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1

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

SciTech Connect

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

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

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

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 ...

Christensen, Eric; Jaffke, Patrick

2013-01-01T23:59:59.000Z

6

FINAL SIMULATION RESULTS FOR DEMONSTRATION CASE 1 AND 2  

SciTech Connect

The goal of this DOE Vision-21 project work scope was to develop an integrated suite of software tools that could be used to simulate and visualize advanced plant concepts. Existing process simulation software did not meet the DOE's objective of ''virtual simulation'' which was needed to evaluate complex cycles. The overall intent of the DOE was to improve predictive tools for cycle analysis, and to improve the component models that are used in turn to simulate equipment in the cycle. Advanced component models are available; however, a generic coupling capability that would link the advanced component models to the cycle simulation software remained to be developed. In the current project, the coupling of the cycle analysis and cycle component simulation software was based on an existing suite of programs. The challenge was to develop a general-purpose software and communications link between the cycle analysis software Aspen Plus{reg_sign} (marketed by Aspen Technology, Inc.), and specialized component modeling packages, as exemplified by industrial proprietary codes (utilized by ALSTOM Power Inc.) and the FLUENT{reg_sign} computational fluid dynamics (CFD) code (provided by Fluent Inc). A software interface and controller, based on an open CAPE-OPEN standard, has been developed and extensively tested. Various test runs and demonstration cases have been utilized to confirm the viability and reliability of the software. ALSTOM Power was tasked with the responsibility to select and run two demonstration cases to test the software--(1) a conventional steam cycle (designated as Demonstration Case 1), and (2) a combined cycle test case (designated as Demonstration Case 2). Demonstration Case 1 is a 30 MWe coal-fired power plant for municipal electricity generation, while Demonstration Case 2 is a 270 MWe, natural gas-fired, combined cycle power plant. Sufficient data was available from the operation of both power plants to complete the cycle configurations. Three runs were completed for each Demonstration Case--(1) an initial baseline run using the existing component libraries in Aspen Plus{reg_sign}, (2) a second run where one of the library components was replaced with an ALSTOM Power proprietary code, and (3) a third run where a cycle component was replaced with a FLUENT{reg_sign} CFD simulation. Each of the three runs was successfully completed over a range of loads. This report documents the case runs and discusses the viability and capabilities of the linkage/interface software.

David Sloan; Woodrow Fiveland

2003-10-15T23:59:59.000Z

7

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

SciTech Connect

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

8

BWR (boiling-water reactor) radiation control: In-plant demonstration at Vermont Yankee: Final report  

SciTech Connect

Results of the RP1934 program, which was established by EPRI in 1981 to demonstrate the adequacy of BRAC program (RP819) principles for BWR radiation control at Vermont Yankee, are presented. Evaluations were performed of the effectiveness of optimization of purification system performance, control of feedwater dissolved oxygen concentrations, minimization of corrosion product and ionic transport, and improved startup, shutdown, and layup practices. The impact on shutdown radiation levels of these corrective actions was assessed based on extensive primary system radiation survey and component gamma scan data. Implementation of the BRAC recommendations was found to be insufficient to reduce the rate of activity buildup on out-of-core surfaces at Vermont Yankee, and additional corrective actions were found necessary. Specifically, replacement of cobalt-bearing materials in the control rod drive pins and rollers and feedwater regulating valves was pursued as was installation of electropolished 316 stainless steel during a recirculation piping replacement program. Aggressive programs to further reduce copper concentrations in the reactor water by improving condensate demineralizer efficiency and to minimize organic ingress to the power cycle by reducing organic concentrations in recycled radwaste also were undertaken. Evaluations of the impact on activity buildup of several pretreatment processes including prefilming in moist air, preexposure to high temperature water containing zinc, and electropolishing also were performed in a test loop installed in the reactor water cleanup system. A significant beneficial impact of electropolishing was shown to be present for periods up to 6000 hours.

Palino, G.F.; Hobart, R.L.; Sawochka, S.G.

1987-10-01T23:59:59.000Z

9

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

Energy.gov (U.S. Department of Energy (DOE))

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

10

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

SciTech Connect

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

11

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

E-Print Network (OSTI)

of the beam port door and exposing detector at low reactor power. In order to permit operation 26 continously with minimal disruption of normal reactor operations and minimal radiation exposure to operating personnel, an object- detector holding apparatus... exposure except from the gamma rays in the beam. Righ quality radiographs were 29 obtained using Type R film with exposures of 20-30 minutes at a reactor power of 1 MW. Type M2/RP and Type AA films were used when speed was desired. Type AA...

Lorenz, Robert Wayne

2012-06-07T23:59:59.000Z

12

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

SciTech Connect

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

13

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

SciTech Connect

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

14

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

SciTech Connect

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

15

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

NLE Websites -- All DOE Office Websites (Extended Search)

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

16

Business Case for Early Orders of New Nuclear Reactors Executive Overview  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Early Orders of New Nuclear Reactors Early Orders of New Nuclear Reactors Executive Overview Page EX - 1 Business Case for New Business Case for New Nuclear Power Plants Nuclear Power Plants Bringing Public and Private Resources Together for Nuclear Energy Mitigating Critical Risks on Early Orders for New Reactors Briefing for NERAC October 1, 2002 Disclaimer: This draft report was prepared to help the Department of Energy determine the barriers related to the deployment of new nuclear power plants but does not necessarily represent the views or policy of the Department. Business Case for Early Orders of New Nuclear Reactors Executive Overview Page EX - 2 Integrated Project Team Process * Integrated project team (IPT) approach facilitated consideration of complex issues involved in the project and to ensure contractor access to important data from NE.

17

An Overview of the Safety Case for Small Modular Reactors  

SciTech Connect

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

18

Reliability study of a special decay heat removal system of a gas-cooled fast reactor demonstrator  

Science Journals Connector (OSTI)

Abstract The European roadmap toward the development of generation IV concepts addresses the safety and reliability assessment of the special system designed for decay heat removal of a gas-cooled fast reactor demonstrator (GFRD). The envisaged system includes the combination of both active and passive means to accomplish the fundamental safety function. Failure probabilities are calculated on various system configurations, according to either pressurized or depressurized accident events under investigation, and integrated with probabilities of occurrence of corresponding hardware components and natural circulation performance assessment. The analysis suggests the improvement of measures against common cause failures (CCF), in terms of an appropriate diversification among the redundant systems, to reduce the system failure risk. Particular emphasis is placed upon passive system reliability assessment, being recognized to be still an open issue, and the approach based on the functional reliability is adopted to address the point. Results highlight natural circulation as a challenging factor for the decay heat removal safety function accomplishment by means of passive devices. With the models presented here, the simplifying assumptions and the limited scenarios considered according to the level of definition of the design, where many systems are not yet established, one can conclude that attention has to be paid to the functional aspects of the passive system, i.e. the ones not pertaining to the “hardware” of the system. In this article the results of the analysis are discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The design diversity of the components undergoing \\{CCFs\\} can be effective for the improvement and some accident management measures are also possible by making use of the long grace period in GFRD.

Luciano Burgazzi

2014-01-01T23:59:59.000Z

19

Cross flow filtration for radwaste applications reverse osmosis demonstration case studies  

SciTech Connect

Today`s radwaste economic and regulatory scenarios signify the importance in the improvement of operational practices to reduce generator liabilities. This action is largely due to the rising cost dealing with burial sites and the imposed waste volume restriction. To control the economical burdens associated with waste burial and to comply with stricter environmental regulations, NPP`s are attempting to modify their radwaste system(s) design and operating philosophy by placing a major emphasis on waste volume reduction and processing techniques. The utilization of reverse osmosis technology as a means for treatment of process and wastewater streams in the nuclear power industry has been investigated for many years. This paper will outline reverse osmosis theory and highlight performance data for process and waste stream purification applications. Case studies performed at 5 nuclear plants have been outlined. The demonstrations were performed on a widely variety of process stream for both a PWR and BWR application. The data provided by the pilot systems, the equipment design, and the economical impact a reverse osmosis unit will have on producing treated (high purity) are as follows.

Malkmus, D. [VECTRA Technologies, Inc., Columbia, SC (United States)

1995-05-01T23:59:59.000Z

20

Demonstrating Online Monitoring of Air Pollutant Photodegradation in a 3D Printed Gas-Phase Photocatalysis Reactor  

Science Journals Connector (OSTI)

We present a demonstration of online monitoring of gas-phase photocatalytic reactions. A cotton cloth impregnated with commercial titanium dioxide nanoparticles is used as a photocatalytic filter to clean air contaminated with a model pollutant. A fan ...

Bozhidar I. Stefanov; Delphine Lebrun; Andreas Mattsson; Claes G. Granqvist; Lars Österlund

2014-11-21T23:59:59.000Z

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.


21

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

SciTech Connect

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

22

Prospects for spheromak fusion reactors  

Science Journals Connector (OSTI)

The reactor study of Hagenson and Krakowski demonstrated the attractiveness of the spheromak as a compact fusion reactor, based on...

