboratory development of the Hybrid Plasma Reactor/Filter to protect against a ‘worst-case’ simulant, methyl bromide (CH3Br), and presents a preliminary engineering assessment of the technology to Joint Expeditionary Collective Protection performance specifications for chemical vapor air purification technologies.

Josephson, Gary B.; Tonkyn, Russell G.; Frye, J. G.; Riley, Brian J.; Rappe, Kenneth G.

2011-04-06T23:59:59.000Z

285

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect (OSTI)

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

Not Available

1991-04-01T23:59:59.000Z

286

Modular Pebble Bed Reactor High Temperature Gas Reactor  

E-Print Network [OSTI]

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

287

Transmutation of Nuclear Waste and the future MYRRHA Demonstrator  

E-Print Network [OSTI]

While a considerable and world-wide growth of the nuclear share in the global energy mix is desirable for many reasons, there are also, in particular in the "old world" major objections. These are both concerns about safety, in particular in the wake of the Fukushima nuclear accident and concerns about the long-term burden that is constituted by the radiotoxic waste from the spent fuel. With regard to the second topic, the present contribution will outline the concept of Partitioning & Transmutation (P&T), as scientific and technological answer. Deployment of P&T may use dedicated "Transmuter" or "Burner" reactors, using a fast neutron spectrum. For the transmutation of waste with a large content (up to 50%) of (very long-lived) Minor Actinides, a sub-critical reactor, using an external neutron source is a most attractive solution. It is constituted by coupling a proton accelerator, a spallation target and a subcritical core. This promising new technology is named ADS, for accelerator-driven system. The present paper aims at a short introduction into the field that has been characterized by a high collaborative activity during the last decade in Europe, in order to focus, in its later part, on the MYRRHA project as the European ADS technology demonstrator.

Alex C. Mueller

2012-10-16T23:59:59.000Z

288

INITIAL IRRADIATION OF THE FIRST ADVANCED GAS REACTOR FUEL DEVELOPMENT AND QUALIFICATION EXPERIMENT IN THE ADVANCED TEST REACTOR  

SciTech Connect (OSTI)

The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

S. Blaine Grover; David A. Petti

2007-09-01T23:59:59.000Z

289

Gaseous reactor control system  

SciTech Connect (OSTI)

This paper describes a nuclear reactor control system for controlling the reactivity of the core of a nuclear reactor. It includes a control gas having a high neutron cross-section; a first tank containing a first supply of the control gas; a first conduit providing a first fluid passage extending into the core, the first conduit being operatively connected to communicate with the first tank; a first valve operatively connected to regulate the flow of the control gas between the first tank and the first conduit; a second conduit concentrically disposed around the first conduit such that a second fluid passage is defined between the outer surface of the first conduit and the inner surface of the second conduit; a second tank containing a second supply of the control gas, the second tank being operatively connected to communicate with the second fluid passage; a second supply valve operatively connected to regulate the flow of the control gas between the second tank and the second fluid passage.

Abdel-Khalik, S.

1991-09-03T23:59:59.000Z

290

Overview of the US stellarator reactor study  

SciTech Connect (OSTI)

This study, which uses a cost-minimization code that incorporates the ARIES costing and reactor component models with a I-D energy transport calculation, shows that a torsatron reactor could be competitive with a tokamak reactor.

Lyon, J.F. [Oak Ridge National Lab., TN (United States); Gulec, K. [Univ. of Tennessee, Knoxville, TN (United States); Miller, R.L. [Los Alamos National Lab., NM (United States); El-Guebaly, L. [Univ. of Wisconsin, Madison, WI (United States)

1993-12-31T23:59:59.000Z

291

Reactor Cost Analysis Brian James  

E-Print Network [OSTI]

Reactor Cost Analysis Brian James Directed Technologies, Inc. 6-7 November 2007 This presentation specification & optimization · Capital cost estimation · Projected hydrogen $/kg #12;Directed Technologies, Inc/WGS Membrane Reactor OTM/ Water-Splitting ANL With WGS #12;Directed Technologies, Inc. 6-7 November 2007 BILIWG

292

Solvent refined coal reactor quench system  

DOE Patents [OSTI]

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

Thorogood, Robert M. (Macungie, PA)

1983-01-01T23:59:59.000Z

293

Solvent refined coal reactor quench system  

DOE Patents [OSTI]

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

Thorogood, R.M.

1983-11-08T23:59:59.000Z

294

Fast reactors and nuclear nonproliferation  

SciTech Connect (OSTI)

Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (authors)

Avrorin, E.N. [Russian Federal Nuclear Center - Zababakhin Institute of Applied Physics, Snezhinsk (Russian Federation); Rachkov, V.I.; Chebeskov, A.N. [State Scientific Center of the Russian Federation - Institute for Physics and Power Engineering, Bondarenko Square, 1, Obninsk, Kaluga region, 249033 (Russian Federation)

2013-07-01T23:59:59.000Z

295

Implications of Fast Reactor Transuranic Conversion Ratio  

SciTech Connect (OSTI)

Theoretically, the transuranic conversion ratio (CR), i.e. the transuranic production divided by transuranic destruction, in a fast reactor can range from near zero to about 1.9, which is the average neutron yield from Pu239 minus 1. In practice, the possible range will be somewhat less. We have studied the implications of transuranic conversion ratio of 0.0 to 1.7 using the fresh and discharge fuel compositions calculated elsewhere. The corresponding fissile breeding ratio ranges from 0.2 to 1.6. The cases below CR=1 (“burners”) do not have blankets; the cases above CR=1 (“breeders”) have breeding blankets. The burnup was allowed to float while holding the maximum fluence to the cladding constant. We graph the fuel burnup and composition change. As a function of transuranic conversion ratio, we calculate and graph the heat, gamma, and neutron emission of fresh fuel; whether the material is “attractive” for direct weapon use using published criteria; the uranium utilization and rate of consumption of natural uranium; and the long-term radiotoxicity after fuel discharge. For context, other cases and analyses are included, primarily once-through light water reactor (LWR) uranium oxide fuel at 51 MWth-day/kg-iHM burnup (UOX-51). For CR<1, the heat, gamma, and neutron emission increase as material is recycled. The uranium utilization is at or below 1%, just as it is in thermal reactors as both types of reactors require continuing fissile support. For CR>1, heat, gamma, and neutron emission decrease with recycling. The uranium utilization exceeds 1%, especially as all the transuranic elements are recycled. exceeds 1%, especially as all the transuranic elements are recycled. At the system equilibrium, heat and gamma vary by somewhat over an order of magnitude as a function of CR. Isotopes that dominate heat and gamma emission are scattered throughout the actinide chain, so the modest impact of CR is unsurprising. Neutron emitters are preferentially found among the higher actinides, so the neutron emission varies much stronger with CR, about three orders of magnitude.

Steven J. Piet; Edward A. Hoffman; Samuel E. Bays

2010-11-01T23:59:59.000Z

296

MOOSE simulating nuclear reactor CRUD buildup  

SciTech Connect (OSTI)

This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

None

2014-02-06T23:59:59.000Z

297

THE MATERIALS OF FAST BREEDER REACTORS  

E-Print Network [OSTI]

metal fast breeder reactor (LMFBR) concern the behavior ofmetal fast breeder reactor (LMFBR). Despite the simplicityinduced by irradiation. LMFBR funding is the largest single

Olander, Donald R.

2013-01-01T23:59:59.000Z

298

MOOSE simulating nuclear reactor CRUD buildup  

ScienceCinema (OSTI)

This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

None

2014-07-21T23:59:59.000Z

299

LIGHT WATER REACTOR SUSTAINABILITY PROGRAM: INTRODUCTION  

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

LIGHT WATER REACTOR SUSTAINABILITY PROGRAM: INTRODUCTION The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1...

300

Granular Dynamics in Pebble Bed Reactor Cores  

E-Print Network [OSTI]

gas reactors with the effective heat transfer of a molten salt coolant and the passive natural circulation safety systems of sodium fast reactors.

Laufer, Michael Robert

2013-01-01T23:59:59.000Z

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


301

Granular Dynamics in Pebble Bed Reactor Cores  

E-Print Network [OSTI]

pebble bed reactor,” Nuclear Engineering and Design, vol.the AVR reactor,” Nuclear Engineering and Design, vol. 121,Operating Experience,” Nuclear Engineering and Design, vol.

Laufer, Michael Robert

2013-01-01T23:59:59.000Z

302

Automatic reactor power control for a pressurized water reactor  

SciTech Connect (OSTI)

An automatic reactor power control system is presented for a pressurized water reactor (PWR). The associated reactor control strategy is called mode K.' The new system implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial shape change, which allows automatic control of the axial power distribution. Thus, the mode K enables precise regulation of both the reactivity and the power distribution, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load-follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1,000-MW (electric) PWR. The simulation results illustrate that the mode K would be a practical reactor power control strategy for the increased automation of nuclear plants.

Jungin Choi (Kyungwon Univ. (Korea, Republic of)); Yungjoon Hah (Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)); Unchul Lee (Seoul National Univ. (Korea, Republic of))

1993-05-01T23:59:59.000Z

303

Demonstration of Energy Efficient Steam Reforming in Microchannels for Automotive Fuel Processing  

SciTech Connect (OSTI)

A compact, energy efficient microchannel steam reforming system has been demonstrated. The unit generates sufficient reformate to provide H2 to a 10 kWe PEM fuel cell (when coupled with a water-gas shift and CO cleanup reactors). The overall volume of the reactor is 4.9 liters while that of the supporting network of heat exchangers is 1.7 liters . Use of a microchannel configuration in the steam reforming reactor produces rapid heat and mass transport which enables fast kinetics for the highly endothermic reaction. Heat is provided to the reactor by a combustion gas flowing in interleaved microchannels in cross flow with the reaction channels. A network of microchannel heat exchangers allows recovery of heat in the reformate product and combustion exhaust streams for use in vaporizing water and fuel, preheating reactants to reactor temperature and preheating combustion air. The microchannel architecture enables very compact and highly effective heat exchangers to be constructed. As a result of the heat exchange network, the system exhaust temperatures are typically ~50?C for the combustion gas and ~130?C for the reformate product while the reactor is operated at 750?C. While reforming isooctane at a rate sufficient to supply a 13.7 kWe fuel cell the system achieved 98.6% conversion with an estimated overall system efficiency after integration with WGS and PEM fuel cell of 44% (electrical output / LHV fuel). The efficiency estimate assumes integration with a WGS reactor (90% conversion CO to CO2 with 100% selectivity) and a PEM fuel cell (64% power conversion effectiveness with 85% H2 utilization for an overall 54% efficiency) and does not include parasitic losses for compression of combustion air. Acknowledgement The work described here was funded by the U.S. Department of Energy, Office of Transportation Technology as part of the OTT Fuel Cells Program.

Whyatt, Greg A.; TeGrotenhuis, Ward E.; Geeting, John GH; Davis, James M.; Wegeng, Robert S.; Pederson, Larry R.

2002-01-01T23:59:59.000Z

304

HTI retrieval demonstration project execution plan  

SciTech Connect (OSTI)

This plan describes the process for demonstrating the retrieval of difficult Hanford tank waste forms utilizing commercial technologies and the private sector to conduct the operations. The demonstration is to be conducted in Tank 241-C-106.

Ellingson, D.R.

1997-09-04T23:59:59.000Z

305

Bayesian nonparametric reward learning from demonstration  

E-Print Network [OSTI]

Learning from demonstration provides an attractive solution to the problem of teaching autonomous systems how to perform complex tasks. Demonstration opens autonomy development to non-experts and is an intuitive means of ...

Michini, Bernard (Bernard J.)

2013-01-01T23:59:59.000Z

306

Nuclear reactor control apparatus  

DOE Patents [OSTI]

Nuclear reactor safety rod release apparatus comprises a ring which carries detents normally positioned in an annular recess in outer side of the rod, the ring being held against the lower end of a drive shaft by magnetic force exerted by a solenoid carried by the drive shaft. When the solenoid is de-energized, the detent-carrying ring drops until the detents contact a cam surface associated with the lower end of the drive shaft, at which point the detents are cammed out of the recess in the safety rod to release the rod from the drive shaft. In preferred embodiments of the invention, an additional latch is provided to release a lower portion of a safety rod under conditions that may interfere with movement of the entire rod.

Sridhar, Bettadapur N. (Cupertino, CA)

1983-10-25T23:59:59.000Z

307

Nuclear reactor control  

DOE Patents [OSTI]

1. In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

Cawley, William E. (Phoenix, AZ); Warnick, Robert F. (Pasco, WA)

1982-01-01T23:59:59.000Z

308

Nuclear reactor control  

SciTech Connect (OSTI)

In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

Cawley, W.E.; Warnick, R.F.

1982-03-30T23:59:59.000Z

309

Webinar: National Hydrogen Learning Demonstration Status  

Broader source: Energy.gov [DOE]

Video recording and text version of the webinar, "National Hydrogen Learning Demonstration Status," originally presented on February 6, 2012.

310

Achieving and Demonstrating Vehicle Technologies Engine Fuel...  

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

Engine Fuel Efficiency Milestones Achieving and Demonstrating Vehicle Technologies Engine Fuel Efficiency Milestones 2009 DOE Hydrogen Program and Vehicle Technologies...

311

Webinar: Hydrogen Storage Materials Database Demonstration  

Broader source: Energy.gov [DOE]

Video recording and text version of the webinar, Hydrogen Storage Materials Database Demonstration, originally presented on December 13, 2011.

312

Industrial Scale Demonstration of Smart Manufacturing Achieving...  

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

Scale Demonstration of Smart Manufacturing Achieving Transformational Energy Productivity Gains Development of an Open Architecture, Widely Applicable Smart Manufacturing...

313

Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One  

SciTech Connect (OSTI)

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials.

Glasstone, S.; Sesonske, A.

1994-12-31T23:59:59.000Z

314

University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor  

SciTech Connect (OSTI)

The Department of Energy’s Idaho National Laboratory, under its programmatic responsibility for managing the University Research Reactor Conversions, has completed the conversion of the reactor at Purdue University Reactor. With this work completed and in anticipation of other impending conversion projects, the INL convened and engaged the project participants in a structured discussion to capture the lessons learned. The lessons learned process has allowed us to capture gaps, opportunities, and good practices, drawing from the project team’s experiences. These lessons will be used to raise the standard of excellence, effectiveness, and efficiency in all future conversion projects.