T. K. Fowler; D. D. Hua

1995-06-01T23:59:59.000Z

23

A case study on effectiveness of structural reliability analysis in nuclear reactor safety assessment  

Science Journals Connector (OSTI)

Problems on reliability of structural integrity occupy an important position in various aspects of nuclear reactor safety. In the present paper, an effective method for quantitative evaluation of structural reliability based on ‘stress strength model’ is developed with the objectives of taking a larger number of factors into the evaluation than before and giving useful results within moderate computing time. The method is applied to the reliability analysis of PWR pressure vessels. The results show the relative importance of inspection as well as the parameter uncertainty for assuring the reliability of the structure, although analysis is limited within the scope of linear elastic fracture mechanics (LEFM). This case study also shows that the analysis of structural reliability is effective for safety assessment of nuclear power plants in general and possibly for the improvements of the consistency in the design code.

A. Yamaguchi; S. Kondo; Y. Togo

1983-01-01T23:59:59.000Z

24

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

SciTech Connect

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

25

Portfolio for fast reactor collaboration  

SciTech Connect

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

26

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

SciTech Connect

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

27

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

28

GATEWAY Demonstrations  

Energy.gov (U.S. Department of Energy (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.

29

Nuclear Reactor (atomic reactor)  

Science Journals Connector (OSTI)

A nuclear reactor splits Uranium or Plutonium nuclei, and the...235 is fissionable but more than 99% of the naturally occurring Uranium is U238 that makes enrichment mandatory. In some reactors U238 and Thorium23...

2008-01-01T23:59:59.000Z

30

Spent fuel pyroprocessing demonstration  

SciTech Connect

A major element of the shutdown of the US liquid metal reactor development program is managing the sodium-bonded spent metallic fuel from the Experimental Breeder Reactor-II to meet US environmental laws. Argonne National Laboratory has refurbished and equipped an existing hot cell facility for treating the spent fuel by a high-temperature electrochemical process commonly called pyroprocessing. Four products will be produced for storage and disposal. Two high-level waste forms will be produced and qualified for disposal of the fission and activation products. Uranium and transuranium alloys will be produced for storage pending a decision by the US Department of Energy on the fate of its plutonium and enriched uranium. Together these activities will demonstrate a unique electrochemical treatment technology for spent nuclear fuel. This technology potentially has significant economic and technical advantages over either conventional reprocessing or direct disposal as a high-level waste option.

McFarlane, L.F.; Lineberry, M.J.

1995-05-01T23:59:59.000Z

31

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

32

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

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

33

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

34

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

35

BWR oxygen control demonstration program  

SciTech Connect

A number of cracks have occurred recently in certain BWR piping systems. The operating environment associated with oxidizing species such as oxygen and hydrogen peroxide is considered one of the factors in the mechanism of cracking. In April 1976, NWT Corporation was contracted to perform a BWR oxygen control demonstration program. Means for reducing reactor water oxygen and hydrogen peroxide concentrations during startup and shutdown transients were defined and demonstrated at Vermont Yankee and Browns Ferry Unit 3. Results of the demonstrations and an analytical review of impacts of major system variables on oxygen transients are discussed herein.

Pearl, W.L.; Kassen, W.R.; Sawochka, S.G.

1981-05-01T23:59:59.000Z

36

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations, which govern the operation of commercial nuclear power plants, require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including

37

Reactor Safety Research Programs  

SciTech Connect

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

38

Ignition reactor and pump pulse parameters in a reactor–laser system  

Science Journals Connector (OSTI)

The experience gained in operating a demonstration nuclear-pumped laser in stand B (Physics and Power- Engineering Institute (FEI)) with a pulsed ignition reactor based on the 235U BARS-6 reactor is analyzed. It ...

P. P. D’yachenko; G. N. Fokin

2012-09-01T23:59:59.000Z

39

Novel Catalytic Membrane Reactors  

SciTech Connect

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

40

NUCLEAR REACTORS.  

E-Print Network (OSTI)

??Nuclear reactors are devices containing fissionable material in sufficient quantity and so arranged as to be capable of maintaining a controlled, self-sustaining NUCLEAR FISSION chain… (more)

Belachew, Dessalegn

2010-01-01T23:59:59.000Z

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 technology assessment and selection utilizing systems engineering approach  

SciTech Connect

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

42

Fluoride Salt-Cooled High-Temperature Reactor Development Roadmap  

SciTech Connect

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

43

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

44

Small Reactor for Deep Space Exploration  

SciTech Connect

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

45

West Valley Demonstration Project  

Energy.gov (U.S. Department of Energy (DOE))

West Valley Demonstration Project compliance agreements, along with summaries of the agreements, can be viewed here.

46

Thermionic Reactor Design Studies  

SciTech Connect

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

47

CALDERON COKEMAKING PROCESS/DEMONSTRATION PROJECT  

SciTech Connect

This project deals with the demonstration of a coking reactor (Process Development Unit-- PDU-11) using Calderon's proprietary technology for making commercially acceptable coke. The activities of the past quarter were focused on the following: 1. Testing and Designing of the Submerged Quenching Closed System for the Process; 2. Usage of the Cracked Desulfurized Gas as a Reducing Gas to Make Directly Reduced Iron (DRI) in Order to Make the Process Economics Viable; 3. Changes in the Ceramic Liners for Supporting Them in the Coking Reactor; 4. Work Towards Testing of U.S. Steel's Coal in the Existing Process Development Unit in Alliance (PDU-1); 5. Permitting.

Albert Calderon

1998-04-08T23:59:59.000Z

48

Prototypical Consolidation Demonstration Project: Final report  

SciTech Connect

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

49

naval reactors  

National Nuclear Security Administration (NNSA)

After operating for 34 years and training over 14,000 sailors, the Department of Energy S1C Prototype Reactor Site in Windsor, Connecticut, was returned to "green field"...

50

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Assessment of High Value Surveillance Materials Assessment of High Value Surveillance Materials Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Assessment of High Value Surveillance Materials The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely

51

LIMB Demonstration Project Extension and Coolside Demonstration  

SciTech Connect

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

52

Temperature effects on chemical reactor  

Science Journals Connector (OSTI)

In this paper we had to study some characteristics of the chemical reactors from which we can understand the reactor operation in different circumstances; from these and the most important factor that has a great effect on the reactor operation is the temperature it is a mathematical processing of a chemical problem that was already studied but it may be developed by introducing new strategies of control; in our case we deal with the analysis of a liquid?gas reactor which can make the flotation of the benzene to produce the ethylene; this type of reactors can be used in vast domains of the chemical industry especially in refinery plants where we find the oil separation and its extractions whether they are gases or liquids which become necessary for industrial technology especially in our century.

M. Azzouzi

2008-01-01T23:59:59.000Z

53

Molecular vibration demonstrations  

Science Journals Connector (OSTI)

Molecular vibration demonstrations ... Two dynamic models that illustrate the normal-mode vibrations of the water and benzene molecules. ...

George Turrell; Robert Demol

1987-01-01T23:59:59.000Z

54

Research reactors - an overview  

SciTech Connect

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

55

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

56

Light Water Reactor Sustainability  

NLE Websites -- All DOE Office Websites (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...

57

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

58

Radiation Emergency Procedure Demonstrations  

NLE Websites -- All DOE Office Websites (Extended Search)

Managing Radiation Emergencies Managing Radiation Emergencies Procedure Demonstrations Procedure Demonstrations Note: RealPlayer is needed for listening to the narration that accompany these demonstrations. Real Player Dressing To Prevent the Spread of Radioactive Contamination This demonstration shows how your team can dress to prevent the spread of radioactive contamination. Click to begin presentation on dressing to prevent the spread of radioactive contamination. Preparing The Area This demonstration shows basic steps you can take to gather equipment and prepare a room to receive a patient who may be contaminated with radioactive material. Click to begin presentation on preparing a room to receive a radioactive contaminated patient. Removing Contaminated Clothing This demonstration shows the procedure for removing clothing from a patient who may be contaminated with radioactive material.

59

LIMB demonstration project extension  

SciTech Connect

The purpose of the DOE limestone injection multistage burner (LIMB) Demonstration Project Extension is to extend the data base on LIMB technology and to expand DOE's list of Clean Coal Technologies by demonstrating the Coolside process as part of the project. The main objectives of this project are: to demonstrate the general applicability of LIMB technology by testing 3 coals and 4 sorbents (total of 12 coal/sorbent combinations) at the Ohio Edison Edgewater plant; and to demonstrate that Coolside is a viable technology for improving precipitator performance and reducing sulfur dioxide emissions while acceptable operability is maintained. Progress is reported. 3 figs.

Not Available

1990-09-21T23:59:59.000Z

60

West Valley Demonstration Project  

Energy.gov (U.S. Department of Energy (DOE))

The West Valley Demonstration Project came into being through the West Valley Demonstration Project Act of 1980. The Act requires that the DOE is responsible for solidifying the high-level waste, disposing of waste created by the solidification, and decommissioning the facilities used in the process.

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

CALDERON COKEMAKING PROCESS/DEMONSTRATION PROJECT  

SciTech Connect

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

62

Modal approximation of xenon oscillations in nuclear reactors — an FEM-based approach  

Science Journals Connector (OSTI)

A modal expansion method for modelling dynamic space-dependent effects (xenon oscillations) in large nuclear reactors is presented. Based on the finite element method (FEM), modal expansion functions for various cases with practical significance can be calculated. In this way it is possible to derive a low order state space model describing the considered effect. The use of this model in on-line controller design and simulation is intended. Modal methods in connection with modern numerical algorithms turned out to be effective to treat distributed parameter systems governed by elliptic differential equations. This will be demonstrated by application to a VVER-1000 pressurized water reactor.

J.Chr. Kalkkuhl; M.G. Döring

1993-01-01T23:59:59.000Z

63

Argonne step closer to safer nuclear reactor  

Science Journals Connector (OSTI)

Argonne step closer to safer nuclear reactor ... "A key technological link" toward development of meltdown-immune nuclear reactors is now in the demonstration phase at Argonne National Laboratory near Chicago. ... The technique is part of Argonne's continuing interest in the sodium-cooled integral fast reactor (IFR), whose immunity to meltdown derives from molten sodium's function as a heat sink and the use of metallic fuel that conducts heat better than conventional oxide fuels. ...

WARD WORTHY

1988-05-30T23:59:59.000Z

64

Core Drilling Demonstration  

Energy.gov (U.S. Department of Energy (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...

65

Chevrolet Volt Vehicle Demonstration  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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...

66

Montana ICTL Demonstration Program  

NLE Websites -- All DOE Office Websites (Extended Search)

Montana ICTL Demonstration Program Montana ICTL Demonstration Program Background The Department of Energy (DOE) funds basic and applied research toward the development of technologies that will allow the U.S. to depend to a greater extent on renewable fuels, especially those derived from domestic sources of energy. Coal is one of the nation's most abundant domestic energy resources; however, conventional technologies using coal release large amounts of carbon dioxide (CO

67

successfully demonstrated the separation  

NLE Websites -- All DOE Office Websites (Extended Search)

successfully demonstrated the separation and capture of 90 percent successfully demonstrated the separation and capture of 90 percent of the c arbon dioxide (CO 2 ) from a pulve rized coal plant. In t he ARRA-funded project, Membrane Technology and Research Inc. (MTR) and its partners tested the Polaris(tm) membrane system, which uses a CO 2 -selective polymeric membrane material and module to capture CO 2 from a plant's flue gas. Since the Polaris(tm) membranes

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

Light Water Reactors Technology Development - Nuclear Reactors  

NLE Websites -- All DOE Office Websites (Extended Search)

Light Water Reactors Light Water Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

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

GATEWAY Demonstration Outdoor Projects  

Energy.gov (U.S. Department of Energy (DOE))

DOE shares the results of completed GATEWAY demonstration projects, publishing detailed reports that include analysis of data collected, projected energy savings, economic analyses, and user feedback. Report briefs summarize key findings in a quick-scan format. Both the reports and briefs are available as Adobe Acrobat PDFs.

72

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

73

GATEWAY Demonstration Indoor Projects  

Energy.gov (U.S. Department of Energy (DOE))

DOE shares the results of completed GATEWAY demonstration projects, publishing detailed reports that include analysis of data collected, projected energy savings, economic analyses, and user feedback. Report briefs summarize key findings in a quick-scan format. Both the reports and briefs are available as Adobe Acrobat PDFs.

74

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

75

LIMB Demonstration Project Extension  

SciTech Connect

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

76

LIMB Demonstration Project Extension  

SciTech Connect

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

1989-03-15T23:59:59.000Z

77

LIMB Demonstration Project Extension  

SciTech Connect

The basic goal of the Limestone Injection Mitigation 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

1989-11-15T23:59:59.000Z

78

LIMB demonstration project extension  

SciTech Connect

The main objectives of this project are: (1) To demonstrate the general applicability of Limestone Injection Multistage Burner (LIMB) technology by testing 3 coals and 4 sorbents (total of 12 coal/sorbent combinations) at the Ohio Edison Edgewater Plant. (2) To demonstrate that Coolside is a viable technology for improving precipitator performance and reducing sulfur dioxide emissions while acceptance operability is maintained. During the past quarter, activities for phase I, design and permitting, and phase II, construction, shakedown and start-up were completed for phase III, operation, data collection, reporting and disposition, activities continued with consol completing the revisions to the Coolside Topical report, the completion of LIMB Extension testing, and the start of demobilization and restoration.

Not Available

1991-12-16T23:59:59.000Z

79

LIMB Demonstration Project Extension  

SciTech Connect

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

1988-12-15T23:59:59.000Z

80

LIMB Demonstration Project Extension  

SciTech Connect

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 and NO 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

1988-09-15T23:59:59.000Z

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

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

1988-03-15T23:59:59.000Z

82

Photocatalytic reactor  

DOE Patents (OSTI)

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

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

1999-01-19T23:59:59.000Z

83

LIMB Demonstration Project Extension  

SciTech Connect

The DOE LIMB Demonstration Project Extension is a continuation of the EPA Limestone Injection Multistage Burner (LIMB) Demonstration. EPA ultimately expects to show that LIMB is a low cost control technology capable of producing moderate SO{sub x} and NO{sub x} control (50--60 percent) with applicability for retrofit to the major portion of the existing coal-fired boiler population. The current EPA Wall-Fired LIMB Demonstration is a four-year project that includes design and installation of a LIMB system at the 105-MW Unit 4 boiler at Ohio Edison's Edgewater Station in Lorain, Ohio. LIMB Extension testing continued during the quarter with lignosulfonated hydrated lime, pulverized limestone, and hydrated dolomitic lime while firing 1.8% and 3% sulfur coals. Sulfur dioxide removal efficiencies were equivalent to the results found during EPA, base LIMB testing. Sulfur dioxide removal efficiencies were lower than expected while testing with pulverized limestone without humidification. A slight increase in sulfur capture was noted while injecting pulverized limestone at the 187' elevation and with the humidifier outlet temperature at 145{degree}F.

Not Available

1990-09-21T23:59:59.000Z

84

LIMB Demonstration Project Extension  

SciTech Connect

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 (1) 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; (2) boiler reliability, operability, and steam production can be maintained at levels existing prior to LIMB retrofit; and (3) 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. The demonstration project consists of several distinct phases: a preliminary phase to develop the LIMB process design applicable to the host boiler, a construction and start-up phase, and an operating and evaluation phase. The first major activity, the development of the Edgewater LIMB design, was completed in January 1986 and detailed engineering is now complete. Major boiler-related components were installed during a September 1986 boiler outage. Start-up activities began in March of 1987 with tuning of the low NO{sub x} burners. Sorbent injection activities were underway as of July 1987. 3 figs.

Not Available

1991-09-15T23:59:59.000Z

85

Evaluation of Torsatrons as reactors  

SciTech Connect

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors. This scoping study, which uses an integrated cost-minimization code that incorporates costing and reactor component models self-consistently with a 1-D energy transport calculation, shows that a torsatron reactor could also be economically competitive with a tokamak reactor. The projected cost of electricity (COE) estimated using the Advanced Reactor Innovation and Evaluation Studies (ARIES) costing algorithms is 65.6 mill/kW(e)h in constant 1992 dollars for a reference 1-GW(e) Compact Torsatron reactor case. The COE is relatively insensitive (<10% variation) over a wide range of assumptions, including variations in the maximum field allowed on the coils, the coil elongation, the shape of the density profile, the beta limit, the confinement multiplier, and the presence of a large loss region for alpha particles. The largest variations in the COE occur for variations in the electrical power output demanded and the plasma-coil separation ratio.

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)

1994-03-01T23:59:59.000Z

86

Synergistic smart fuel for in-pile nuclear reactor measurements  

SciTech Connect

The thermo-acoustic fuel rod sensor developed in this research has demonstrated a novel technique for monitoring the temperature within the core of a nuclear reactor or the temperature of the surrounding heat-transfer fluid. It uses the heat from the nuclear fuel to generate sustained acoustic oscillations whose frequency will be indicative of the temperature. Converting a nuclear fuel rod into this type of thermo-acoustic sensor simply requires the insertion of a porous material (stack). This sensor has demonstrated a synergy with the elevated temperatures that exist within the nuclear reactor using materials that have only minimal susceptibility to high-energy particle fluxes. When the sensor is in operation, the sound waves radiated from the fuel rod resonator will propagate through the surrounding cooling fluid. The frequency of these oscillations is directly correlated with an effective temperature within the fuel rod resonator. This device is self-powered and is operational even in case of total loss of power of the reactor.

Smith, J.A.; Kotter, D.K. [Idaho National Laboratories, Idaho Falls (United States); Ali, R.A.; Garrett, S.L. [Penn State University, University Park, State College, PA 16801 (United States)

2013-07-01T23:59:59.000Z

87

SHARP Assembly-Scale Multiphysics Demonstration Simulations | Argonne  

NLE Websites -- All DOE Office Websites (Extended Search)

SHARP Assembly-Scale Multiphysics Demonstration Simulations SHARP Assembly-Scale Multiphysics Demonstration Simulations Title SHARP Assembly-Scale Multiphysics Demonstration Simulations Publication Type Report Year of Publication 2013 Authors Tautges, TJ, Fischer, PF, Grindeanu, I, Jain, R, Mahajan, A, Obabko, AV, Smith, MA, Merzari, E, Ferencz, R Document Number ANL/MCS-NE-13-9 Abstract The NEAMS Reactor Product Line effort aims to develop an integrated multiphysics simulation capability for the design and analysis of future generations of nuclear power plants. The Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms.

88

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

89

GEN-IV Reactors  

Science Journals Connector (OSTI)

Generation-IV reactors are a set of nuclear reactors currently being developed under international collaborations targeting ... economics, proliferation resistance, and physical protection of nuclear energy. Nuclear

Taek K. Kim

2013-01-01T23:59:59.000Z

90

The Netherlands Reactor Centre  

Science Journals Connector (OSTI)

... Two illustrated brochures in English have recently J. been issued by the Netherlands Reactor Centre ( ... Centre (Reactor Centrum Nederland). The first* gives a general survey of the ...