Eric C. Woolstenhulme; Dana M. Hewit

2008-09-01T23:59:59.000Z

315

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect (OSTI)

This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

Not Available

1993-11-01T23:59:59.000Z

316

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

SciTech Connect (OSTI)

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

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

2012-07-01T23:59:59.000Z

317

Materials Degradation in Light Water Reactors: Life After 60,???  

SciTech Connect (OSTI)

Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the key issue with materials aging and cable/piping as the top concerns for plant reliability. Materials degradation within a nuclear power plant is very complex. There are many different types of materials within the reactor itself: over 25 different metal alloys can be found with can be found within the primary and secondary systems, not to mention the concrete containment vessel, instrumentation and control, and other support facilities. When this diverse set of materials is placed in the complex and harsh environment coupled with load, degradation over an extended life is indeed quite complicated. To address this issue, the USNRC has developed a Progressive Materials Degradation Approach (NUREG/CR-6923). This approach is intended to develop a foundation for appropriate actions to keep materials degradation from adversely impacting component integrity and safety and identify materials and locations where degradation can reasonably be expected in the future. Clearly, materials degradation will impact reactor reliability, availability, and potentially, safe operation. Routine surveillance and component replacement can mitigate these factors, although failures still occur. With reactor life extensions to 60 years or beyond or power uprates, many components must tolerate the reactor environment for even longer times. This may increase susceptibility for most components and may introduce new degradation modes. While all components (except perhaps the reactor vessel) can be replaced, it may not be economically favorable. Therefore, understanding, controlling, and mitigating materials degradation processes are key priorities for reactor operation, power uprate considerations, and life extensions. This document is written to give an overview of some of the materials degradation issues that may be key for extend reactor service life. A detailed description of all the possible forms of degradation is beyond the scope of this short paper and has already been described in other documents (for example, the NUREG/CR-6923). The intent of this document is to present an overview of current materials issues in the existing reactor fleet and a brief analysis of the potential impact of extending life beyond 60 years. Discussion is presented in six distinct areas: (1) Reactor pressure vessel; (2) Reactor core and primary systems; (3) Reactor secondary systems; (4) Weldments; (5) Concrete; and (6) Modeling and simulations. Following each of these areas, some research thrust directions to help identify and mitigate lifetime extension issues are proposed. Note that while piping and cabling are important for extended service, these components are discussed in more depth in a separate paper. Further, the materials degradation issues associated with fuel cladding and fuel assemblies are not discussed in this section as these components are replaced periodically and will not influence the overall lifetime of the reactor.

Busby, Jeremy T [ORNL; Nanstad, Randy K [ORNL; Stoller, Roger E [ORNL; Feng, Zhili [ORNL; Naus, Dan J [ORNL

2008-04-01T23:59:59.000Z

318

A Multivariate Time Series Method for Monte Carlo Reactor Analysis  

SciTech Connect (OSTI)

A robust multivariate time series method has been established for the Monte Carlo calculation of neutron multiplication problems. The method is termed Coarse Mesh Projection Method (CMPM) and can be implemented using the coarse statistical bins for acquisition of nuclear fission source data. A novel aspect of CMPM is the combination of the general technical principle of projection pursuit in the signal processing discipline and the neutron multiplication eigenvalue problem in the nuclear engineering discipline. CMPM enables reactor physicists to accurately evaluate major eigenvalue separations of nuclear reactors with continuous energy Monte Carlo calculation. CMPM was incorporated in the MCNP Monte Carlo particle transport code of Los Alamos National Laboratory. The great advantage of CMPM over the traditional Fission Matrix method is demonstrated for the three space-dimensional modeling of the initial core of a pressurized water reactor.

Taro Ueki

2008-08-14T23:59:59.000Z

319

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

SciTech Connect (OSTI)

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

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

2002-09-01T23:59:59.000Z

320

Reduced enrichment for research and test reactors: Proceedings  

SciTech Connect (OSTI)

The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

Not Available

1988-05-01T23:59:59.000Z

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


321

Reactor Neutrino Flux Uncertainty Suppression on Multiple Detector Experiments  

E-Print Network [OSTI]

This publication provides a coherent treatment for the reactor neutrino flux uncertainties suppression, specially focussed on the latest $\\theta_{13}$ measurement. The treatment starts with single detector in single reactor site, most relevant for all reactor experiments beyond $\\theta_{13}$. We demonstrate there is no trivial error cancellation, thus the flux systematic error can remain dominant even after the adoption of multi-detector configurations. However, three mechanisms for flux error suppression have been identified and calculated in the context of Double Chooz, Daya Bay and RENO sites. Our analysis computes the error {\\it suppression fraction} using simplified scenarios to maximise relative comparison among experiments. We have validated the only mechanism exploited so far by experiments to improve the precision of the published $\\theta_{13}$. The other two newly identified mechanisms could lead to total error flux cancellation under specific conditions and are expected to have major implications o...

Cucoanes, Andi; Cabrera, Anatael; Fallot, Muriel; Onillon, Anthony; Obolensky, Michel; Yermia, Frederic

2015-01-01T23:59:59.000Z

322

NEUTRON RADIOGRAPHY (NRAD) REACTOR 64-ELEMENT CORE UPGRADE  

SciTech Connect (OSTI)

The neutron radiography (NRAD) reactor is a 250 kW TRIGA (registered) (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The interim critical configuration developed during the core upgrade, which contains only 62 fuel elements, has been evaluated as an acceptable benchmark experiment. The final 64-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (approximately +/-1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

John D. Bess

2014-03-01T23:59:59.000Z

323

Reactor physics design of supercritical CO?-cooled fast reactors  

E-Print Network [OSTI]

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

Pope, Michael A. (Michael Alexander)

2004-01-01T23:59:59.000Z

324

Reactor protection system design alternatives for sodium fast reactors  

E-Print Network [OSTI]

Historically, unprotected transients have been viewed as design basis events that can significantly challenge sodium-cooled fast reactors. The perceived potential consequences of a severe unprotected transient in a ...

DeWitte, Jacob D. (Jacob Dominic)

2011-01-01T23:59:59.000Z

325

United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support  

SciTech Connect (OSTI)

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

Douglas Morrell

2011-03-01T23:59:59.000Z

326

Nuclear reactor downcomer flow deflector  

DOE Patents [OSTI]

A nuclear reactor having a coolant flow deflector secured to a reactor core barrel in line with a coolant inlet nozzle. The flow deflector redirects incoming coolant down an annulus between the core barrel and the reactor vessel. The deflector has a main body with a front side facing the fluid inlet nozzle and a rear side facing the core barrel. The rear side of the main body has at least one protrusion secured to the core barrel so that a gap exists between the rear side of the main body adjacent the protrusion and the core barrel. Preferably, the protrusion is a relief that circumscribes the rear side of the main body.

Gilmore, Charles B. (Greensburg, PA); Altman, David A. (Pittsburgh, PA); Singleton, Norman R. (Murrysville, PA)

2011-02-15T23:59:59.000Z

327

Spent Nuclear Fuel Dry Transfer System Cold Demonstration Project Final Report  

SciTech Connect (OSTI)

The spent nuclear fuel dry transfer system (DTS) provides an interface between large and small casks and between storage-only and transportation casks. It permits decommissioning of reactor pools after shutdown and allows the use of large storage-only casks for temporary onsite storage of spent nuclear fuel irrespective of reactor or fuel handling limitations at a reactor site. A cold demonstration of the DTS prototype was initiated in August 1996 at the Idaho National Engineering and Environmental Laboratory (INEEL). The major components demonstrated included the fuel assembly handling subsystem, the shield plug/lid handling subsystem, the cask interface subsystem, the demonstration control subsystem, a support frame, and a closed circuit television and lighting system. The demonstration included a complete series of DTS operations from source cask receipt and opening through fuel transfer and closure of the receiving cask. The demonstration included both normal operations and recovery from off-normal events. It was designed to challenge the system to determine whether there were any activities that could be made to jeopardize the activities of another function or its safety. All known interlocks were challenged. The equipment ran smoothly and functioned as designed. A few "bugs" were corrected. Prior to completion of the demonstration testing, a number of DTS prototype systems were modified to apply lessons learned to date. Additional testing was performed to validate the modifications. In general, all the equipment worked exceptionally well. The demonstration also helped confirm cost estimates that had been made at several points in the development of the system.

Christensen, Max R; McKinnon, M. A.

1999-12-01T23:59:59.000Z

328

Environmentally assisted cracking of light-water reactor materials  

SciTech Connect (OSTI)

Environmentally assisted cracking (EAC) of lightwater reactor (LWR) materials has affected nuclear reactors from the very introduction of the technology. Corrosion problems have afflicted steam generators from the very introduction of pressurized water reactor (PWR) technology. Shippingport, the first commercial PWR operated in the United States, developed leaking cracks in two Type 304 stainless steel (SS) steam generator tubes as early as 1957, after only 150 h of operation. Stress corrosion cracks were observed in the heat-affected zones of welds in austenitic SS piping and associated components in boiling-water reactors (BRWs) as early as 1965. The degradation of steam generator tubing in PWRs and the stress corrosion cracking (SCC) of austenitic SS piping in BWRs have been the most visible and most expensive examples of EAC in LWRs, and the repair and replacement of steam generators and recirculation piping has cost hundreds of millions of dollars. However, other problems associated with the effects of the environment on reactor structures and components am important concerns in operating plants and for extended reactor lifetimes. Cast duplex austenitic-ferritic SSs are used extensively in the nuclear industry to fabricate pump casings and valve bodies for LWRs and primary coolant piping in many PWRs. Embrittlement of the ferrite phase in cast duplex SS may occur after 10 to 20 years at reactor operating temperatures, which could influence the mechanical response and integrity of pressure boundary components during high strain-rate loading (e.g., seismic events). The problem is of most concern in PWRs where slightly higher temperatures are typical and cast SS piping is widely used.

Chopra, O.K.; Chung, H.M.; Kassner, T.F.; Shack, W.J.

1996-02-01T23:59:59.000Z

329

Evaluating Russian space nuclear reactor technology for United States applications  

SciTech Connect (OSTI)

Space nuclear power and nuclear electric propulsion are considered important technologies for planetary exploration, as well as selected earth orbit applications. The Nuclear Electric Propulsion Space Test Program (NEPSTP) was intended to provide an early flight demonstration of these technologies at relatively low cost through extensive use of existing Russian technology. The key element of Russian technology employed in the program was the Topaz II reactor. Refocusing of the activities of the Ballistic Missile Defense Organization (BMDO), combined with budgetary pressures, forced the cancellation of the NEPSTP at the end of the 1993 fiscal year. The NEPSTP was faced with many unique flight qualification issues. In general, the launch of a spacecraft employing a nuclear reactor power system complicates many spacecraft qualification activities. However, the NEPSTP activities were further complicated because the reactor power system was a Russian design. Therefore, this program considered not only the unique flight qualification issues associated with space nuclear power, but also with differences between Russian and United States flight qualification procedures. This paper presents an overview of the NEPSTP. The program goals, the proposed mission, the spacecraft, and the Topaz II space nuclear power system are described. The subject of flight qualification is examined and the inherent difficulties of qualifying a space reactor are described. The differences between United States and Russian flight qualification procedures are explored. A plan is then described that was developed to determine an appropriate flight qualification program for the Topaz II reactor to support a possible NEPSTP launch.

Polansky, G.F. [Phillips Lab., Albuquerque, NM (United States); Schmidt, G.L. [New Mexico Engineering Research Institute, Albuquerque, NM (United States); Voss, S.S. [Los Alamos National Lab., NM (United States); Reynolds, E.L. [Applied Physics Lab., Laurel, MD (United States)

1994-08-01T23:59:59.000Z

330

2012 Annual Report Research Reactor Infrastructure Program  

SciTech Connect (OSTI)

The content of this report is the 2012 Annual Report for the Research Reactor Infrastructure Program.

Douglas Morrell

2012-11-01T23:59:59.000Z

331

PPPL3224, Preprint: February 1997, UC420, 426 Plasma Transport Control and SelfSustaining Fusion Reactor*  

E-Print Network [OSTI]

to demonstrate this concept with D­T reactor­grade plasmas. For edge transport control, a method based an economical and environmentally sound fusion reactor prototype is the ultimate goal of controlled fusionPPPL­3224, Preprint: February 1997, UC­420, 426 Plasma Transport Control and Self­Sustaining Fusion

332

PPPL-3224, Preprint: February 1997, UC-420, 426 Plasma Transport Control and Self-Sustaining Fusion Reactor*  

E-Print Network [OSTI]

to demonstrate this concept with D-T reactor-grade plasmas. For edge transport control, a method based an economical and environmentally sound fusion reactor prototype is the ultimate goal of controlled fusionPPPL-3224, Preprint: February 1997, UC-420, 426 Plasma Transport Control and Self-Sustaining Fusion

333

Unique features of space reactors  

SciTech Connect (OSTI)

Space reactors are designed to meet a unique set of requirements; they must be sufficiently compact to be launched in a rocket to their operational location, operate for many years without maintenance and servicing, operate in extreme environments, and reject heat by radiation to space. To meet these restrictions, operating temperatures are much greater than in terrestrial power plants, and the reactors tend to have a fast neutron spectrum. Currently, a new generation of space reactor power plants is being developed. The major effort is in the SP-100 program, where the power plant is being designed for seven years of full power, and no maintenance operation at a reactor outlet operating temperature of 1350 K. 8 refs., 3 figs., 1 tab.

Buden, D.

1990-01-01T23:59:59.000Z

334

Nuclear Reactors and Technology; (USA)  

SciTech Connect (OSTI)

Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

Cason, D.L.; Hicks, S.C. (eds.)

1991-01-01T23:59:59.000Z

335

Spectral shift reactor control method  

SciTech Connect (OSTI)

The method is described of closely controlling the reactor water coolant temperature of an operating spectral-shift nuclear reactor, the reactor comprising a core formed of fuel assemblies through which the reactor water coolant flows; different types of elongated elements operable to be controllably moved into and out of the core; one type of the elongated elements comprising control rods formed of neutron absorbing material and operable to decrease reactivity through neutron absorption when inserted into the core; another of the types of elongated elements comprising displacer rods formed of material which has a low absorption for neutrons and which have overall neutron-absorbing and moderating characteristics essentially not exceeding those of hollow tubular Zircaloy members with a filling zirconium oxide or aluminum oxide, the displacer rods operating to displace an equivalent volume of water coolant fluid from the core when inserted therein to decrease reactivity and to increase reactivity when moved from the core.

Impink, A.J. Jr.

1987-08-18T23:59:59.000Z

336

Reactor core isolation cooling system  

DOE Patents [OSTI]

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom. 1 figure.

Cooke, F.E.

1992-12-08T23:59:59.000Z

337

University Reactor Matching Grants Program  

SciTech Connect (OSTI)

During the 2002 Fiscal year, funds from the DOE matching grant program, along with matching funds from the industrial sponsors, have been used to support research in the area of thermal-hydraulics. Both experimental and numerical research projects have been performed. Experimental research focused on two areas: (1) Identification of the root cause mechanism for axial offset anomaly in pressurized water reactors under prototypical reactor conditions, and (2) Fluid dynamic aspects of thin liquid film protection schemes for inertial fusion reactor chambers. Numerical research focused on two areas: (1) Multi-fluid modeling of both two-phase and two-component flows for steam conditioning and mist cooling applications, and (2) Modeling of bounded Rayleigh-Taylor instability with interfacial mass transfer and fluid injection through a porous wall simulating the ''wetted wall'' protection scheme in inertial fusion reactor chambers. Details of activities in these areas are given.

John Valentine; Farzad Rahnema; Said Abdel-Khalik

2003-02-14T23:59:59.000Z

338

Interfacial effects in fast reactors  

E-Print Network [OSTI]

The problem of increased resonance capture rates near zone interfaces in fast reactor media has been examined both theoretically and experimentally. An interface traversing assembly was designed, constructed and employed ...

Saidi, Mohammad Said

1979-01-01T23:59:59.000Z

339

Teaching About Nature's Nuclear Reactors  

E-Print Network [OSTI]

Naturally occurring nuclear reactors existed in uranium deposits on Earth long before Enrico Fermi built the first man-made nuclear reactor beneath Staggs Field in 1942. In the story of their discovery, there are important lessons to be learned about scientific inquiry and scientific discovery. Now, there is evidence to suggest that the Earth's magnetic field and Jupiter's atmospheric turbulence are driven by planetary-scale nuclear reactors. The subject of planetocentric nuclear fission reactors can be a jumping off point for stimulating classroom discussions about the nature and implications of planetary energy sources and about the geomagnetic field. But more importantly, the subject can help to bring into focus the importance of discussing, debating, and challenging current thinking in a variety of areas.

Herndon, J M

2005-01-01T23:59:59.000Z

340

Reactor core isolation cooling system  

DOE Patents [OSTI]

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom.

Cooke, Franklin E. (San Jose, CA)

1992-01-01T23:59:59.000Z

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


341

Reactor physics project final report  

E-Print Network [OSTI]

This is the final report in an experimental and theoretical program to develop and apply single- and few-element methods for the determination of reactor lattice parameters. The period covered by the report is January 1, ...

Driscoll, Michael J.

1970-01-01T23:59:59.000Z

342

Automatic safety rod for reactors  

DOE Patents [OSTI]

An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-core flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

Germer, John H. (San Jose, CA)

1988-01-01T23:59:59.000Z

343

Computer aided nuclear reactor modeling  

E-Print Network [OSTI]

CHAPTER Page IV ALPHA ARCHITECTURE Design Philosophy Abstract Data Type Based Modules Grouping by Functions Miscellaneous Design Influences Architecture . . X Window System . Editor Library Model Library User Interface Library . V CONCLUSIONS... Connected Model . . . . , . . . 31 12 13 Header Section Editor Editing a "Choice" Attribute A Table of Vectors . 32 33 . 34 14 15 16 Current Reactor Modeling Schematic Reactor Modeling Schematic with Alpha Public Header File of Vertex Module...

Warraich, Khalid Sarwar

1995-01-01T23:59:59.000Z

344

Fast quench reactor and method  

DOE Patents [OSTI]

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

Detering, Brent A. (Idaho Falls, ID); Donaldson, Alan D. (Idaho Falls, ID); Fincke, James R. (Idaho Falls, ID); Kong, Peter C. (Idaho Falls, ID)

1998-01-01T23:59:59.000Z

345

Design and simulation of a plant control system for a GCFR demonstration plant  

SciTech Connect (OSTI)

A plant control system is being designed for a 300 MW(e) Gas Cooled Fast Breeder Reactor (GCFR) demonstration plant. Control analysis is being performed as an integral part of the plant design process to ensure that control requirements are satisfied as the plant design evolves. Plant models and simulations are being developed to generate information necessary to further define control system requirements for subsequent plant design iterations.

Estrine, E.A.; Greiner, H.G.

1980-02-01T23:59:59.000Z

346

Cryogenic Barrier Demonstration Project. Final report  

SciTech Connect (OSTI)

A long-term frozen soil barrier was implemented at the HRE (Homogeneous Reactor Experiment) Pond facility at the Oak Ridge National Laboratory in 1997. This was performed to verify the technical feasibility and costs of deploying a frozen barrier at a radiologically contaminated site. Work began in September 1996 and progressed through to December 1999. The frozen barrier has been operational since November 1997. Verification of the barrier integrity was performed independently by the EPA's SITE Program. This project showed frozen barriers offer a proven technology to retain below grade hazardous substances at relatively low costs with minimal effect on the environment.

Johnson, L.A.; Yarmak, E.; Long, E.L.

2000-03-01T23:59:59.000Z

347

Manufacturing Demonstration Facility Workshop | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Careerlumens_placard-green.eps More Documents &Small Modular ReactorsNuclear Safety at theFlow

348

Test and Demonstration Assets of New Mexico  

SciTech Connect (OSTI)

This document was developed by the Arrowhead Center of New Mexico State University as part of the National Security Preparedness Project (NSPP), funded by a DOE/NNSA grant. The NSPP has three primary components: business incubation, workforce development, and technology demonstration and validation. The document contains a survey of test and demonstration assets in New Mexico available for external users such as small businesses with security technologies under development. Demonstration and validation of national security technologies created by incubator sources, as well as other sources, are critical phases of technology development. The NSPP will support the utilization of an integrated demonstration and validation environment.

None

2008-03-31T23:59:59.000Z

349

Grid Connectivity Research, Development & Demonstration Projects  

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

and communication technology * Engage demonstration partners to validate NIST smart grid standards for utility network and sub-meter requirements (includes ANSI, NEMA...

350

Learning Demonstration Progress Report -- Spring 2008  

SciTech Connect (OSTI)

This report documents key results from DOE's Controlled Hydrogen Fleet and Infrastructure Validation and Demonstration Project based on data through December 2007.

Wipke, K.; Sprik, S.; Kurtz, J.

2008-04-01T23:59:59.000Z

351

Enterprise Assessments Review, West Valley Demonstration Project...  

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

conducted an independent oversight review of activity-level implementation of the radiation protection program at the West Valley Demonstration Project. The onsite review was...

352

Achieving and Demonstrating Vehicle Technologies Engine Fuel...  

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

Vehicle Technologies Engine Fuel Efficiency Milestones Achieving and Demonstrating Vehicle Technologies Engine Fuel Efficiency Milestones 2010 DOE Vehicle Technologies and Hydrogen...

353

Independent Oversight Review, West Valley Demonstration Project...  

Office of Environmental Management (EM)

West Valley Demonstration Project - December 2014 3Q CY2005 (PDF), Facility Representative Program Performance Indicators Quarterly Report EA-1552: Final Environmental Assessment...

354

The Smithsonian American Art Museum GATEWAY Demonstration  

Broader source: Energy.gov [DOE]

View the video about using LEDs in a GATEWAY demonstration at the Smithsonian American Art Museum in Washington, DC, including an interview with lighting designer Scott Rosenfeld.

355

Oak Ridge City Center Technology Demonstration Project  

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

Oak Ridge City Center Technology Demonstration Project David Thrash, Principal Investigator Oak Ridge City Center, LLC Track Name May 18, 2010 This presentation does not contain...

356

Oak Ridge National Laboratory Manufacturing Demonstration Facility  

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

Oak Ridge National Laboratory Manufacturing Demonstration Facility Technology Collaborations | Proposal Guidelines Proposal Guidelines Proposals should be no more than 5 single...

357

Sapphire Energy, Inc. Demonstration-Scale Project  

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

and run the facility. The success of this project will demonstrate the technical and economic feasibility of the algae-to- fuels process for commercial-scale biorefineries....

358

Independent Activity Report, West Valley Demonstration Project...  

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

July 2012 Operational Awareness Oversight of the West Valley Demonstration Project HIAR WVDP-2012-07-30 This Independent Activity Report documents an operational awareness...

359

Daemen Alternative Energy/Geothermal Technologies Demonstration...  

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

Daemen Alternative EnergyGeothermal Technologies Demonstration Program Erie County Robert C. Beiswanger, Jr. Daemen College May 20, 2010 This presentation does not contain any...

360

Selection Criteria for Demonstration Projects | Department of...  

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

and Logistical Challenges to Smart Grid Implementation Demonstration and Deployment Workshop Day 1 Offshore Resource Assessment and Design Conditions Public Meeting Summary Report...

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


361

Commercial Building Demonstration and Deployment Overview - 2014...  

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

Office's Commercial Building Demonstration and Deployment activities. Through robust feedback, the BTO Program Peer Review enhances existing efforts and improves future designs....

362

Smart Grid Demonstration Funding Opportunity Announcement DE...  

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

Frequently asked questions about the Smart Grid Demonstration and Energy Storage Funding Opportunity Announcement released as part of the American Recovery and Reinvestment Act,...

363

Enforcement Documents - West Valley Demonstration Project | Department...  

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

Services - EA-1999-09 Issued to West Valley Nuclear Services, related to a High-Level Radioactive Waste Contamination Event at the West Valley Demonstration...

364

Demonstrating Fuel Consumption and Emissions Reductions with...  

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

Fuel Consumption and Emissions Reductions with Next Generation Model-Based Diesel Engine Control Demonstrating Fuel Consumption and Emissions Reductions with Next Generation...

365

Solar solids reactor  

DOE Patents [OSTI]

A solar powered kiln is provided, that is of relatively simple design and which efficiently uses solar energy. The kiln or solids reactor includes a stationary chamber with a rearward end which receives solid material to be reacted and a forward end through which reacted material is disposed of, and a screw conveyor extending along the bottom of the chamber for slowly advancing the material between the chamber ends. Concentrated solar energy is directed to an aperture at the forward end of the chamber to heat the solid material moving along the bottom of the chamber. The solar energy can be reflected from a mirror facing at an upward incline, through the aperture and against a heat-absorbing material near the top of the chamber, which moves towards the rear of the chamber to distribute heat throughout the chamber. Pumps at the forward and rearward ends of the chamber pump heated sweep gas through the length of the chamber, while minimizing the flow of gas through an open aperture through which concentrated sunlight is received.

Yudow, B.D.

1986-02-24T23:59:59.000Z

366

Solar solids reactor  

DOE Patents [OSTI]

A solar powered kiln is provided, that is of relatively simple design and which efficiently uses solar energy. The kiln or solids reactor includes a stationary chamber with a rearward end which receives solid material to be reacted and a forward end through which reacted material is disposed of, and a screw conveyor extending along the bottom of the chamber for slowly advancing the material between the chamber ends. Concentrated solar energy is directed to an aperture at the forward end of the chamber to heat the solid material moving along the bottom of the chamber. The solar energy can be reflected from a mirror facing at an upward incline, through the aperture and against a heat-absorbing material near the top of the chamber, which moves towards the rear of the chamber to distribute heat throughout the chamber. Pumps at the forward and rearward ends of the chamber pump heated sweep gas through the length of the chamber, while minimizing the flow of gas through an open aperture through which concentrated sunlight is received.

Yudow, Bernard D. (Chicago, IL)

1987-01-01T23:59:59.000Z

367

Nuclear reactor control rod  

SciTech Connect (OSTI)

This patent describes a vertically oriented bottom entry control rod from a nuclear reactor: a frame including an elongated central spine of cruciform cross section connected between an upper support member and a lower support member both of cruciform shape having four laterally extending arms. The arms are in alignment with the arms of the lower support member and each aligned upper and lower support members has a sheath extending between; absorber plates of neutron absorber material, different from the material of the frame, one of the absorber plates is positioned within a sheath beneath each of the arms; attachment means suspends the absorber plates from the arms of the upper support member within a sheath; elongated absorber members positioned within a sheath between each of the suspended absorber plates and an arm of the lower support member; and joint means between the upper ends of the absorber members and the lower ends of the suspended absorber plates for minimizing gaps; the sheath means encloses the suspended absorber plates and the absorber members extending between aligned arms of the upper and lower support members and secured.

Cearley, J.E.; Izzo, K.R.

1987-06-30T23:59:59.000Z

368

Reactor pressure vessel nozzle  

DOE Patents [OSTI]

A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough. 2 figs.

Challberg, R.C.; Upton, H.A.

1994-10-04T23:59:59.000Z

369

Integrated monitoring and surveillance system demonstration project: Phase I accomplishments  

SciTech Connect (OSTI)

The authors present the results of the Integrated Monitoring and Surveillance System (IMSS) demonstration project Phase I efforts. The rationale behind IMSS development is reviewed and progress in each of the 5 basic tasks is detailed. Significant results include decisions to use Echelon LonWorks networking protocol and Microsoft Access for the data system needs, a preliminary design for the plutonium canning system glovebox, identification of facilities and materials available for the demonstration, determination of possibly affected facility documentation, and a preliminary list of available sensor technologies. Recently imposed changes in the overall project schedule and scope are also discussed and budgetary requirements for competition of Phase II presented. The results show that the IMSS demonstration project team has met and in many cases exceeded the commitments made for Phase I deliverables.