S. WEINTROUB

1964-04-04T23:59:59.000Z

91

Fuel Cell Demonstration Program  

SciTech Connect

In an effort to promote clean energy projects and aid in the commercialization of new fuel cell technologies the Long Island Power Authority (LIPA) initiated a Fuel Cell Demonstration Program in 1999 with six month deployments of Proton Exchange Membrane (PEM) non-commercial Beta model systems at partnering sites throughout Long Island. These projects facilitated significant developments in the technology, providing operating experience that allowed the manufacturer to produce fuel cells that were half the size of the Beta units and suitable for outdoor installations. In 2001, LIPA embarked on a large-scale effort to identify and develop measures that could improve the reliability and performance of future fuel cell technologies for electric utility applications and the concept to establish a fuel cell farm (Farm) of 75 units was developed. By the end of October of 2001, 75 Lorax 2.0 fuel cells had been installed at the West Babylon substation on Long Island, making it the first fuel cell demonstration of its kind and size anywhere in the world at the time. Designed to help LIPA study the feasibility of using fuel cells to operate in parallel with LIPA's electric grid system, the Farm operated 120 fuel cells over its lifetime of over 3 years including 3 generations of Plug Power fuel cells (Lorax 2.0, Lorax 3.0, Lorax 4.5). Of these 120 fuel cells, 20 Lorax 3.0 units operated under this Award from June 2002 to September 2004. In parallel with the operation of the Farm, LIPA recruited government and commercial/industrial customers to demonstrate fuel cells as on-site distributed generation. From December 2002 to February 2005, 17 fuel cells were tested and monitored at various customer sites throughout Long Island. The 37 fuel cells operated under this Award produced a total of 712,635 kWh. As fuel cell technology became more mature, performance improvements included a 1% increase in system efficiency. Including equipment, design, fuel, maintenance, installation, and decommissioning the total project budget was approximately $3.7 million.

Gerald Brun

2006-09-15T23:59:59.000Z

92

CCUS Demonstrations Making Progress  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

9, First Quarter, 2013 9, First Quarter, 2013 www.fossil.energy.gov/news/energytoday.html HigHligHts inside 2 CCUS Demonstrations Making Progress A Column from the Director of Clean Energy Sys- tems, Office of Clean Coal 4 LNG Exports DOE Releases Third Party Study on Impact of Natural Gas Exports 5 Providing Emergency Relief Petroleum Reservers Helps Out with Hurricane Relief Efforts 7 Game-Changing Membranes FE-Funded Project Develops Novel Membranes for CCUS 8 Shale Gas Projects Selected 15 Projects Will Research Technical Challenges of Shale Gas Development A project important to demonstrat- ing the commercial viability of carbon capture, utilization and storage (CCUS) technology has completed the first year of inject-

93

Jennings Demonstration PLant  

SciTech Connect

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

94

Fusion Power Demonstration III  

SciTech Connect

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

Lee, J.D. (ed.)

1985-07-01T23:59:59.000Z

95

Business Opportunities for Small Reactors  

SciTech Connect

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

Minato, Akio; Nishimura, Satoshi [Central Research Institute of Electric Power Industry - CRIEPI, 2-11-1 Iwado-Kita, Komae, Tokyo 201-8511 (Japan); Brown, Neil W. [Lawrence Livermore National Laboratory - LLNL, PO Box 808, Livermore, CA 94551 (United States)

2007-07-01T23:59:59.000Z

96

(Liquid metal reactor/fast breeder reactor research and development)  

SciTech Connect

The second meeting of the UJCC was held in Japan on June 6--8, 1990. The first day was devoted to presentations of the status of the US and Japanese Fast Breeder Reactor (FBR) programs and the status of specific areas of cooperative work. Briefly, the Japanese are following the FBR development program which has been in place since the 1970s. This program includes an FBR test reactor (JOYO), a pilot-scale reactor (MONJU), a demonstration-scale plant, and commercial-scale plants by about 2020. The US program has been redirected toward an actinide recycle mission using metal fuel and pyroprocessing of spent fuel to recovery both Pu and the higher actinides for return to the Liquid Metal Reactor (LMR). The second day was spent traveling from Tokyo to Tsuruga for a tour of the MONJU reactor. The tour was especially interesting. The third day was spent writing the minutes of the meeting and the return trip to Tokyo.

Homan, F.J.

1990-06-20T23:59:59.000Z

97

SRS Small Modular Reactors  

SciTech Connect

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

None

2012-04-27T23:59:59.000Z

98

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

99

RELAP5 Simulation of Thermal-Hydraulic Behavior in a CANDU Reactor - Assessments of RD-14 Experiments  

SciTech Connect

The critical reactor header break and the thermosiphoning experiments in the RD-14 test facility were simulated with the RELAP5/MOD3.1 code. The RELAP5 code has been developed for best-estimate transient simulation of pressurized water reactors and associated systems, but it has not been assessed for a Canada deuterium uranium (CANDU) reactor. Therefore, this study has been initiated with an aim to identify the code applicability in a CANDU reactor by simulating some of the tests performed in the RD-14 facility. The RD-14 test facility at Whiteshell Nuclear Research Establishment is a full-scale pressurized-water loop. The RD-14 is not a scale model of any particular CANDU reactor. Rather, it possesses many geometric features of a CANDU reactor heat transport system and is capable of operating at conditions similar to those expected to occur in a reactor under normal operation and some postulated accident conditions. In this study, two critical reactor header break tests (B8711 and B8713) and three thermosiphoning tests (T8513, T8515, and T8517) were analyzed with the RELAP5 code. The results were compared with experimental data and those of CATHENA performed by Atomic Energy of Canada Ltd. The RELAP5 analyses demonstrate the code's capability to predict reasonably the main phenomena occurring in the transient, in both the qualitative and the quantitative view. However, some discrepancies after the emergency coolant injection for the critical break case and also related to the behaviors of the mass flow rate and the primary pressure for the thermosiphoning case were observed.

Lee, Sukho; Kim, In-Goo [Korea Institute of Nuclear Safety (Korea, Republic of)

2000-04-15T23:59:59.000Z

100

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

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

Operational control of boiling water reactor stability  

SciTech Connect

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

102

Thermionic Reactor Design Studies  

SciTech Connect

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

103

Rod consolidation at the West Valley Demonstration Project  

SciTech Connect

A rod consolidation demonstration with irradiated pressurized water reactor fuel was recently conducted by personnel from Nuclear Assurance Corporation and West Valley Nuclear Services Company at the West Valley Demonstration Project in West Valley, New York. The rod consolidation demonstration involved pulling all of the fuel rods from six fuel Assemblies. In general, the rod pulling proceeded smoothly. The highest compaction ratio attained was 1:8:1. Among the total of 1074 fuel rods were some known degraded rods (they had collapsed cladding, a result of in-reactor fuel densification), but no rods were broken or dropped during the demonstration. One aim was to gather information on the effect of rod consolidation operations on the integrity of the fuel rods during subsequent handling and storage. Another goal was to collect information on the condition and handling of intact, damaged, and failed fuel that has been in storage for an extended period. 9 refs., 8 figs., 1 tab.

Bailey, W.J.

1986-12-01T23:59:59.000Z

104

Dry Storage Demonstration for High-Burnup Spent Nuclear Fuel-Feasibility Study  

SciTech Connect

Initially, casks for dry storage of spent fuel were licensed for assembly-average burnup of about 35 GWd/MTU. Over the last two decades, the discharge burnup of fuel has increased steadily and now exceeds 45 GWd/MTU. With spent fuel burnups approaching the licensing limits (peak rod burnup of 62 GWd/MTU for pressurized water reactor fuel) and some lead test assemblies being burned beyond this limit, a need for a confirmatory dry storage demonstration program was first identified after the publication in May 1999 of the U.S. Nuclear Regulatory Commissions (NRC) Interim Staff Guidance 11 (ISG-11). With the publication in July 2002 of the second revision of ISG-11, the desirability for such a program further increased to obtain confirmatory data about the potential changes in cladding mechanical properties induced by dry storage, which would have implications to the transportation, handling, and disposal of high-burnup spent fuel. While dry storage licenses have kept pace with reactor discharge burnups, transportation licenses have not and are considered on a case by case basis. Therefore, this feasibility study was performed to examine the options available for conducting a confirmatory experimental program supporting the dry storage, transportation, and disposal of spent nuclear fuel with burnups well in excess of 45 GWd/MTU.

McKinnon, Mikal A. (BATTELLE (PACIFIC NW LAB)); Cunningham, Mitchel E. (BATTELLE (PACIFIC NW LAB))

2003-09-09T23:59:59.000Z

105

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

106

Elementary Reactor Physics  

Science Journals Connector (OSTI)

... THERE are few subjects which have developed at the rate at which reactor physics and ... physics and reactor theory have done. This, of course, is largely due to the circumstances in ...

J. F. HILL

1962-02-10T23:59:59.000Z

107

Colliding Beam Fusion Reactors  

Science Journals Connector (OSTI)

The recirculating power for virtually all types of fusion reactors has previously been calculated [1] with the Fokker–Planck equation. The reactors involve non-Maxwellian plasmas. The calculations are ... the rec...