Aumeier, S.E.; Walters, B.G.; Crawford, D.C. [and others

1997-01-15T23:59:59.000Z

370

Propellant actuated nuclear reactor steam depressurization valve  

DOE Patents [OSTI]

A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

Ehrke, Alan C. (San Jose, CA); Knepp, John B. (San Jose, CA); Skoda, George I. (Santa Clara, CA)

1992-01-01T23:59:59.000Z

371

When Do Commercial Reactors Permanently Shut Down?  

Reports and Publications (EIA)

For those wishing to obtain current data, the following resources are available: U.S. reactors, go to the Energy Information Administration's nuclear reactor shutdown list. (Note: As of April 30, 2010, the last U.S. reactor to permanently shut down was Big Rock Point in 1997.) Foreign Reactors, go to the Power Reactor Information System (PRIS) on the International Atomic Energy Agency's website.

2011-01-01T23:59:59.000Z

372

Reactor Technology Options Study for Near-Term Deployment of GNEP Grid-Appropriate Reactors  

SciTech Connect (OSTI)

World energy demand is projected to significantly increase over the coming decades. The International Energy Agency projects that electricity demand will increase 50% by 2015 and double by 2030, with most of the increase coming in developing countries as they experience double-digit rates of economic growth and seek to improve their standards of living. Energy is the necessary driver for human development, and the demand for energy in these countries will be met using whatever production technologies are available. Recognizing this inevitable energy demand and its implications for the United States, the U.S. National Security Strategy has proposed the Global Nuclear Energy Partnership (GNEP) to work with other nations to develop and deploy advanced nuclear recycling and reactor technologies. This initiative will help provide reliable, emission-free energy with less of the waste burden of older technologies and without making available separated plutonium that could be used by rogue states or terrorists for nuclear weapons. These new technologies will make possible a dramatic expansion of safe, clean nuclear energy to help meet the growing global energy demand. In other words, GNEP seeks to create an international regime to support large-scale growth in the worldwide use of nuclear energy without increasing the risk of nuclear weapon proliferation. This global expansion of nuclear power is strategically important to the United States for several reasons, including the following: (1) National security, by reducing the competition and potential for conflict over increasingly scarce fossil energy resources; (2) Economic security, by helping maintain stable prices for nonrenewable resources such as oil, gas, and coal; (3) Environmental security, by replacing or off-setting large-scale burning of greenhouse gas-emitting fuels for electricity production; and (4) Regaining technical leadership, through deployment of innovative U.S. technology-based reactors. Fully meeting the GNEP vision may require the deployment of thousands of reactors during the next century in dozens of countries, many of which are in the developing world where nuclear energy is not used currently. Such a large-scale deployment will have significant implications related to both fuel supply and spent fuel/waste management, both domestically and worldwide. Consequently, GNEP must address the development and demonstration of proliferation-resistant technologies to ensure both a safe and sustainable nuclear fuel cycle, and reactor designs that are appropriate for the range of needs across the global community. The focus of this report is the latter need, that is, the development and demonstration of proliferation-resistant reactors that are well matched to the needs and capabilities of developing countries.

Ingersoll, Daniel T [ORNL; Poore III, Willis P [ORNL

2007-09-01T23:59:59.000Z

373

DOE reactor-pumped laser program  

SciTech Connect (OSTI)

FALCON is a high-power, steady-state, nuclear reactor-pumped laser (RPL) concept that is being developed by the Department of Energy. The FALCON program has experimentally demonstrated reactor-pumped lasing in various mixtures of xenon, argon, neon, and helium at at wavelengths of 585, 703, 725, 1271, 1733, 1792, 2032, 2630, 2650, and 3370 nm with intrinsic efficiency as high as 2.5%. The major strengths of a reactor-pumped laser are continuous high-power operation, modular construction, self-contained power, compact size, and a variety of wavelengths (from visible to infrared). These characteristics suggest numerous applications not easily accessible to other laser types. A ground-based RPL could beam its power to space for such activities as illuminating geosynchronous communication satellites in the earth`s shadow to extend their lives, beaming power to orbital transfer vehicles, removing space debris, and providing power (from earth) to a lunar base during the long lunar night. The compact size and self-contained power also makes an RPL very suitable for ship basing so that power-beaming activities could be situated around the globe. The continuous high power of an RPL opens many potential manufacturing applications such as deep-penetration welding and cutting of thick structures, wide-area hardening of metal surfaces by heat treatment or cladding application, wide-area vapor deposition of ceramics onto metal surfaces, production of sub-micron sized particles for manufacturing of ceramics, wide-area deposition of diamond-like coatings, and 3-D ceramic lithography.

Felty, J.R. [USDOE, Germantown, MD (United States). Defense Programs; Lipinski, R.J.; McArthur, D.A.; Pickard, P.S. [Sandia National Labs., Albuquerque, NM (United States)

1993-12-31T23:59:59.000Z

374

Stability of tubular and autothermal packed bed reactors using phase plane analysis  

SciTech Connect (OSTI)

The regions of stability and parametric sensitivity of countercurrent reactor/heat exchangers are determined explicitly in the plane of inlet feed temperature-inlet coolant temperature. The concept of phase plane analysis is generalized to include all orders of reaction rate expressions, a broader range of system parameters, and is extended to the case of autothermal reactors. An industrial hydrocarbon oxidation reactor model and an autothermal CO oxidation reactor model have been used to illustrate and to evaluate the analysis method. The approach presented here is appealing since the region of safe inlet temperatures is determined explicitly and the region of safe operation can be optimized with respect to the reactor design parameters.

Chylla, R.W. Jr.; Adomaitis, R.A.; Cinar, A.

1987-07-01T23:59:59.000Z

375

Building Cognitive Tutors with Programming by Demonstration  

E-Print Network [OSTI]

, W. W., & Koedinger, K. R. (2005). Building Cognitive Tutors with Program- ming by Demonstration41 Building Cognitive Tutors with Programming by Demonstration Noboru Matsuda, William W. Cohen, Kenneth R. Koedinger School of Computer Science, Carnegie Mellon University 5000 Forbes Ave., Pittsburgh

Cohen, William W.

376

Buried waste integrated demonstration technology integration process  

SciTech Connect (OSTI)

A Technology integration Process was developed for the Idaho National Energy Laboratories (INEL) Buried Waste Integrated Demonstration (BWID) Program to facilitate the transfer of technology and knowledge from industry, universities, and other Federal agencies into the BWID; to successfully transfer demonstrated technology and knowledge from the BWID to industry, universities, and other Federal agencies; and to share demonstrated technologies and knowledge between Integrated Demonstrations and other Department of Energy (DOE) spread throughout the DOE Complex. This document also details specific methods and tools for integrating and transferring technologies into or out of the BWID program. The document provides background on the BWID program and technology development needs, demonstrates the direction of technology transfer, illustrates current processes for this transfer, and lists points of contact for prospective participants in the BWID technology transfer efforts. The Technology Integration Process was prepared to ensure compliance with the requirements of DOE's Office of Technology Development (OTD).

Ferguson, J.S.; Ferguson, J.E.

1992-04-01T23:59:59.000Z

377

Buried waste integrated demonstration technology integration process  

SciTech Connect (OSTI)

A Technology integration Process was developed for the Idaho National Energy Laboratories (INEL) Buried Waste Integrated Demonstration (BWID) Program to facilitate the transfer of technology and knowledge from industry, universities, and other Federal agencies into the BWID; to successfully transfer demonstrated technology and knowledge from the BWID to industry, universities, and other Federal agencies; and to share demonstrated technologies and knowledge between Integrated Demonstrations and other Department of Energy (DOE) spread throughout the DOE Complex. This document also details specific methods and tools for integrating and transferring technologies into or out of the BWID program. The document provides background on the BWID program and technology development needs, demonstrates the direction of technology transfer, illustrates current processes for this transfer, and lists points of contact for prospective participants in the BWID technology transfer efforts. The Technology Integration Process was prepared to ensure compliance with the requirements of DOE`s Office of Technology Development (OTD).

Ferguson, J.S.; Ferguson, J.E.

1992-04-01T23:59:59.000Z

378

Passive Safety Features for Small Modular Reactors  

SciTech Connect (OSTI)

The rapid growth in the size and complexity of commercial nuclear power plants in the 1970s spawned an interest in smaller, simpler designs that are inherently or intrinsically safe through the use of passive design features. Several designs were developed, but none were ever built, although some of their passive safety features were incorporated into large commercial plant designs that are being planned or built today. In recent years, several reactor vendors are actively redeveloping small modular reactor (SMR) designs with even greater use of passive features. Several designs incorporate the ultimate in passive safety they completely eliminate specific accident initiators from the design. Other design features help to reduce the likelihood of an accident or help to mitigate the accident s consequences, should one occur. While some passive safety features are common to most SMR designs, irrespective of the coolant technology, other features are specific to water, gas, or liquid-metal cooled SMR designs. The extensive use of passive safety features in SMRs promise to make these plants highly robust, protecting both the general public and the owner/investor. Once demonstrated, these plants should allow nuclear power to be used confidently for a broader range of customers and applications than will be possible with large plants alone.

Ingersoll, Daniel T [ORNL] [ORNL

2010-01-01T23:59:59.000Z

379

On the interest of carbon-coated plasma reactor for advanced gate stack etching processes  

SciTech Connect (OSTI)

In integrated circuit fabrication the most wide spread strategy to achieve acceptable wafer-to-wafer reproducibility of the gate stack etching process is to dry-clean the plasma reactor walls between each wafer processed. However, inherent exposure of the reactor walls to fluorine-based plasma leads to formation and accumulation of nonvolatile fluoride residues (such as AlF{sub x}) on reactor wall surfaces, which in turn leads to process drifts and metallic contamination of wafers. To prevent this while keeping an Al{sub 2}O{sub 3} reactor wall material, a coating strategy must be used, in which the reactor is coated by a protective layer between wafers. It was shown recently that deposition of carbon-rich coating on the reactor walls allows improvements of process reproducibility and reactor wall protection. The authors show that this strategy results in a higher ion-to-neutral flux ratio to the wafer when compared to other strategies (clean or SiOCl{sub x}-coated reactors) because the carbon walls load reactive radical densities while keeping the same ion current. As a result, the etching rates are generally smaller in a carbon-coated reactor, but a highly anisotropic etching profile can be achieved in silicon and metal gates, whose etching is strongly ion assisted. Furthermore, thanks to the low density of Cl atoms in the carbon-coated reactor, silicon etching can be achieved almost without sidewall passivation layers, allowing fine critical dimension control to be achieved. In addition, it is shown that although the O atom density is also smaller in the carbon-coated reactor, the selectivity toward ultrathin gate oxides is not reduced dramatically. Furthermore, during metal gate etching over high-k dielectric, the low level of parasitic oxygen in the carbon-coated reactor also allows one to minimize bulk silicon reoxidation through HfO{sub 2} high-k gate dielectric. It is then shown that the BCl{sub 3} etching process of the HfO{sub 2} high-k material is highly selective toward the substrate in the carbon-coated reactor, and the carbon-coating strategy thus allows minimizing the silicon recess of the active area of transistors. The authors eventually demonstrate that the carbon-coating strategy drastically reduces on-wafer metallic contamination. Finally, the consumption of carbon from the reactor during the etching process is discussed (and thus the amount of initial deposit that is required to protect the reactor walls) together with the best way of cleaning the reactor after a silicon etching process.

Ramos, R.; Cunge, G.; Joubert, O. [Freescale Semiconductor Inc., 850 Rue Jean Monnet, 38921 Crolles Cedex (France) and Laboratoire des Technologies de la Microelectronique, CNRS, 17 Rue des Martyrs (c/o CEA-LETI), 38054 Grenoble Cedex 9 (France); Laboratoire des Technologies de la Microelectronique, CNRS, 17 Rue des Martyrs (c/o CEA-LETI), 38054 Grenoble Cedex 9 (France)

2007-03-15T23:59:59.000Z

380

Reactor Pressure Vessel Head Packaging & Disposal  

SciTech Connect (OSTI)

Reactor Pressure Vessel (RPV) Head replacements have come to the forefront due to erosion/corrosion and wastage problems resulting from the susceptibility of the RPV Head alloy steel material to water/boric acid corrosion from reactor coolant leakage through the various RPV Head penetrations. A case in point is the recent Davis-Besse RPV Head project, where detailed inspections in early 2002 revealed significant wastage of head material adjacent to one of the Control Rod Drive Mechanism (CRDM) nozzles. In lieu of making ASME weld repairs to the damaged head, Davis-Besse made the decision to replace the RPV Head. The decision was made on the basis that the required weld repair would be too extensive and almost impractical. This paper presents the packaging, transport, and disposal considerations for the damaged Davis-Besse RPV Head. It addresses the requirements necessary to meet Davis Besse needs, as well as the regulatory criteria, for shipping and burial of the head. It focuses on the radiological characterization, shipping/disposal package design, site preparation and packaging, and the transportation and emergency response plans that were developed for the Davis-Besse RPV Head project.

Wheeler, D. M.; Posivak, E.; Freitag, A.; Geddes, B.

2003-02-26T23:59:59.000Z

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


381

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

SciTech Connect (OSTI)

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

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

2012-07-01T23:59:59.000Z

382

Experiments illustrating the importance of automated reasoning in spacecraft reactor control  

SciTech Connect (OSTI)

The importance of automated reasoning to the control of spacecraft nuclear reactors is discussed. Presented are results from experiments that demonstrate the role of planning, prediction, and assessment in the realization of autonomous control. These experiments were performed under closed-loop conditions on the Annular Core Research Reactor that is operated by the Sandia National Laboratories. Automated diagnostics is identified as an area that requires much further research.

Bernard, J.A. (Nuclear Reactor Laboratory, Massachusetts Institute of Technology, 138 Albany Street Cambridge, Massachusetts (USA)); Wyant, F.J. (Nuclear Technology Projects Division, Sandia National Laboratories, P.O. Box 5800 Albuquerque, New Mexico (USA))

1991-01-10T23:59:59.000Z

383

Conceptual design of a new homogeneous reactor for medical radioisotope Mo-99/Tc-99m production  

SciTech Connect (OSTI)

To partly solve the global and regional shortages of Mo-99 supply, a conceptual design of a nitrate-fuel-solution based homogeneous reactor dedicated for Mo-99/Tc-99m medical radioisotope production is proposed. The modified LEU Cintichem process for Mo-99 extraction which has been licensed and demonstrated commercially for decades by BATAN is taken into account as a key design consideration. The design characteristics and main parameters are identified and the advantageous aspects are shown by comparing with the BATAN's existing Mo-99 supply chain which uses a heterogeneous reactor (RSG GAS multipurpose reactor)

Liem, Peng Hong [Nippon Advanced Information Service (NAIS Co., Inc.) Scientific Computational Division, 416 Muramatsu, Tokaimura, Ibaraki (Japan); Tran, Hoai Nam [Chalmers University of Technology, Dept. of Applied Physics, Div. of Nuclear Engineering, SE-412 96 Gothenburg (Sweden); Sembiring, Tagor Malem [National Nuclear Energy Agency (BATAN), Center for Reactor Technology and Nuclear Safety, Kawasan Puspiptek, Serpong, Tangerang Selatan, Banten (Indonesia); Arbie, Bakri [PT MOTAB Technology, Kedoya Elok Plaza Blok DA 12, Jl. Panjang, Kebun Jeruk, Jakarta Barat (Indonesia)

2014-09-30T23:59:59.000Z

384

Advances in the development of wire mesh reactor for coal gasification studies - article no. 084102  

SciTech Connect (OSTI)

In an effort to further understand the coal gasification behavior in entrained-flow gasifiers, a high pressure and high temperature wire mesh reactor with new features was recently built. An advanced LABVIEW-based temperature measurement and control system were adapted. Molybdenum wire mesh with aperture smaller than 70 {mu} m and type D thermocouple were used to enable high carbon conversion ({gt}90%) at temperatures {gt}1000 {sup o}C. Gaseous species from wire mesh reactor were quantified using a high sensitivity gas chromatography. The material balance of coal pyrolysis in wire mesh reactor was demonstrated for the first time by improving the volatile's quantification techniques.

Zeng, C.; Chen, L.; Liu, G.; Li, W.H.; Huang, B.M.; Zhu, H.D.; Zhang, B.; Zamansky, V. [GE Global Research Shanghai, Shanghai (China)

2008-08-15T23:59:59.000Z

385

Uncertainties in the Anti-neutrino Production at Nuclear Reactors  

E-Print Network [OSTI]

neutrino Production at Nuclear Reactors Z. Djurcic 1 , ?emission rates from nuclear reactors are determined fromlarge commercial nuclear reactors are playing an important

Djurcic, Zelimir

2009-01-01T23:59:59.000Z

386

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network [OSTI]

Fundamental aspects of nuclear reactor fuel elements.Unlike permanent nuclear reactor core components, nuclearof the first nuclear reactors, commercial nuclear fuel still

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

387

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

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

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

388

Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation  

E-Print Network [OSTI]

selected as part of the Generation IV reactors .. - 4 -The development of Generation IV fast reactors can make aconcepts selected for the Generation IV reactors, three,

Heidet, Florent

2010-01-01T23:59:59.000Z

389

Global Optimization of Chemical Reactors and Kinetic Optimization  

E-Print Network [OSTI]

Model; 3-D; Monolith; Reactor; Optimization Introduction TheAngeles Global Optimization of Chemical Reactors and KineticGlobal Optimization of Chemical Reactors and Kinetic

ALHUSSEINI, ZAYNA ISHAQ

2013-01-01T23:59:59.000Z

390

Marketing Plan for Demonstration and Validation Assets  

SciTech Connect (OSTI)

The National Security Preparedness Project (NSPP), is to be sustained by various programs, including technology demonstration and evaluation (DEMVAL). This project assists companies in developing technologies under the National Security Technology Incubator program (NSTI) through demonstration and validation of technologies applicable to national security created by incubators and other sources. The NSPP also will support the creation of an integrated demonstration and validation environment. This report documents the DEMVAL marketing and visibility plan, which will focus on collecting information about, and expanding the visibility of, DEMVAL assets serving businesses with national security technology applications in southern New Mexico.

None

2008-05-30T23:59:59.000Z

391

Application of an exact model matching technique to coupled-core nuclear reactor control  

SciTech Connect (OSTI)

In this Note the control problem of linearized coupled-core multivariable nuclear reactors is treated by using a recent exact model matching technique in the frequency domain. The case of state feedback control is first considered and then the results are used where only the output variables of the reactor are available for feedback. A numerical example of a three coupled-core nuclear reactor model with one delayed neutron group for each core and short neutron travel time between cores is included.

Tzafestas, S.G.; Chrysochoides, N.G.; Rokkos, K.

1984-08-01T23:59:59.000Z

392

UCLA program in reactor studies: The ARIES tokamak reactor study  

SciTech Connect (OSTI)

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D-{sup 3}He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs.

Not Available

1991-01-01T23:59:59.000Z

393

LCEs for Naval Reactor Benchmark Calculations  

SciTech Connect (OSTI)

The purpose of this engineering calculation is to document the MCNP4B2LV evaluations of Laboratory Critical Experiments (LCEs) performed as part of the Disposal Criticality Analysis Methodology program. LCE evaluations documented in this report were performed for 22 different cases with varied design parameters. Some of these LCEs (10) are documented in existing references (Ref. 7.1 and 7.2), but were re-run for this calculation file using more neutron histories. The objective of this analysis is to quantify the MCNP4B2LV code system's ability to accurately calculate the effective neutron multiplication factor (k{sub eff}) for various critical configurations. These LCE evaluations support the development and validation of the neutronics methodology used for criticality analyses involving Naval reactor spent nuclear fuel in a geologic repository.

W.J. Anderson

1999-07-19T23:59:59.000Z

394

Rapid starting methanol reactor system  

DOE Patents [OSTI]

The invention relates to a methanol-to-hydrogen cracking reactor for use with a fuel cell vehicular power plant. The system is particularly designed for rapid start-up of the catalytic methanol cracking reactor after an extended shut-down period, i.e., after the vehicular fuel cell power plant has been inoperative overnight. Rapid system start-up is accomplished by a combination of direct and indirect heating of the cracking catalyst. Initially, liquid methanol is burned with a stoichiometric or slightly lean air mixture in the combustion chamber of the reactor assembly. The hot combustion gas travels down a flue gas chamber in heat exchange relationship with the catalytic cracking chamber transferring heat across the catalyst chamber wall to heat the catalyst indirectly. The combustion gas is then diverted back through the catalyst bed to heat the catalyst pellets directly. When the cracking reactor temperature reaches operating temperature, methanol combustion is stopped and a hot gas valve is switched to route the flue gas overboard, with methanol being fed directly to the catalytic cracking reactor. Thereafter, the burner operates on excess hydrogen from the fuel cells.

Chludzinski, Paul J. (38 Berkshire St., Swampscott, MA 01907); Dantowitz, Philip (39 Nancy Ave., Peabody, MA 01960); McElroy, James F. (12 Old Cart Rd., Hamilton, MA 01936)

1984-01-01T23:59:59.000Z

395

Reactor Neutrino Flux Uncertainty Suppression on Multiple Detector Experiments  

E-Print Network [OSTI]

This publication provides a coherent treatment for the reactor neutrino flux uncertainties suppression, specially focussed on the latest $\\theta_{13}$ measurement. The treatment starts with single detector in single reactor site, most relevant for all reactor experiments beyond $\\theta_{13}$. We demonstrate there is no trivial error cancellation, thus the flux systematic error can remain dominant even after the adoption of multi-detector configurations. However, three mechanisms for flux error suppression have been identified and calculated in the context of Double Chooz, Daya Bay and RENO sites. Our analysis computes the error {\\it suppression fraction} using simplified scenarios to maximise relative comparison among experiments. We have validated the only mechanism exploited so far by experiments to improve the precision of the published $\\theta_{13}$. The other two newly identified mechanisms could lead to total error flux cancellation under specific conditions and are expected to have major implications on the global $\\theta_{13}$ knowledge today. First, Double Chooz, in its final configuration, is the only experiment benefiting from a negligible reactor flux error due to a $\\sim$90\\% geometrical suppression. Second, Daya Bay and RENO could benefit from their partial geometrical cancellation, yielding a potential $\\sim$50\\% error suppression, thus significantly improving the global $\\theta_{13}$ precision today. And third, we illustrate the rationale behind further error suppression upon the exploitation of the inter-reactor error correlations, so far neglected. So, our publication is a key step forward in the context of high precision neutrino reactor experiments providing insight on the suppression of their intrinsic flux error uncertainty, thus affecting past and current experimental results, as well as the design of future experiments.

Andi Cucoanes; Pau Novella; Anatael Cabrera; Muriel Fallot; Anthony Onillon; Michel Obolensky; Frederic Yermia

2015-01-02T23:59:59.000Z

396

U.S. Department of Energy Program of International Technical Cooperation for Research Reactor Utilization  

SciTech Connect (OSTI)

The U.S. Department of Energy, National Nuclear Security Administration (DOE/NNSA) has initiated collaborations with the national nuclear authorities of Egypt, Peru, and Romania for the purpose of advancing the commercial potential and utilization of their respective research reactors. Under its Office of International Safeguards ''Sister Laboratory'' program, DOE/NNSA has undertaken numerous technical collaborations over the past decade intended to promote peaceful applications of nuclear technology. Among these has been technical assistance in research reactor applications, such as neutron activation analysis, nuclear analysis, reactor physics, and medical radioisotope production. The current collaborations are intended to provide the subject countries with a methodology for greater commercialization of research reactor products and services. Our primary goal is the transfer of knowledge, both in administrative and technical issues, needed for the establishment of an effective business plan and utilization strategy for the continued operation of the countries' research reactors. Technical consultation, cooperation, and the information transfer provided are related to: identification, evaluation, and assessment of current research reactor capabilities for products and services; identification of opportunities for technical upgrades for new or expanded products and services; advice and consultation on research reactor upgrades and technical modifications; characterization of markets for reactor products and services; identification of competition and estimation of potential for market penetration; integration of technical constraints; estimation of cash flow streams; and case studies.

Chong, D.; Manning, M.; Ellis, R.; Apt, K.; Flaim, S.; Sylvester, K.

2004-10-03T23:59:59.000Z

397

Calderon Cokemaking Process/Demonstration Project  

SciTech Connect (OSTI)

This project deals with the demonstration of a coking process using proprietary technology of Calderon with the following objectives in order to enable its commercialization: (i) making coke of such quality as to be suitable for use in high driving (highly productive) blast furnaces; (ii) providing proof that such process is continuous and environmentally closed to prevent emissions; and (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. The activities of the past quarter were entirely focused on operating the Calderon Process Development Unit (PDU-I) in Alliance, Ohio conducting a series of tests under steady state using coal from Bethlehem Steel and U.S. Steel in order to demonstrate the above. The objectives mentioned above were successfully demonstrated.

None

1998-04-08T23:59:59.000Z

398

Technoclimat- Green Technologies Demonstration Program (Quebec, Canada)  

Broader source: Energy.gov [DOE]

The Green technologies demonstration program aiming to reduce greenhouse gas emissions is a product of Measure 20 of the 2006-2012 Climate Change Action Plan (CCAP). This CCAP measure encourages...

399

National Hydrogen Learning Demonstration Status (Presentation)  

SciTech Connect (OSTI)

This presentation discusses U.S. DOE Learning Demonstration Project goals, fuel cell vehicle and H2 station deployment status, and technical highlights of vehicle and infrastructure analysis results and progress.

Wipke, K.; Sprik, S.; Kurtz, J.; Ramsden, T.; Ainscough, C.; Saur, G.

2012-02-01T23:59:59.000Z

400

"deinewahl02" A Demonstrator for Interactive Television  

E-Print Network [OSTI]

"deinewahl02" A Demonstrator for Interactive Television Dr. Sepideh Chakaveh Olaf Geuer Fraunhofer Institute Media Communication Fraunhofer Institute Media Communication Competence Center Interactive innovation. In here we are considering to reverse this process, i.e. implementing internet type applications

Masthoff, Judith

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


401

Demonstration and Deployment Strategy Workshop: Summary  

Broader source: Energy.gov [DOE]

This report is based on the proceedings of the U.S. DOE’s Bioenergy Technologies Office Demonstration and Deployment Strategy Workshop, held on March 12–13, 2014, at Argonne National Laboratory.

402

Landfill Gas Fueled HCCI Demonstration System  

E-Print Network [OSTI]

Journal of Engineering for Gas Turbines and Power, 121:569-operations with natural gas: Fuel composition implications,”USA ICEF2006-1578 LANDFILL GAS FUELED HCCI DEMONSTRATION

Blizman, Brandon J.; Makel, Darby B.; Mack, John Hunter; Dibble, Robert W.

2006-01-01T23:59:59.000Z

403

Status of the MAJORANA DEMONSTRATOR experiment  

SciTech Connect (OSTI)

The MAJORANA DEMONSTRATOR neutrinoless double beta-decay experiment is currently under construction at the Sanford Underground Research Facility in South Dakota, USA. An overview and status of the experiment are given.

Martin, R. D. [Nuclear Science Division, Lawrence Berkeley National Laboratory, Berkeley, CA, USA and Department of Physics, University of South Dakota, Vermillion, SD (United States); Abgrall, N.; Chan, Y-D.; Hegai, A.; Mertens, S.; Poon, A. W. P.; Vetter, K. [Nuclear Science Division, Lawrence Berkeley National Laboratory, Berkeley, CA (United States); Aguayo, E.; Fast, J. E.; Hoppe, E. W.; Kouzes, R. T.; LaFerriere, B. D.; Orrell, J. L.; Overman, N. R.; Soin, A. [Pacific Northwest National Laboratory, Richland, WA (United States); Avignone III, F. T. [Department of Physics and Astronomy, University of South Carolina, Columbia, SC, USA and Oak Ridge National Laboratory, Oak Ridge, TN (United States); Barabash, A. S.; Konovalov, S. I.; Yumatov, V. [Institute for Theoretical and Experimental Physics, Moscow (Russian Federation); Bertrand, F. E. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); and others

2014-06-24T23:59:59.000Z

404

CALDERON COKEMAKING PROCESS/DEMONSTRATION PROJECT  

SciTech Connect (OSTI)

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; (4) conducting a blast furnace test to demonstrate the compatibility of the coke produced; and (5) 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: Detailed workings of the team; Proposal to FETC for Phase II; Permitting and Environmental Work; and Engineering Progress.