Norman Rostoker; Artan Qerushi; Michl Binderbauer

2003-06-01T23:59:59.000Z

108

Technology Demonstrations | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Demonstrations Demonstrations Technology Demonstrations Efficient new building technologies can help meet our country's energy goals, stimulate U.S. manufacturing, create jobs, and improve the environment. However, many high-performing technologies are not readily adopted in the marketplace due to lack of information about their real-world performance. To address this gap in information, the DOE frequently supports demonstrations to assess technologies' energy performance, installation procedures, operations, and maintenance characteristics. The information from these demonstrations helps consumers make more informed decisions and helps U.S. manufacturers validate the performance of their products. Frequently Asked Questions How does DOE prioritize demonstration projects?

109

CALDERON COKEMAKING PROCESS/DEMONSTRATION PROJECT  

SciTech Connect

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

Advanced Test Reactor Tour  

SciTech Connect

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

111

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

112

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

113

Materials Degradation in Light Water Reactors: Life After 60 | Department  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

114

Materials Degradation in Light Water Reactors: Life After 60 | Department  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

115

The Integral Fast Reactor: A practical approach to waste management  

SciTech Connect

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

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

The Argonaut Reactor - Reactors designed/built by Argonne National  

NLE Websites -- All DOE Office Websites (Extended Search)

Achievements > Achievements > Argonne Reactors > Training Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

118

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

119

Micro-Scale Catalytic Reactor for Syngas Production  

Science Journals Connector (OSTI)

Micro-Scale Catalytic Reactor for Syngas Production ... The H2 yield (?H2) was evaluated as moles of H2 produced per mole of CH4 converted in the reforming channel. ... In particular, the maximum temperature moves from a location close to the reactor center (for MCH4-sr/MCH4-co = 2) toward the reactor inlet producing in the case of MCH4-sr/MCH4-co =2.75 decreasing temperature profiles without a maximum. ...

S. Vaccaro; L. Malangone; P. Ciambelli

2010-07-01T23:59:59.000Z

120

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

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

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

122

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

SciTech Connect

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

123

POTENTIAL BENCHMARKS FOR ACTINIDE PRODUCTION IN HANFORD REACTORS  

SciTech Connect

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

124

Research, Development, Demonstration, and Deployment  

Energy.gov (U.S. Department of Energy (DOE))

The Bioenergy Technologies Office's research, development, demonstration, and deployment (RDD&D) efforts are organized around five key technical and three cross-cutting elements. The first two...

125

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

126

Hydrogen Storage Materials Database Demonstration  

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

| Fuel Cell Technologies Program Source: US DOE 4252011 eere.energy.gov Hydrogen Storage Materials Database Demonstration FUEL CELL TECHNOLOGIES PROGRAM Ned Stetson Storage Tech...

127

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

SciTech Connect

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

128

NETL: News Release - DOE-Supported Project Demonstrates Benefits of  

NLE Websites -- All DOE Office Websites (Extended Search)

DOE-Supported Project Demonstrates Benefits of Constructed Wetlands to Treat Non-Traditional Water Sources DOE-Supported Project Demonstrates Benefits of Constructed Wetlands to Treat Non-Traditional Water Sources Flue gas desulfurization water was treated in a constructed wetlands system consisting of five reactors planted with vegetation found in natural wetlands. The water to be treated was received from an operating coal-fired power plant in the south-eastern United States. Flue gas desulfurization water was treated in a constructed wetlands system consisting of five "reactors" planted with vegetation found in natural wetlands. The water to be treated was received from an operating coal-fired power plant in the south-eastern United States. Washington, DC - In a pilot-scale test supported by the U.S. Department of Energy (DOE) Office of Fossil Energy, Clemson University researchers

129

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

E-Print Network (OSTI)

The design of passive heat removal systems is one of the main concerns for the modular Very High Temperature Gas-Cooled Reactors (VHTR) vessel cavity. The Reactor Cavity Cooling System (RCCS) is an important heat removal system in case of accidents...

Frisani, Angelo

2011-08-08T23:59:59.000Z

130

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

SciTech Connect

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

131

Ex-Core CFD Analysis Results for the Prometheus Gas Reactor  

SciTech Connect

This paper presents the initial nozzle-to-nozzle (N2N) reactor vessel model scoping studies using computational fluid dynamics (CFD) analysis methods. The N2N model has been solved under a variety of different boundary conditions. This paper presents some of the basic hydraulic results from the N2N CFD analysis effort. It also demonstrates how designers were going to apply the analysis results to modify a number of the design features. The initial goals for developing a preliminary CFD N2N model were to establish baseline expectations for pressure drops and flow fields around the reactor core. Analysis results indicated that the averaged reactor vessel pressure drop for all analyzed cases was 46.9 kPa ({approx}6.8 psid). In addition, mass flow distributions to the three core fuel channel regions exhibited a nearly inverted profile to those specified for the in-core thermal/hydraulic design. During subsequent design iterations, the goal would have been to modify or add design features that would have minimized reactor vessel pressure drop and improved flow distribution to the inlet of the core.

Lorentz, Donald G. [Space Engineering, Bechtel Bettis, Inc. West Mifflin, PA 15122 (United States)

2007-01-30T23:59:59.000Z

132

Progress and status of the integral fast reactor (IFR) development program  

SciTech Connect

This paper discusses the Integral Fast Reactor (IFR) development program, in which the entire reactor system - reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. Detailed discussions on the present status of the IFR technology development activities in the areas of fuels, pyroprocessing, safety, core design, and fuel cycle demonstration are also presented.

Chang, Y.I. (Argonne National Lab., Argonne, IL (US))

1992-01-01T23:59:59.000Z

133

Geothermal EGS Demonstration Photo Library  

Energy.gov (U.S. Department of Energy (DOE))

EGS Demonstrations make up the most advanced research and science investments in the geothermal sector. Five active demonstration sites nationwide are proving the spectrum of EGS potential, in and near existing hydrothermal operations, with infrastructure, and in the longer-term greenfield settings, where no previous geothermal development is operating.

134

Generation -IV Reactor Concepts  

NLE Websites -- All DOE Office Websites (Extended Search)

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

135

BWR radiation control: plant demonstration  

SciTech Connect

The first year's progress is presented for a four-year program intended to implement and evaluate BRAC radiation reduction operational guidelines at the Vermont Yankee BWR and to document the results in sufficient detail to provide guidance to other BWR owners. Past operational, chemistry and radiation level data have been reviewed to provide a historical base of reference. Extensive sampling and chemistry monitoring systems have been installed to evaluate plant chemistry status and the effects of program implemented changes. Radiation surveys and piping gamma scans are being performed at targeted locations to quantify radiation level trends and to identify and quantify piping isotopics. Contact radiation levels on the recirculation line at Vermont Yankee have been increasing at a rate of 175 mR/h-EFPY since 1978. A materials survey of feedwater and reactor components in contact with the process liquid has been performed to identify sources of corrosion product release, particularly cobalt and nickel. A feedwater oxygen injection system has been installed to evaluate the effects of oxygen control on feedwater materials corrosion product releases. A baseline performance evaluation of the condensate treatment and reactor water cleanup systems has been completed. Data on organics and ionics at Vermont Yankee have been obtained. A methodology of BWR feedwater system layup during extended outages was developed, and an evaluation performed of layup and startup practices utilized at Vermont Yankee during the fall 1980 and 1981 refueling outages.

Palino, G.F.; Hobart, R.L.; Wall, P.S.; Sawochka, S.G.

1982-11-01T23:59:59.000Z

136

Major Demonstrations | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Major Demonstrations Major Demonstrations Major Demonstrations A state-of-the-art integrated coal gasification combined-cycle (IGCC) power plant, Tampa Electric's Polk Power Station produces enough electricity to serve 75,000 homes. A state-of-the-art integrated coal gasification combined-cycle (IGCC) power plant, Tampa Electric's Polk Power Station produces enough electricity to serve 75,000 homes. The Office of Fossil Energy is co-funding large-scale demonstrations of clean coal technologies to hasten their adoption into the commercial marketplace. Through the year 2030, electricity consumption in the United States is expected to grow by about 1 percent per year. The ability of coal-fired generation to help meet this demand could be limited by concerns over greenhouse gas emissions. While the Major Demonstrations performed to date

137

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

SciTech Connect

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

138

Early Site Permit Demonstration Program: Station design alternatives report  

SciTech Connect

This report provides the results of investigating the basis for including Station Design Alternatives (SDAs) in the regulatory guidance given for nuclear plant environmental reports (ERs), explains approaches or processes for evaluating SDAs at the early site permit (ESP) stage, and applies one of the processes to each of the ten systems or subsystems considered as SDAS. The key objective o this report s to demonstrate an adequate examination of alternatives can be performed without the extensive development f design data. The report discusses the Composite Suitability Approach and the Established Cutoff Approach in evaluating station design alternatives and selects one of these approaches to evaluate alternatives for each of the plant or station that were considered. Four types of ALWRs have been considered due to the availability of extensive plant data: System 80+, AP600, Advanced Boiling Reactor (ABWR), and Simplified Boiling Water Reactor (SBWR). This report demonstrates the feasibility of evaluating station design alternatives when reactor design detail has not been determined, quantitatively compares the potential ental impacts of alternatives, and focuses the ultimate selection of a alternative on cost and applicant-specific factors. The range of alternatives system is deliberately limited to a reasonable number to demonstrate the or to the three most commonly used at operating plants.