Albert Calderon

1999-09-22T23:59:59.000Z

405

Electrometallurgical treatment demonstration at ANL-West  

SciTech Connect (OSTI)

Electrometallurgical treatment (EMT) was developed by Argonne National Laboratory (ANL) to ready sodium-bonded spent nuclear fuel for geological disposal. A demonstration of this technology was successfully completed in August 1999. EMT was used to condition irradiated EBR-II driver and blanket fuel at ANL-West. The results of this demonstration, including the production of radioactive high-level waste forms, are presented.

Goff, K. M.; Benedict, R. W.; Johnson, S. G.; Mariani, R. D.; Simpson, M. F.; Westphal, B. R.

2000-03-20T23:59:59.000Z

406

Computational fluid dynamic modeling of fluidized-bed polymerization reactors  

SciTech Connect (OSTI)

Polyethylene is one of the most widely used plastics, and over 60 million tons are produced worldwide every year. Polyethylene is obtained by the catalytic polymerization of ethylene in gas and liquid phase reactors. The gas phase processes are more advantageous, and use fluidized-bed reactors for production of polyethylene. Since they operate so close to the melting point of the polymer, agglomeration is an operational concern in all slurry and gas polymerization processes. Electrostatics and hot spot formation are the main factors that contribute to agglomeration in gas-phase processes. Electrostatic charges in gas phase polymerization fluidized bed reactors are known to influence the bed hydrodynamics, particle elutriation, bubble size, bubble shape etc. Accumulation of electrostatic charges in the fluidized-bed can lead to operational issues. In this work a first-principles electrostatic model is developed and coupled with a multi-fluid computational fluid dynamic (CFD) model to understand the effect of electrostatics on the dynamics of a fluidized-bed. The multi-fluid CFD model for gas-particle flow is based on the kinetic theory of granular flows closures. The electrostatic model is developed based on a fixed, size-dependent charge for each type of particle (catalyst, polymer, polymer fines) phase. The combined CFD model is first verified using simple test cases, validated with experiments and applied to a pilot-scale polymerization fluidized-bed reactor. The CFD model reproduced qualitative trends in particle segregation and entrainment due to electrostatic charges observed in experiments. For the scale up of fluidized bed reactor, filtered models are developed and implemented on pilot scale reactor.

Rokkam, Ram [Ames Laboratory

2012-11-02T23:59:59.000Z

407

Guidance manual for conducting technology demonstration activities  

SciTech Connect (OSTI)

This demonstration guidance manual has been prepared to assist Martin Marietta Energy Systems, Inc. (Energy Systems), staff in conducting demonstrations. It is prepared in checklist style to facilitate its use and assumes that Energy Systems personnel have project management responsibility. In addition to a detailed step-by-step listing of procedural considerations, a general checklist, logic flow diagram, and several examples of necessary plans are included to assist the user in developing an understanding of the many complex activities required to manage technology demonstrations. Demonstrations are pilot-scale applications of often innovative technologies to determine the commercial viability of the technologies to perform their designed function. Demonstrations are generally conducted on well-defined problems for which existing technologies or processes are less than satisfactory in terms of effectiveness, cost, and/or regulatory compliance. Critically important issues in demonstration management include, but are not limited to, such factors as communications with line and matrix management and with the US Department of Energy (DOE) and Energy Systems staff responsible for management oversight, budgetary and schedule requirements, regulatory compliance, and safety.

Jolley, R.L.; Morris, M.I.; Singh, S.P.N.

1991-12-01T23:59:59.000Z

408

What is the Federal Demonstration Project  

SciTech Connect (OSTI)

The Federal Demonstration Project is a cooperative effort between a number of universities, a private research institute, and several federal agencies to increase research productivity by eliminating unnecessary administrative procedures and by streamlining and standardizing needed controls. The Project aims to locate responsibility for decision-making as close as possible to principal investigators while maintaining necessary institutional and agency oversight to ensure accountability. By freeing researchers from some of their paperwork burden, more efficient research administration systems will enable investigators to spend more of their time doing science and engineering. The Federal Demonstration Project is an outgrowth of an earlier activity sponsored by five major federal R D agencies at the Florida State University System and the University of Miami. In Florida, the focus was on standardizing and streamlining procedures for administering research grants after the grants had been awarded to the universities. (See Attachment 1 for descriptions of the demonstrations carried out under the Florida Demonstration Project). In May 1988, the most successful of the demonstrated procedures were approved by the US Office of Management and Budget for use in grants awarded by any federal agency to any research organization. The new procedures give agencies authority to waive requirements that grantees obtain federal approval prior to taking a number of administrative actions with respect to grant management. The FDP institutions together with the participating federal agencies are designing and demonstrating innovative research administration procedures and are assessing the impact of those new procedures.

Not Available

1990-01-01T23:59:59.000Z

409

Reactor control rod timing system  

SciTech Connect (OSTI)

A fluid driven jet-edge whistle timing system for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (Above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

Wu, P.T.

1982-02-09T23:59:59.000Z

410

Reactor control rod timing system  

DOE Patents [OSTI]

A fluid driven jet-edge whistle timing system for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

Wu, Peter T. K. (Clifton Park, NY)

1982-01-01T23:59:59.000Z

411

Completing the Design of the Advanced Gas Reactor Fuel Development and Qualification Experiments for Irradiation in the Advanced Test Reactor  

SciTech Connect (OSTI)

The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

S. Blaine Grover

2006-10-01T23:59:59.000Z

412

DOE's way-out reactors  

SciTech Connect (OSTI)

The SP-100 reactor, envisioned long before Star Wars, was to power civilian structures such as the space station and orbiting commercial labs. According to the SDI Organization, it will be the cornerstone for SDI, used as a no-maintenance, general source of energy for the military's infrastructure - weapons scale power will come later. DOE wants to spend $72 in FY 1977 to design and build these reactors. Funding problems with Congress, as well as some of the technology and timetables are discussed here.

Marshall, E.

1986-03-21T23:59:59.000Z

413

Nuclear reactor safety heat transfer  

SciTech Connect (OSTI)

Reviewed is a book which has 5 parts: Overview, Fundamental Concepts, Design Basis Accident-Light Water Reactors (LWRs), Design Basis Accident-Liquid-Metal Fast Breeder Reactors (LMFBRs), and Special Topics. It combines a historical overview, textbook material, handbook information, and the editor's personal philosophy on safety of nuclear power plants. Topics include thermal-hydraulic considerations; transient response of LWRs and LMFBRs following initiating events; various accident scenarios; single- and two-phase flow; single- and two-phase heat transfer; nuclear systems safety modeling; startup and shutdown; transient response during normal and upset conditions; vapor explosions, natural convection cooling; blockages in LMFBR subassemblies; sodium boiling; and Three Mile Island.

Jones, O.C.

1982-07-01T23:59:59.000Z

414

Buried Waste Integrated Demonstration Plan. Revision 1  

SciTech Connect (OSTI)

This document presents the plan of activities for the Buried Waste Integrated Demonstration (BWID) program which supports the environmental restoration (ER) objectives of the Department of Energy (DOE) Complex. Discussed in this plan are the objectives, organization, roles and responsibilities, and the process for implementing and managing BWID. BWID is hosted at the Idaho National Engineering Laboratory (INEL), but involves participants from throughout the DOE Complex, private industry, universities, and the international community. These participants will support, demonstrate, and evaluate a suite of advanced technologies representing a comprehensive remediation system for the effective and efficient remediation of buried waste. The processes for identifying technological needs, screening candidate technologies for applicability and maturity, selecting appropriate technologies for demonstration, field demonstrating, evaluation of results and transferring technologies to environmental restoration programs are also presented. This document further describes the elements of project planning and control that apply to BWID. It addresses the management processes, operating procedures, programmatic and technical objectives, and schedules. Key functions in support of each demonstration such as regulatory coordination, safety analyses, risk evaluations, facility requirements, and data management are presented.

Kostelnik, K.M.

1991-12-01T23:59:59.000Z

415

Uncertainties in the Anti-neutrino Production at Nuclear Reactors  

E-Print Network [OSTI]

reactors are determined from thermal power measure- ments and ?ssion rate calculations.of a reactor’s ther- mal power is given by a calculation ofCALCULATIONS During the power cycle of a nuclear reactor,

Djurcic, Zelimir

2009-01-01T23:59:59.000Z

416

Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation  

E-Print Network [OSTI]

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

Heidet, Florent

2010-01-01T23:59:59.000Z

417

Computational Analysis of Fluid Flow in Pebble Bed Modular Reactor  

E-Print Network [OSTI]

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

Gandhir, Akshay

2012-10-19T23:59:59.000Z

418

Reactivity control assembly for nuclear reactor  

DOE Patents [OSTI]

Reactivity control assembly for nuclear reactor comprises supports stacked above reactor core for holding control rods. Couplers associated with the supports and a vertically movable drive shaft have lugs at their lower ends for engagement with the supports.

Bollinger, Lawrence R. (Schenectady, NY)

1984-01-01T23:59:59.000Z

419

Reactor accelerator coupling experiments: a feasability study  

E-Print Network [OSTI]

The Reactor Accelerator Coupling Experiments (RACE) are a set of neutron source driven subcritical experiments under temperature feedback conditions. These experiments will involve coupling an accelerator driven neutron source to a TRIGA reactor...

Woddi Venkat Krishna, Taraknath

2006-08-16T23:59:59.000Z

420

New fast-reactor approach. [LMFBR  

SciTech Connect (OSTI)

The design parameters for a 1000 MW LMFBR type reactor are presented. The design requires the multiple primary coolant pumps and heat exchangers to be located around the core within the reactor vessel.

Folkrod, J.R.; Kann, W.J.; Klocksieben, R.H.

1983-01-01T23:59:59.000Z

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


421

Stability analysis of supercritical water cooled reactors  

E-Print Network [OSTI]

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

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

2005-01-01T23:59:59.000Z

422

Multi-Applications Small Light Water Reactor - NERI Final Report  

SciTech Connect (OSTI)

The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle.

S. Michale Modro; James E. Fisher; Kevan D. Weaver; Jose N. Reyes, Jr.; John T. Groome; Pierre Babka; Thomas M. Carlson

2003-12-01T23:59:59.000Z

423

Johnson Noise Thermometry for Advanced Small Modular Reactors  

SciTech Connect (OSTI)

Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor s physical condition. In and near core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurement due to their fundamental natures. Small, Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of this project is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.

Britton Jr, Charles L [ORNL; Roberts, Michael [ORNL; Bull, Nora D [ORNL; Holcomb, David Eugene [ORNL; Wood, Richard Thomas [ORNL

2012-10-01T23:59:59.000Z

424

Johnson Noise Thermometry for Advanced Small Modular Reactors  

SciTech Connect (OSTI)

Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor’s physical condition. In and near the core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurement due to their fundamental natures. Small Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of the current ORNL-led project, conducted under the Instrumentation, Controls, and Human-Machine Interface (ICHMI) research pathway of the U.S. Department of Energy (DOE) Advanced SMR Research and Development (R&D) program, is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.

Britton, C.L.,Jr.; Roberts, M.; Bull, N.D.; Holcomb, D.E.; Wood, R.T.

2012-09-15T23:59:59.000Z

425

Application of the ''reactivity constraint approach'' to automatic reactor control  

SciTech Connect (OSTI)

The ''reactivity constraint approach'' is described and demonstrated to be an effective and reliable means for the automatic control of power in nuclear reactors. This approach functions by restricting the effect of the delayed neutron populations to that which can be balanced by an induced change in the prompt population. This is done by limiting the net reactivity to the amount that can be offset by reversing the direction of motion of the automated control mechanism. The necessary reactivity constraints are obtained from the dynamic period equation, which gives the instantaneous reactor period as a function of the reactivity and the rate of change of reactivity. The derivation of this equation is described with emphasis on the recently obtained ''alternate'' formulation. Following a discussion of the behavior of each term of this alternate equation as a function of reactivity, its use in the design and operation of a nonlinear, closed-loop, digital controller for reactor power is in the design and operation of a nonlinear, closed-loop, digital controller for reactor power is described. Details of the initial experimental trials of the resulting controller are given.

Bernard, J.A.; Henry, A.F.; Lanning, D.D.

1988-02-01T23:59:59.000Z

426

Recent Sodium Technology Development for the Decommissioning of the Rapsodie and Superphenix Reactors and the Management of Sodium Wastes  

SciTech Connect (OSTI)

The Commissariat a l'Energie Atomique (CEA) has recently developed and/or conducted experiments on several processes in support of the decommissioning of two French liquid-metal fast reactors (LMFRs), Rapsodie and Superphenix, as well as on the treatment of CEA sodium wastes. CEA has demonstrated that it is possible to define appropriate and efficient processes to meet the different situations encountered in decommissioning LMFRs. Mechanical techniques derived from standard technologies have been successfully applied to fast reactor decommissioning to complete primary vessel draining from sodium. In addition, specific chemical processes have been developed to deal safely with metallic sodium reactivity. Sodium-contaminated equipment has been successfully cleaned by reacting sodium with water mist in an atmosphere with carbon dioxide to form inert sodium carbonate. Bulk sodium has been successfully converted into aqueous caustic soda by injection of liquid-metallic sodium into sodium hydroxide solution. Several processes were also defined to deal with specific sodium wastes. In all cases the principle is based on a sodium/water chemical reaction where the released hydrogen and heat are controlled. With the development of a wide variety of processes, all steps in the decommissioning of LMFRs are assumed to be now properly mastered.

Rodriguez, G.; Gastaldi, O.; Baque, F. [Commissariat a l'Energie Atomique Cadarache (France)

2005-04-15T23:59:59.000Z

427

Digital computer operation of a nuclear reactor  

DOE Patents [OSTI]

A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

Colley, Robert W. (Richland, WA)

1984-01-01T23:59:59.000Z

428

Digital computer operation of a nuclear reactor  

DOE Patents [OSTI]

A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

Colley, R.W.

1982-06-29T23:59:59.000Z

429

Liquid metal cooled nuclear reactor plant system  

DOE Patents [OSTI]

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

1993-01-01T23:59:59.000Z

430

Westinghouse Advanced Reactors Division Plutonium Fuel Laboratories  

Office of Legacy Management (LM)

Radiological Condition of the Westinghouse Advanced Reactors Division Plutonium Fuel Laboratories Cheswick, Pennsylvania -. -, -- AGENCY: Office of Operational Safety, Department...

431

Light Water Reactor Sustainability (LWRS) Program  

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

Light Water Reactor Sustainability (LWRS) Program Login Instructions go here. User ID: Password: Log In Forgot your password?...

432

How far is a Fusion Power Reactor from an Experimental Reactor?  

E-Print Network [OSTI]

be able to move directly and safely to a "first of a kind" reactor. The main conditions to be satisfied / experimental evidence. To assess the reactor relevance of ITER, rather than a comparison between ITER and one1 How far is a Fusion Power Reactor from an Experimental Reactor? R. Toschi(1) , P. Barabaschi(2

433

Eddy current NDE performance demonstrations using simulation tools  

SciTech Connect (OSTI)

To carry out performance demonstrations of the Eddy-Current NDE processes applied on French nuclear power plants, EDF studies the possibility of using simulation tools as an alternative to measurements on steam generator tube mocks-up. This paper focuses on the strategy led by EDF to assess and use code{sub C}armel3D and Civa, on the case of Eddy-Current NDE on wears problem which may appear in the U-shape region of steam generator tubes due to the rubbing of anti-vibration bars.

Maurice, L. [EDF - CEIDRE, 2 rue Ampere, 93206 Saint-Denis Cedex 1 (France); Costan, V.; Guillot, E.; Thomas, P. [EDF - R and D, THEMIS, 1, avenue du General de Gaulle, 92141 Clamart (France)

2013-01-25T23:59:59.000Z

434

Indication of anomalous heat energy production in a reactor device  

E-Print Network [OSTI]

An experimental investigation of possible anomalous heat production in a special type of reactor tube named E-Cat HT is carried out. The reactor tube is charged with a small amount of hydrogen loaded nickel powder plus some additives. The reaction is primarily initiated by heat from resistor coils inside the reactor tube. Measurement of the produced heat was performed with high-resolution thermal imaging cameras, recording data every second from the hot reactor tube. The measurements of electrical power input were performed with a large bandwidth three-phase power analyzer. Data were collected in two experimental runs lasting 96 and 116 hours, respectively. An anomalous heat production was indicated in both experiments. The 116-hour experiment also included a calibration of the experimental set-up without the active charge present in the E-Cat HT. In this case, no extra heat was generated beyond the expected heat from the electric input. Computed volumetric and gravimetric energy densities were found to be far above those of any known chemical source. Even by the most conservative assumptions as to the errors in the measurements, the result is still one order of magnitude greater than conventional energy sources.

Giuseppe Levi; Evelyn Foschi; Torbjörn Hartman; Bo Höistad; Roland Pettersson; Lars Tegnér; Hanno Essén

2013-06-07T23:59:59.000Z

435

Indication of anomalous heat energy production in a reactor device  

E-Print Network [OSTI]

An experimental investigation of possible anomalous heat production in a special type of reactor tube named E-Cat HT is carried out. The reactor tube is charged with a small amount of hydrogen loaded nickel powder plus some additives. The reaction is primarily initiated by heat from resistor coils inside the reactor tube. Measurement of the produced heat was performed with high-resolution thermal imaging cameras, recording data every second from the hot reactor tube. The measurements of electrical power input were performed with a large bandwidth three-phase power analyzer. Data were collected in two experimental runs lasting 96 and 116 hours, respectively. An anomalous heat production was indicated in both experiments. The 116-hour experiment also included a calibration of the experimental set-up without the active charge present in the E-Cat HT. In this case, no extra heat was generated beyond the expected heat from the electric input. Computed volumetric and gravimetric energy densities were found to be fa...

Levi, Giuseppe; Hartman, Torbjörn; Höistad, Bo; Pettersson, Roland; Tegnér, Lars; Essén, Hanno

2013-01-01T23:59:59.000Z

436

Development and Demonstration of Ultrafiltration Simulants  

SciTech Connect (OSTI)

According to Bechtel National, Inc. (BNI) Test Specification 24590-PTF-TSP-RT-06-006, Rev 0, Simulant Development to Support the Development and Demonstration of Leaching and Ultrafiltration Pretreatment Processes,” simulants for boehmite, gibbsite, and filtration are to be developed that can be used in subsequent bench and integrated testing of the leaching/filtration processes for the waste treatment plant (WTP). These simulants will then be used to demonstrate the leaching process and to help refine processing conditions which may impact safety basis considerations (Smith 2006). This report documents the results of the filtration simulant development.

Russell, Renee L.; Billing, Justin M.; Peterson, Reid A.; Rinehart, Donald E.; Smith, Harry D.

2009-02-24T23:59:59.000Z

437

Experimental demonstration of superdirective dielectric antenna  

SciTech Connect (OSTI)

We propose and demonstrate experimentally a simple approach for achieving superdirectivity of emitted radiation for electrically small antennas based on a spherical dielectric resonator with a notch excited by a dipole source. Superdirectivity is achieved without using complex antenna arrays and for a wide range of frequencies. We also demonstrate the steering effect for a subwavelength displacement of the source. Finally, unlike previously known superdirective antennas, our design has significantly smaller losses, at the operation frequency radiation efficiency attains 80%, and matching holds in the 3%-wide frequency band without any special matching technique.

Krasnok, Alexander E.; Filonov, Dmitry S.; Belov, Pavel A. [National Research University of Information Technologies, Mechanics and Optics (ITMO), St. Petersburg 197101 (Russian Federation); Simovski, Constantin R. [National Research University of Information Technologies, Mechanics and Optics (ITMO), St. Petersburg 197101 (Russian Federation); Aalto University, School of Electric and Electronic Engineering, Aalto FI76000 (Finland); Kivshar, Yuri S. [National Research University of Information Technologies, Mechanics and Optics (ITMO), St. Petersburg 197101 (Russian Federation); Nonlinear Physics Center, Research School of Physics and Engineering, Australian National University, Canberra ACT 0200 (Australia)

2014-03-31T23:59:59.000Z

438

JEA successfully completes world's largest CFB demonstration  

SciTech Connect (OSTI)

JEA (formerly the Jacksonville Electric Authority) has successfully completed an eighth year landmark demonstration project that continues in baseload commercial operation. It scales up atmospheric fluidized-bed technology demonstration to the near-300-MW size, providing important data on a technology that can achieve > 90% SO{sub 2} removal and 60% NOx reduction at relatively high efficiencies and at costs comparable to those of conventional pulverized coal plants. The article recounts the history of the project. Performance tests showed a blend of coal and petcoke were most efficient as a feedstock. 3 figs.

NONE

2005-09-30T23:59:59.000Z

439

Result Demonstration: A Method That Works  

E-Print Network [OSTI]

E-449 05-07 ?A producer doing a result demonstration on his or her personal property is the ultimate in education: Hands-on and remembered! Don?t stop and file it. You?ve educated one, now educate the masses! Let the world know about it. Put.... County Extension educators in San Patri- cio County, in cooperation with the local Field Crops Committee, implemented several in-depth demonstration/applied research and educa- tional programs to help the local producers more effectively manage...

Boleman, Chris; Dromgoole, Darrell A.

2007-05-24T23:59:59.000Z

440

Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation  

E-Print Network [OSTI]

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

Heidet, Florent

2010-01-01T23:59:59.000Z

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


441

MULTIPHASE REACTOR MODELING FOR ZINC CHLORIDE CATALYZED COAL LIQUEFACTION  

E-Print Network [OSTI]

for the Coal Slurry Reactor Calculations are shown here for= Total reactor pressure, psi. The calculation is iterative,

Joyce, Peter James

2011-01-01T23:59:59.000Z

442

D Ris-R-406 Department of Reactor  

E-Print Network [OSTI]

of a Nuclear District Heating Reactor ... 17 3. REACTOR PHYSICS AND DYNAMICS 13 3.1. Core Follow Studies: Bti

443

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

SciTech Connect (OSTI)

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

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

2011-05-01T23:59:59.000Z

444

Interdisciplinary Institute for Innovation Nuclear reactors' construction  

E-Print Network [OSTI]

Interdisciplinary Institute for Innovation Nuclear reactors' construction costs: The role of lead@mines-paristech.fr hal-00956292,version1-6Mar2014 #12;hal-00956292,version1-6Mar2014 #12;Nuclear reactors' construction reactor construction costs in France and the United States. Studying the cost of nuclear power has often

Paris-Sud XI, Université de

445

The Reactor An ObjectOriented Framework  

E-Print Network [OSTI]

The Reactor An Object­Oriented Framework for Event Demultiplexing and Event Handler Dispatching Douglas C. Schmidt 1 Overview ffl The Reactor is an object­oriented frame­ work that encapsulates OS event demul­ tiplexing mechanisms -- e.g., the Reactor API runs transparently atop both Wait

Schmidt, Douglas C.

446

International Journal of Chemical Reactor Engineering  

E-Print Network [OSTI]

International Journal of Chemical Reactor Engineering Volume 3 2005 Article A17 Optimal Operation, a single re- action takes place in the reactor and the operational objective is to compute the optimal feed is illustrated via simulation of two semi-batch reactor applications. KEYWORDS: Dynamic Optimization, Batch

Palanki, Srinivas

447

Laminar Entrained Flow Reactor (Fact Sheet)  

SciTech Connect (OSTI)

The Laminar Entrained Flow Reactor (LEFR) is a modular, lab scale, single-user reactor for the study of catalytic fast pyrolysis (CFP). This system can be employed to study a variety of reactor conditions for both in situ and ex situ CFP.

Not Available

2014-02-01T23:59:59.000Z

448

Nozzle for electric dispersion reactor  

DOE Patents [OSTI]

A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 4 figs.

Sisson, W.G.; Basaran, O.A.; Harris, M.T.

1995-11-07T23:59:59.000Z

449

Nozzle for electric dispersion reactor  

DOE Patents [OSTI]

A nozzle for an electric dispersion reactor includes two coaxial cylindrical bodies, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 5 figs.

Sisson, W.G.; Harris, M.T.; Scott, T.C.; Basaran, O.A.

1998-06-02T23:59:59.000Z

450

Nozzle for electric dispersion reactor  

DOE Patents [OSTI]

A nozzle for an electric dispersion reactor includes two coaxial cylindrical bodies, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 5 figs.

Sisson, W.G.; Harris, M.T.; Scott, T.C.; Basaran, O.A.

1996-04-02T23:59:59.000Z

451

Nozzle for electric dispersion reactor  

DOE Patents [OSTI]

A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 4 figs.

Sisson, W.G.; Basaran, O.A.; Harris, M.T.

1998-04-14T23:59:59.000Z

452

Nozzle for electric dispersion reactor  

DOE Patents [OSTI]

A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode.

Sisson, Warren G. (Oak Ridge, TN); Basaran, Osman A. (Oak Ridge, TN); Harris, Michael T. (Knoxville, TN)

1995-01-01T23:59:59.000Z

453

Nozzle for electric dispersion reactor  

DOE Patents [OSTI]

A nozzle for an electric dispersion reactor includes two coaxial cylindrical bodies, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode.

Sisson, Warren G. (Oak Ridge, TN); Harris, Michael T. (Knoxville, TN); Scott, Timothy C. (Knoxville, TN); Basaran, Osman A. (Oak Ridge, TN)

1998-01-01T23:59:59.000Z

454

Nozzle for electric dispersion reactor  

DOE Patents [OSTI]

A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode.

Sisson, Warren G. (Oak Ridge, TN); Basaran, Osman A. (Oak Ridge, TN); Harris, Michael T. (Knoxville, TN)

1998-01-01T23:59:59.000Z

455

Nozzle for electric dispersion reactor  

DOE Patents [OSTI]

A nozzle for an electric dispersion reactor includes two coaxial cylindrical bodies, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode.

Sisson, Warren G. (Oak Ridge, TN); Harris, Michael T. (Knoxville, TN); Scott, Timothy C. (Knoxville, TN); Basaran, Osman A. (Oak Ridge, TN)

1996-01-01T23:59:59.000Z

456

Nuclear Reactor Safety Design Criteria  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The order establishes nuclear safety criteria applicable to the design, fabrication, construction, testing, and performance requirements of nuclear reactor facilities and safety class structures, systems, and components (SSCs) within these facilities. Cancels paragraphs 8a and 8b of DOE 5480.6. Cancels DOE O 5480.6 in part. Certified 11-18-10.

1993-01-19T23:59:59.000Z

457

Starfire - a commercial tokamak reactor  

SciTech Connect (OSTI)

The basic objective of the STARFIRE Project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. 10 refs.

Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Kokoszenski, J.; Graumann, D.

1981-01-01T23:59:59.000Z

458

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

SciTech Connect (OSTI)

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

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

2014-03-01T23:59:59.000Z

459

Reactor Start-up and Control Methodologies: Consideration of the Space Radiation Environment  

SciTech Connect (OSTI)

The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable the accomplishment of ambitious space exploration missions. The natural radiation environment in space provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Initial investigation using MCNPX 2.5.b for proton transport through the SAFE-400 reactor indicates a secondary neutron net current of 1.4x107 n/s at the core-reflector interface, with an incoming current of 3.4x106 n/s due to neutrons produced in the Be reflector alone. This neutron population could provide a reliable startup source for a space reactor. Additionally, this source must be considered in developing a reliable control strategy during reactor startup, steady-state operation, and power transients. An autonomous control system is developed and analyzed for application during reactor startup, accounting for fluctuations in the radiation environment that result from changes in vehicle location (altitude, latitude, position in solar system) or due to temporal variations in the radiation field, as may occur in the case of solar flares. One proposed application of a nuclear electric propulsion vehicle is in a tour of the Jovian system, where the time required for communication to Earth is significant. Hence, it is important that a reactor control system be designed with feedback mechanisms to automatically adjust to changes in reactor temperatures, power levels, etc., maintaining nominal operation without user intervention. This paper will evaluate the potential use of secondary neutrons produced by proton interactions in the reactor vessel as a startup source for a space reactor and will present a potential control methodology for reactor startup procedures in the event of source fluctuations.