Not Available

1993-03-01T23:59:59.000Z

139

Power Reactor Progress  

Science Journals Connector (OSTI)

Argonne kicks off EBWR; Allis-Chalmers plans power reactor using both nuclear and conventional fuels ... NUCLEAR POWER took two giant steps last week. ... Just as the first nuclear power system in the U. S. designed and built solely for the generation of electric power went into full operation at Argonne, Allis-Chalmers came up with a new twist in power reactors—a controlled recirculation boiling reactor (CRBR) using both nuclear and conventional fuels (C&EN, Feb. 18, page 7). ...

1957-02-25T23:59:59.000Z

140

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

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

AEC Pushes Fusion Reactors  

Science Journals Connector (OSTI)

AEC Pushes Fusion Reactors ... Project Sherwood, as the study program is called, began in 1951-52 soon after the first successful thermonuclear explosion in the Pacific. ...

1955-10-10T23:59:59.000Z

142

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

143

Advanced Nuclear Research Reactor  

SciTech Connect

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

Lolich, J.V.

2004-10-06T23:59:59.000Z

144

Critical Heat Flux -CHF in Liquid Metal in Presence of a Magnetic Field with Particular Reference to Fusion Reactor Project  

Science Journals Connector (OSTI)

Knowledge of the critical heat flux q??crit is a cornerstone of reactor design fission, but as will demonstrate also in fusion reactors. This quantity cannot be deduced directly,...

F. J. Arias

2010-04-01T23:59:59.000Z

145

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

146

The integral fast reactor fuel cycle  

SciTech Connect

The liquid-metal reactor (LMR) has the potential to extend the uranium resource by a factor of 50 to 100 over current commercial light water reactors (LWRs). In the integral fast reactor (IFR) development program, the entire reactor system - reactor, fuel cycle, and waste process - is being developed and optimized at the same time as a single integral entity. A key feature of the IFR concept is the metallic fuel. The lead irradiation tests on the new U-Pu-Zr metallic fuel in the Experimental Breeder Reactor II have surpassed 185000 MWd/t burnup, and its high burnup capability has now been fully demonstrated. The metallic fuel also allows a radically improved fuel cycle technology. Pyroprocessing, which utilizes high temperatures and molten salt and molten metal solvents, can be advantageously utilized for processing metal fuels because the product is metal suitable for fabrication into new fuel elements. Direct production of a metal product avoids expensive and cumbersome chemical conversion steps that would result from use of the conventional Purex solvent extraction process. The key step in the IFR process is electrorefining, which provides for recovery of the valuable fuel constituents, uranium and plutonium, and for removal of fission products. A notable feature of the IFR process is that the actinide elements accompany plutonium through the process. This results in a major advantage in the high-level waste management.

Chang, Y.I. (Argonne National Lab., IL (United States))

1990-01-01T23:59:59.000Z

147

Offsite demonstrations for MWLID technologies  

SciTech Connect

The goal of the Offsite Demonstration Project for Mixed Waste Landfill Integrated Demonstration (MWLID)-developed environmental site characterization and remediation technologies is to facilitate the transfer, use, and commercialization of these technologies to the public and private sector. The meet this goal, the project identified environmental restoration needs of mixed waste and/or hazardous waste landfill owners (Native American, municipal, DOE, and DoD); documenting potential demonstration sites and the contaminants present at each site; assessing the environmental regulations that would effect demonstration activities; and evaluating site suitability for demonstrating MWLID technologies at the tribal and municipal sites identified. Eighteen landfill sites within a 40.2-km radius of Sandia National Laboratories are listed on the CERCLIS Site/Event Listing for the state of New Mexico. Seventeen are not located within DOE or DoD facilities and are potential offsite MWLID technology demonstration sites. Two of the seventeen CERCLIS sites, one on Native American land and one on municipal land, were evaluated and identified as potential candidates for off-site demonstrations of MWLID-developed technologies. Contaminants potentially present on site include chromium waste, household/commercial hazardous waste, volatile organic compounds, and petroleum products. MWLID characterization technologies applicable to these sites include Magnetometer Towed Array, Cross-borehole Electromagnetic Imaging, SitePlanner {trademark}/PLUME, Hybrid Directional Drilling, Seamist{trademark}/Vadose Zone Monitoring, Stripping Analyses, and x-ray Fluorescence Spectroscopy for Heavy Metals.

Williams, C. [Sandia National Labs., Albuquerque, NM (United States); Gruebel, R. [Tech. Reps., Inc., Albuquerque, NM (United States)

1995-04-01T23:59:59.000Z

148

The physics of magnetic fusion reactors  

Science Journals Connector (OSTI)

During the past two decades there have been substantial advances in magnetic fusion research. On the experimental front, progress has been led by the mainline tokamaks, which have achieved reactor-level values of temperature and plasma pressure. Comparable progress, when allowance is made for their smaller programs, has been made in complementary configurations such as the stellarator, reversed-field pinch and field-reversed configuration. In this paper, the status of understanding of the physics of toroidal plasmas is reviewed. It is shown how the physics performance, constrained by technological and economic realities, determines the form of reference toroidal reactors. A comparative study of example reactors is not made, because the level of confidence in projections of their performance varies widely, reflecting the vastly different levels of support which each has received. Success with the tokamak has led to the initiation of the International Thermonuclear Experimental Reactor project. It is designed to produce 1500 MW of fusion power from a deuterium-tritium plasma for pulses of 1000 s or longer and to demonstrate the integration of the plasma and nuclear technologies needed for a demonstration reactor.

John Sheffield

1994-07-01T23:59:59.000Z

149

EGS Demonstration | Open Energy Information  

Open Energy Info (EERE)

source source History View New Pages Recent Changes All Special Pages Semantic Search/Querying Get Involved Help Apps Datasets Community Login | Sign Up Search Page Edit History Facebook icon Twitter icon » EGS Demonstration Jump to: navigation, search Geothermal ARRA Funded Projects for EGS Demonstration Loading map... {"format":"googlemaps3","type":"ROADMAP","types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"limit":200,"offset":0,"link":"all","sort":[""],"order":[],"headers":"show","mainlabel":"","intro":"","outro":"","searchlabel":"\u2026

150

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....

151

Handbook of Reactor Physics  

Science Journals Connector (OSTI)

... THIS handbook is one volume in a series sponsored by the United States Atomic Energy Commission with ... data and reference information in the field of reactors. The volume is devoted to reactor physics and radiation shielding, the latter subject occupying approximately a quarter of the book.

PETER W. MUMMERY

1956-08-25T23:59:59.000Z

152

Fast reactor safety  

Science Journals Connector (OSTI)

... SIR, - In his article on fast reactor safety (26 July, page 270) Norman Dombey claims to introduce to non-specialists ... , page 270) Norman Dombey claims to introduce to non-specialists some features of fast reactors that are not available outside the technical literature. The non-specialist would do well ...

R.D. SMITH

1979-08-23T23:59:59.000Z

153

Instrumentation of Nuclear Reactors  

Science Journals Connector (OSTI)

... s Lecture Theatre on January 8, a symposium of papers on the instrumentation of nuclear reactors was organized, at which about five hundred members and visitors attended, including guests from ... the Institution, took the chair and introduced Sir John Cockcroft, whose lecture on "Nuclear Reactors and their Applications" provided a general background for the three specialized papers which followed. ...

1953-03-07T23:59:59.000Z

154

Nuclear Research Reactors  

Science Journals Connector (OSTI)

... their countries for the advent of nuclear power. A few countries had built large research reactors for the production of isotopes and to study the behaviour of nuclear fuel, but ... production of isotopes and to study the behaviour of nuclear fuel, but the small training reactor had not been developed. Since then, research ...

T. E. ALLIBONE

1963-07-20T23:59:59.000Z

155

Neutral reactors on shunt compensated EHV lines  

SciTech Connect

This paper examines the applications of a neutral reactor in limiting resonance overvoltages induced on deenergized conductors due to parallel energized circuits and stuck breaker conditions. These applications are demonstrated through the planned 243 mile long Mead-Phoenix 500 kV line running on the same right of way as the existing Mead-Liberty 345 kV line. Reducing the secondary arc current during single pole reclosing is also examined. In addition to its applications, a procedure for sizing, rating and protection of the neutral reactor is explained.

Atmuri, S.R. [Teshmont Consultants Inc., Winnipeg, Manitoba (Canada); Thallam, R.S.; Gerlach, D.W.; Lundquist, T.G. [Salt River Project, Phoenix, AZ (United States); Selin, D.A. [Arizona Public Service Co., Phoenix, AZ (United States)

1994-12-31T23:59:59.000Z

156

Distributed computing and nuclear reactor analysis  

SciTech Connect

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

157

Canadian university research reactors  

SciTech Connect

In Canada there are seven university research reactors: one medium-power (2-MW) swimming pool reactor at McMaster University and six low-power (20-kW) SLOWPOKE reactors at Dalhousie University, Ecole Polytechnique, the Royal Military College, the University of Toronto, the University of Saskatchewan, and the University of Alberta. This paper describes primarily the McMaster Nuclear Reactor (MNR), which operates on a wider scale than the SLOWPOKE reactors. The MNR has over a hundred user groups and is a very broad-based tool. The main applications are in the following areas: (1) neutron activation analysis (NAA); (2) isotope production; (3) neutron beam research; (4) nuclear engineering; (5) neutron radiography; and (6) nuclear physics.

Ernst, P.C.; Collins, M.F.

1989-11-01T23:59:59.000Z

158

Demonstration and Deployment Strategy Workshop | Department of...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Demonstration and Deployment Strategy Workshop Demonstration and Deployment Strategy Workshop The Bioenergy Technologies Office's (BETO's) Demonstration and Deployment Strategy...

159

Reactor & Nuclear Systems Publications | ORNL  

NLE Websites -- All DOE Office Websites (Extended Search)

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

160

LIMB demonstration project extension and Coolside demonstration: A DOE assessment  

SciTech Connect

The goal of the US Department of Energy (DOE) Clean Coal Technology (CCT) program is to furnish the energy marketplace with a number of advanced, more efficient, and environmentally responsible coal utilization technologies through demonstration projects. These projects seek to establish the commercial feasibility of the most promising advanced coal technologies that have already reached the proof-of-concept stage. This document serves as a DOE post-project assessment of the CCT Round 1 project ``LIMB Demonstration Project Extension and Coolside Demonstration'', described in a report to Congress (Babcock and Wilcox 1987), a paper by DePero et al. (1992), and in a report by Goots et al. (1992). The original limestone injection multistage burner (LIMB) demonstration work was conducted by Babcock and Wilcox Company (B and W) beginning in 1984, under the sponsorship of the US Environmental Protection Agency (EPA) and the State of Ohio Coal Development Office (OCDO). In 1987, B and W and the Ohio Edison Company agreed to extend the full-scale demonstration of LIMB technology under the sponsorship of DOE through its CCT Program, and with support from OCDO and Consolidation Coal Company, now known as CONSOL. In a separate effort, CONSOL had been developing another flue gas desulfurization (FGD) technology known as the Coolside process. Both LIMB and Coolside use sorbent injection to remove SO{sub 2}. The LIMB process injects the sorbent into the furnace and the Coolside injects the sorbent into the flue gas duct. In addition, LIMB uses low-NO{sub x} burners to reduce NO{sub x} emissions; hence it is categorized as a combination SO{sub 2}/NO{sub x} control technology. To take advantage of synergism between the two processes, the CCT project was structured to incorporate demonstration of both the LIMB and Coolside processes. Coolside testing was accomplished between July 1989 and February 1990, and the LIMB Extension test program was conducted between April 1990 and August 1991. The host site for both tests was the 105 MWe coal-fired Unit 4 at Ohio Edison's Edgewater Station in Lorain, Ohio. The major performance objectives of this project were successfully achieved, with SO{sub 2} emissions reductions of up to 70% demonstrated in both processes.

National Energy Technology Laboratory

2000-04-30T23:59:59.000Z

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

Engine ground demonstration test approach  

SciTech Connect

The hardware portion of the current phase of the Integrated Solar Upper Stage (ISUS) program culminates in a system ground demonstration test. The potential application of ISUS technology to a wide array of future missions complicates the process of selecting from among demonstration system design options and test approaches. The approach to this system demonstration has been to maximize system technology readiness level for the entire array of potential missions within the constraints of the program. To this end, system design and test operations planning has been carried out with a premium on demonstrating those elements of the system common to all missions. In addition, test planning has been managed to allow margin for testing those portions of the system envelope needed to confirm acceptable operation for scenarios within the mission set that are specific to a given mission or mission type. Examples drawn from the specific Engine Ground Demonstration (EGD) design selections are used to illuminate this approach, with the result that the EGD system design is not only described, but the reasons for its particular characteristics are made evident.

Kudija, C.T. [Rockwell Aerospace, Canoga Park, CA (United States). Rocketdyne Div.

1996-12-31T23:59:59.000Z

162

Buried Waste Integrated Demonstration Plan  

SciTech Connect

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

163

Shallow Carbon Sequestration Demonstration Project  

NLE Websites -- All DOE Office Websites (Extended Search)

Shallow Carbon SequeStration Shallow Carbon SequeStration DemonStration ProjeCt Background The Shallow Carbon Sequestration Pilot Demonstration Project is a cooperative effort involving City Utilities of Springfield (CU); Missouri Department of Natural Resources (MDNR); Missouri State University (MSU); Missouri University of Science & Technology (MS&T); AmerenUE; Aquila, Inc.; Associated Electric Cooperative, Inc.; Empire District Electric Company; and Kansas City Power & Light. The purpose of this project is to assess the feasibility of carbon sequestration at Missouri power plant sites. The six electric utilities involved in the project account for approximately 90 percent of the electric generating capacity in Missouri. Description The pilot demonstration will evaluate the feasibility of utilizing the Lamotte and

164

QuickPEP Tool Demonstration  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

QuickPEP Tool Demonstration QuickPEP Tool Demonstration Riyaz Papar, PE, CEM Director, Energy Assets & Optimization Hudson Technologies Company William Orthwein, CEM US Department of Energy February 26, 2009 Agenda * Introduction * Plant Energy Profiling * QuickPEP Demonstration * New features in Quick 2.0 * Wrap Up * There are different levels of Plant Energy Profiling - 10,000 ft level - Overall Plant * Phone interview * 1-day plant walkthrough * Using QuickPEP - 1,000 ft level - System level * Gap Analysis (Qualitative only) * 1-day plant walkthrough * 3-day plant Energy Savings Assessments (ESA) * Using US DOE BestPractices System Tools Plant Energy Profiling 10,000 ft approach - The Big Picture in your Plant * Looking at the forest first - Understanding your plant from an energy supply & demand perspective

165

Propane Vehicle Demonstration Grant Program  

SciTech Connect

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

166

MDF | Manufacturing Demonstration Facility | ORNL  

NLE Websites -- All DOE Office Websites (Extended Search)

BTRIC CNMS CSMB CFTF HFIR MDF Working with MDF NTRC OLCF SNS Titanium robotic hand holding sphere fabricated using additive manufacturing Home | User Facilities | MDF MDF | Manufacturing Demonstration Facility SHARE As the nation's premier research laboratory, ORNL is one of the world's most capable resources for transforming the next generation of scientific discovery into solutions for rebuilding and revitalizing America's manufacturing industries. Manufacturing industries engage ORNL's expertise in materials synthesis, characterization, and process technology to reduce technical risk and validate investment for innovations targeting products of the future. DOE's Manufacturing Demonstration Facility, established at ORNL, helps industry adopt new manufacturing technologies to reduce life-cycle energy

167

MDF | Manufacturing Demonstration Facility | ORNL  

NLE Websites -- All DOE Office Websites (Extended Search)

Working with MDF Working with MDF Titanium robotic hand holding sphere fabricated using additive manufacturing Home | User Facilities | MDF MDF | Manufacturing Demonstration Facility SHARE As the nation's premier research laboratory, ORNL is one of the world's most capable resources for transforming the next generation of scientific discovery into solutions for rebuilding and revitalizing America's manufacturing industries. Manufacturing industries engage ORNL's expertise in materials synthesis, characterization, and process technology to reduce technical risk and validate investment for innovations targeting products of the future. DOE's Manufacturing Demonstration Facility, established at ORNL, helps industry adopt new manufacturing technologies to reduce life-cycle energy

168

2012_AdvReactor_Factsheet.indd  

NLE Websites -- All DOE Office Websites (Extended Search)

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

169

LIMB Demonstration Project Extension and Coolside Demonstration. [Final report  

SciTech Connect

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

170

Neutron-Proton Exchange Demonstrated  

Science Journals Connector (OSTI)

Neutron-Proton Exchange Demonstrated ... EVIDENCE of the exchange of charge between protons and neutrons has recently been obtained from studies in the high power cyclotron, according to Ernest O. Lawrence, professor of physics at the University of California a* Berkeley. ...

1947-11-03T23:59:59.000Z

171

Clean Coal Technology Demonstration Program  

Energy.gov (U.S. Department of Energy (DOE))

The Office of Fossil Energy’s Clean Coal Technology Demonstration Program (1986-1993) laid the foundation for effective technologies now in use that have helped significantly lower emissions of sulfur dioxide (SO2), nitrogen oxides (NOx) and airborne particulates (PM10).

172

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

173

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

174

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

175

International Stationary Fuel Cell Demonstration  

NLE Websites -- All DOE Office Websites (Extended Search)

STATIONARY FUEL CELL DEMONSTRATION STATIONARY FUEL CELL DEMONSTRATION John Vogel, Plug Power Inc. Yu-Min Tsou, PEMEAS E-TEK 14 February, 2007 Clean, Reliable On-site Energy SAFE HARBOR STATEMENT This presentation contains forward-looking statements, including statements regarding the company's future plans and expectations regarding the development and commercialization of fuel cell technology. All forward-looking statements are subject to risks, uncertainties and assumptions that could cause actual results to differ materially from those projected. The forward-looking statements speak only as of the date of this presentation. The company expressly disclaims any obligation or undertaking to release publicly any updates or revisions to any such statements to reflect any change in the company's expectations or any change in

176

Electric thermal storage demonstration program  

SciTech Connect

In early 1989, MMWEC, a joint action agency comprised of 30 municipal light departments in Massachusetts and on affiliate in Rhode Island, responded to a DOE request to proposal for the Least Cost Utility Planning program. The MMWEC submission was for the development of a program, focused on small rural electric utilities, to promote the use of electric thermal storage heating systems in residential applications. This report discusses the demonstration of ETS equipment at four member light departments.