Bragg-Sitton, Shannon M.; Holloway, James Paul [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States)

2004-02-04T23:59:59.000Z

460

Device for actuating a nuclear reactor control rod  

SciTech Connect (OSTI)

A first assembly for low-speed translational displacement and dropping of a control rod comprises a leak-tight casing tube which communicates with the reactor vessel, a screw and nut system coupled to the control rod by means of the nut and a screw -retaining device which is capable of releasing the screw. In the event of accidental failure of the casing tube, upward displacement of the control rod and of the broken portion of the casing tube is limited by a second assembly comprising means for rotational locking of the screw drive rotor and calibrated elastic means for producing action on the rotor in opposition to the vertical thrust produced by failure of the casing tube.

Etienne, C.; Fardouet, C.; L'abbe, G.M.

1982-02-09T23:59:59.000Z

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


461

Nuclear reactor safeguards and monitoring with antineutrino detectors A. Bernsteina)  

E-Print Network [OSTI]

Nuclear reactor safeguards and monitoring with antineutrino detectors A. Bernsteina) Sandia of nuclear reactor types, including power reactors, research reactors, and plutonium production reactors-understood principles that govern the core's evolution in time, can be used to determine whether the reactor is being

Gratta, Giorgio

462

Title: Robot Arm Demonstration Supervisor: Alexander Frster  

E-Print Network [OSTI]

UROP 2009 Title: Robot Arm Demonstration Supervisor: Alexander Förster Sponsor: Luca Gambardella / UROP Keywords: programming language C++, Qt-framework, Katana robot arm, computer vision Prerequisites based controlled robot arm. The computer vision system has to detect the position and orientation

Krause, Rolf

463

THE REQUIREMENTS OF A FUSION DEMONSTRATION  

E-Print Network [OSTI]

and will analyze the strategies, facilities, and prior testing requirements needed to demonstrate electric power Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes constitute or imply its endorsement, recommendation, or favoring by the United States Government or any

Najmabadi, Farrokh

464

Energy Harvesting Communication Networks: Optimization and Demonstration  

E-Print Network [OSTI]

) and the UK (Imperial College London). Index Terms--energy harvesting; energy packet net- works; Markov models harvesting devices. EH capability can scavenge ambient energy, such as vibrations, thermal gradients or solar1 Energy Harvesting Communication Networks: Optimization and Demonstration (The E-CROPS Project

Gesbert, David

465

CubeSat deformable mirror demonstration  

E-Print Network [OSTI]

The goal of the CubeSat Deformable Mirror Demonstration (DeMi) is to characterize the performance of a small deformable mirror over a year in low-Earth orbit. Small form factor deformable mirrors are a key technology needed ...

Cahoy, Kerri

466

CALDERON COKEMAKING PROCESS/DEMONSTRATION PROJECT  

SciTech Connect (OSTI)

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; and (v) demonstrating that coke can be produced economically, at a level competitive with coke imports. The activities of the past quarter continued to be focused on the following: Concluding the Negotiation and completing Contracts among Stakeholders of the Team; Revision of Final Report for Phase I; Engineering Design Progress; Selection of Systems Associates, Inc. for design of Control System; Conclusion of Secrecy Agreement with Carborundum (St. Gobain); and Permitting Work and Revisions.

Albert Calderon

2000-03-22T23:59:59.000Z

467

Calderon cokemaking process/demonstration project  

SciTech Connect (OSTI)

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 a 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; (4) conducting a blast furnace test to demonstrate the compatibility of the coke produced; and (5) demonstrating that coke can be produced economically, at a level competitive with coke imports. The activities of the past quarter continued to be focused on the following: Drafting of Contracts among the Stakeholders of the Team, Completion and Delivery of Proposal for Phase 2 Permitting and Environmental Work Engineering Progress Preparation of Final Report for Phase 1 DCAA Audit Funding for Phase 2.

Albert Calderon

1999-12-21T23:59:59.000Z

468

CALDERON COKEMAKING PROCESS/DEMONSTRATION PROJECT  

SciTech Connect (OSTI)

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; and (v) demonstrating that coke can be produced economically, at a level competitive with coke imports. The activities of the past quarter continued to be focused on the following: Concluding the Negotiation and completing Contracts among Stakeholders of the Team; Revision of Final Report for Phase I; Engineering Design Progress; Selection of Systems Associates, Inc. for design of Control System; Conclusion of Secrecy Agreement with Carborundum (St. Gobain); and Permitting Work and Revisions.

Albert Calderon

2000-06-21T23:59:59.000Z

469

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

SciTech Connect (OSTI)

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

Not Available

1980-10-09T23:59:59.000Z

470

Acoustic emission monitoring of hot functional testing: Watts Bar Unit 1 Nuclear Reactor  

SciTech Connect (OSTI)

Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar, Unit 1 Nuclear Power Plant during hot functional preservice testing is described in this report. The report deals with background, methodology, and results. The work discussed here is a major milestone in a program supported by NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. The subject work demonstrated that anticipated problem areas can be overcome. Work is continuing toward AE monitoring during reactor operation.

Hutton, P.H.; Dawson, J.F.; Friesel, M.A.; Harris, J.C.; Pappas, R.A.

1984-06-01T23:59:59.000Z

471

Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants  

SciTech Connect (OSTI)

Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

Not Available

1993-05-13T23:59:59.000Z

472

Spectral shift reactor control method  

SciTech Connect (OSTI)

This patent describes the method of operating a pressurized-water fissile-material-fueled spectral-shift nuclear reactor in such manner that short-term reactivity requirement variations can be satisfied without making control rod or chemical shim changes. The reactor includes a pressure vessel enclosing a reactor core and having an inlet and an outlet for circulating a water coolant moderator in heat transfer relationship with the core. The core comprises fuel assemblies disposed therein for generating heat by nuclear fission. The reactor provided with neutron-absorbing control rods which are vertically movable into and out of the core so that movement of the control rods into the core will substantially decrease reactivity and withdrawal of the control rods from the core will substantially increase reactivity. The control rods when inserted into the core displace an equivalent volume of the water coolant moderator. The reactor also provides neutron-spectral-shift rods which have a lower absorptivity for neutrons than the control rods, the neutron-spectral shift rods when inserted into the core displacing an equaivalent volume of the water coolant moderator. The neutron-spectral-shift rods comprises two different types of rods, a first of the different types of the neutron-spectral-shift rods comprising displacer rods which have a low absorptivity for neutrons, the remainder of the neutron-spectral-shift rods comprising gray rods which have an absorption for neutrons which is intermediate the neutron absorption of the control rods and the low neutron absorption of the displacer rods. Each neutron-spectral-shift displacer rod comprises a hollow thin-walled Zircaloy member containing a filling of solid or annular zirconium- or aluminum-containing material for providing internal support and mass for the thin-walled tubular member.

Impink, A.J. Jr.

1987-12-29T23:59:59.000Z

473

Heterogeneous Recycling in Fast Reactors  

SciTech Connect (OSTI)

Current sodium fast reactor (SFR) designs have avoided the use of depleted uranium blankets over concerns of creating weapons grade plutonium. While reducing proliferation risks, this restrains the reactor design space considerably. This project will analyze various blanket and transmutation target configurations that could broaden the design space while still addressing the non-proliferation issues. The blanket designs will be assessed based on the transmutation efficiency of key minor actinide (MA) isotopes and also on mitigation of associated proliferation risks. This study will also evaluate SFR core performance under different scenarios in which depleted uranium blankets are modified to include minor actinides with or without moderators (e.g. BeO, MgO, B4C, and hydrides). This will be done in an effort to increase the sustainability of the reactor and increase its power density while still offering a proliferation resistant design with the capability of burning MA waste produced from light water reactors (LWRs). Researchers will also analyze the use of recycled (as opposed to depleted) uranium in the blankets. The various designs will compare MA transmutation efficiency, plutonium breeding characteristics, proliferation risk, shutdown margins and reactivity coefficients with a current reference sodium fast reactor design employing homogeneous recycling. The team will also evaluate the out-of-core accumulation and/or burn-down rates of MAs and plutonium isotopes on a cycle-by-cycle basis. This cycle-by-cycle information will be produced in a format readily usable by the fuel cycle systems analysis code, VISION, for assessment of the sustainability of the deployment scenarios.

Dr. Benoit Forget; Michael Pope; Piet, Steven J.; Michael Driscoll

2012-07-30T23:59:59.000Z

474

Control of reactor coolant flow path during reactor decay heat removal  

DOE Patents [OSTI]

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

Hunsbedt, Anstein N. (Los Gatos, CA)

1988-01-01T23:59:59.000Z

475

Newberry Volcano EGS Demonstration - Phase I Results  

SciTech Connect (OSTI)

Phase I of the Newberry Volcano Enhanced Geothermal System (EGS) Demonstration included permitting, community outreach, seismic hazards analysis, initial microseismic array deployment and calibration, final MSA design, site characterization, and stimulation planning. The multi-disciplinary Phase I site characterization supports stimulation planning and regulatory permitting, as well as addressing public concerns including water usage and induced seismicity. A review of the project'Ă?Â?Ă?Â?s water usage plan by an independent hydrology consultant found no expected impacts to local stakeholders, and recommended additional monitoring procedures. The IEA Protocol for Induced Seismicity Associated with Enhanced Geothermal Systems was applied to assess site conditions, properly inform stakeholders, and develop a comprehensive mitigation plan. Analysis of precision LiDAR elevation maps has concluded that there is no evidence of recent faulting near the target well. A borehole televiewer image log of the well bore revealed over three hundred fractures and predicted stress orientations. No natural, background seismicity has been identified in a review of historic data, or in more than seven months of seismic data recorded on an array of seven seismometers operating around the target well. A seismic hazards and induced seismicity risk assessment by an independent consultant concluded that the Demonstration would contribute no additional risk to residents of the nearest town of La Pine, Oregon. In Phase II of the demonstration, an existing deep hot well, NWG 55-29, will be stimulated using hydroshearing techniques to create an EGS reservoir. The Newberry Volcano EGS Demonstration is allowing geothermal industry and academic experts to develop, validate and enhance geoscience and engineering techniques, and other procedures essential to the expansion of EGS throughout the country. Successful development will demonstrate to the American public that EGS can play a significant role in reducing foreign energy dependence, and provide clean, renewable, baseload geothermal power generation in the State of Oregon.

William L. Osborn, Susan Petty, Trenton T. Cladouhos, Joe Iovenitti, Laura Nofziger, Owen Callahan, Douglas S. Perry and Paul L. Stern

2011-10-23T23:59:59.000Z

476

MYRRHA a multi-purpose hybrid research reactor for high-tech applications  

SciTech Connect (OSTI)

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is the flexible experimental accelerator driven system (ADS) in development at SCK-CEN. MYRRHA is able to work both in subcritical (ADS) as in critical mode. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for generation IV (GEN IV) systems, material developments for fusion reactors, radioisotope production and industrial applications, such as Si-doping. MYRRHA will also demonstrate the ADS full concept by coupling the three components (accelerator, spallation target and subcritical reactor) at reasonable power level to allow operation feedback, scalable to an industrial demonstrator and allow the study of efficient transmutation of high-level nuclear waste. MYRRHA is based on the heavy liquid metal technology and so it will contribute to the development of lead fast reactor (LFR) technology and in critical mode, MYRRHA will play the role of European technology pilot plant in the roadmap for LFR. In this paper the historical evolution of MYRRHA and the rationale behind the design choices is presented and the latest configuration of the reactor core and primary system is described. (authors)

Abderrahim, H. A.; Baeten, P. [SCK CEN, Boeretang 200, 2400 Mol (Belgium)

2012-07-01T23:59:59.000Z

477

Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach  

SciTech Connect (OSTI)

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

David Petti; Jim Kinsey; Dave Alberstein

2014-01-01T23:59:59.000Z

478

Heat pipe cooled reactors for multi-kilowatt space power supplies  

SciTech Connect (OSTI)

Three nuclear reactor space power system designs are described that demonstrate how the use of high temperature heat pipes for reactor heat transport, combined with direct conversion of heat to electricity, can result in eliminating pumped heat transport loops for both primary reactor cooling and heat rejection. The result is a significant reduction in system complexity that leads to very low mass systems with high reliability, especially in the power range of 1 to 20 kWe. In addition to removing heat exchangers, electromagnetic pumps, and coolant expansion chambers, the heat pipe/direct conversion combination provides such capabilities as startup from the frozen state, automatic rejection of reactor decay heat in the event of emergency or accidental reactor shutdown, and the elimination of single point failures in the reactor cooling system. The power system designs described include a thermoelectric system that can produce 1 to 2 kWe, a bimodal modification of this system to increase its power level to 5 kWe and incorporate high temperature hydrogen propulsion capability, and a moderated thermionic reactor concept with 5 to 20 kWe power output that is based on beryllium modules that thermally couple cylindrical thermionic fuel elements (TFEs) to radiator heat pipes.

Ranken, W.A.; Houts, M.G.

1995-01-01T23:59:59.000Z

479

Fast-acting nuclear reactor control device  

DOE Patents [OSTI]

A fast-acting nuclear reactor control device for moving and positioning a fety control rod to desired positions within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump/motor, an electric gear motor, and solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

Kotlyar, Oleg M. (Idaho Falls, ID); West, Phillip B. (Idaho Falls, ID)

1993-01-01T23:59:59.000Z

480

Shutdown system for a nuclear reactor  

DOE Patents [OSTI]

An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion. 8 figs.

Groh, E.F.; Olson, A.P.; Wade, D.C.; Robinson, B.W.

1984-06-05T23:59:59.000Z

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