Not Available

1992-01-01T23:59:59.000Z

177

Electric thermal storage demonstration program  

SciTech Connect

In early 1989, MMWEC, a joint action agency comprised of 30 municipal light departments in Massachusetts and on affiliate in Rhode Island, responded to a DOE request to proposal for the Least Cost Utility Planning program. The MMWEC submission was for the development of a program, focused on small rural electric utilities, to promote the use of electric thermal storage heating systems in residential applications. This report discusses the demonstration of ETS equipment at four member light departments.

Not Available

1992-02-01T23:59:59.000Z

178

Reactor Thermal-Hydraulics  

NLE Websites -- All DOE Office Websites (Extended Search)

Thermal-Hydraulics Thermal-Hydraulics Dr. Tanju Sofu, Argonne National Laboratory In a power reactor, the energy produced in fission reaction manifests itself as heat to be removed by a coolant and utilized in a thermodynamic energy conversion cycle to produce electricity. A simplified schematic of a typical nuclear power plant is shown in the diagram below. Primary coolant loop Steam Reactor Heat exchanger Primary pump Secondary pump Condenser Turbine Water Although this process is essentially the same as in any other steam plant configuration, the power density in a nuclear reactor core is typically four orders of magnitude higher than a fossil fueled plant and therefore it poses significant heat transfer challenges. Maximum power that can be obtained from a nuclear reactor is often limited by the

179

Reactor hot spot analysis  

SciTech Connect

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

180

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

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

Advanced Reactor Research and Development Funding Opportunity Announcement  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

182

Advanced Reactor Research and Development Funding Opportunity Announcement  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

183

Small Modular Reactors Presentation to Secretary of Energy Advisory Board -  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

184

Development of Materials for Supercritical-Water-Cooled Reactor |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Development of Materials for Supercritical-Water-Cooled Reactor Development of Materials for Supercritical-Water-Cooled Reactor Development of Materials for Supercritical-Water-Cooled Reactor Supercritical-Water-Cooled Reactor (SCWR) was selected as one of the promising candidates in Generation IV reactors for its prominent advantages; those are the high thermal efficiency, the system simplification, the R&D cost minimization and the flexibility for core design. As the demand for advanced nuclear system increases, Japanese R&D project started in 1999 aiming to provide technical information essential to demonstration of SCPR technologies through three sub-themes of 1. Plant conceptual design, 2. Thermal-hydraulics, and 3. Material. Although the material development is critical issue of SCWR development, previous studies were limited for the screening tests on commercial alloys

185

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

186

F Reactor Inspection  

SciTech Connect

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

187

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Initial Modeling of a Pressurized Water Reactor Completed Using Initial Modeling of a Pressurized Water Reactor Completed Using RELAP-7 Initial Modeling of a Pressurized Water Reactor Completed Using RELAP-7 January 29, 2013 - 12:06pm Addthis Schematic of the OECD PWR benchmark used in the initial RELAP-7 demonstration Schematic of the OECD PWR benchmark used in the initial RELAP-7 demonstration RELAP-7 is a nuclear reactor system safety analysis code. Development started in October 2011, and during the past quarter the initial capabilities of RELAP-7 were demonstrated by simulating a steady-state single-phase pressurized water reactor (PWR) with two parallel loops and multiple reactor core flow channels (Fig. 1). The PWR configuration matched that of the Three Mile Island 1 LWR, which is a benchmark problem from the

188

Microsoft Word - RISMC ATR Case Study - Final.docx  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

2-27015 2-27015 Revision 0 Light Water Reactor Sustainability Program Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study August 2012 DOE Office of Nuclear Energy DISCLAIMER This information was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness, of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. References herein to any specific commercial product, process, or service by trade name, trade mark,

189

Reliability analysis of a passive cooling system using a response surface with an application to the Flexible Conversion Ratio Reactor .  

E-Print Network (OSTI)

??A comprehensive risk-informed methodology for passive safety system design and performance assessment is presented and demonstrated on the Flexible Conversion Ratio Reactor (FCRR). First, the… (more)

Fong, Christopher J. (Christopher Joseph)

2008-01-01T23:59:59.000Z

190

U.S. Department of Energy Provides Report to Congress on the Demonstration  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Provides Report to Congress on the Provides Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel from Decommissioned Nuclear Power Reactor Sites U.S. Department of Energy Provides Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel from Decommissioned Nuclear Power Reactor Sites December 9, 2008 - 8:51am Addthis Washington D.C. - The U.S. Department of Energy (DOE) today released its Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel from Decommissioned Nuclear Power Reactor Sites (DOE/RW-0596, December 2008). The report was prepared pursuant to direction in the House Appropriations Committee Report that accompanied the Consolidated Appropriations Act of 2008 that the Department develop a plan to take

191

A Compact Nuclear Fusion Reactor for Space Flights  

SciTech Connect

A small-scale nuclear fusion reactor is suggested based on the concepts of plasma confinement (with a high pressure gas) which have been patented by the author. The reactor considered can be used as a power setup in space flights. Among the advantages of this reactor is the use of a D3He fuel mixture which at burning gives main reactor products -- charged particles. The energy balance considerably improves, as synchrotron radiation turn out 'captured' in the plasma volume, and dangerous, in the case of classical magnetic confinement, instabilities in the direct current magnetic field configuration proposed do not exist. As a result, the reactor sizes are quite suitable (of the order of several meters). A possibility of making reactive thrust due to employment of ejection of multiply charged ions formed at injection of pellets from some adequate substance into the hot plasma center is considered.

Nastoyashchiy, Anatoly F. [SRC Troitsk Institute for Innovation and Fusion Research, TRINITI 142190 Troitsk Moscow Reg. (Russian Federation)

2006-05-02T23:59:59.000Z

192

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

193

Oversight Reports - West Valley Demonstration Project | Department...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

West Valley Demonstration Project Oversight Reports - West Valley Demonstration Project August 24, 2012 Independent Activity Report, West Valley Demonstration Project - July 2012...

194

REACTOR GROUT THERMAL PROPERTIES  

SciTech Connect

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

195

Reactor Safety Planning for Prometheus Project, for Naval Reactors Information  

SciTech Connect

The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

P. Delmolino

2005-05-06T23:59:59.000Z

196

B Reactor | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Operational Management » History » Manhattan Project » Signature Operational Management » History » Manhattan Project » Signature Facilities » B Reactor B Reactor B Reactor Completed in September 1944, the B Reactor was the world's first large-scale plutonium production reactor. As at Oak Ridge, the need for labor turned Hanford into an atomic boomtown, with the population reaching 50,000 by summer 1944. Similar to the X-10 Graphite Reactor at Oak Ridge in terms of loading and unloading fuel, the B Reactor was built on a much larger scale and used water rather than air as a coolant. Whereas the X-10 had an initial design output of 1,000 kilowatts, the B Reactor was designed to operate at 250,000 kilowatts. Consisting of a 28- by 36-foot, 1,200-ton graphite cylinder lying on its side, the reactor was penetrated through its

197

Microchannel Reactor System for Catalytic Hydrogenation  

SciTech Connect

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

198

Advanced Coal Conversion Process Demonstration  

NLE Websites -- All DOE Office Websites (Extended Search)

Clean Coal Technology Program Clean Coal Technology Program Advanced Coal Conversion Process Demonstration A DOE Assessment DOE/NETL-2005/1217 U.S. Department of Energy Office of Fossil Energy National Energy Technology Laboratory April 2005 2 Disclaimer This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference therein to any specific commercial product, process, or service by trade name,

199

UDC Demonstrates Phosphorescent OLED Systems  

Energy.gov (U.S. Department of Energy (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

200

Clean Coal Diesel Demonstration Project  

SciTech Connect

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

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

Demonstration of integrated optimization software  

SciTech Connect

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

202

Hydrogen Storage Materials Database Demonstration  

NLE Websites -- All DOE Office Websites (Extended Search)

| Fuel Cell Technologies Program Source: US DOE 4/25/2011 eere.energy.gov | Fuel Cell Technologies Program Source: US DOE 4/25/2011 eere.energy.gov Hydrogen Storage Materials Database Demonstration FUEL CELL TECHNOLOGIES PROGRAM Ned Stetson Storage Tech Team Lead Fuel Cell Technologies Program U.S. Department of Energy 12/13/2011 Hydrogen Storage Materials Database Marni Lenahan December 13, 2011 Database Background * The Hydrogen Storage Materials Database was built to retain information from DOE Hydrogen Storage funded research and make these data more accessible. * Data includes properties of hydrogen storage materials investigated such as synthesis conditions, sorption and release conditions, capacities, thermodynamics, etc. http://hydrogenmaterialssearch.govtools.us Current Status * Data continues to be collected from DOE funded research.

203

Performance Demonstration Program Management Plan  

SciTech Connect

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

204

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

SciTech Connect

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

205

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

206

Solar energy storage: A demonstration experiment  

Science Journals Connector (OSTI)

Solar energy storage: A demonstration experiment ... A demonstration of a phase transition that can be used for heat storage. ...

Howard S. Kimmel; Reginald P. T. Tomkins

1979-01-01T23:59:59.000Z

207

Honeywell Demonstrates Automated Demand Response Benefits for...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

208

Electric Drive Vehicle Demonstration and Vehicle Infrastructure...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

209

Webinar: Hydrogen Storage Materials Database Demonstration |...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

210

Hydrogen Storage Materials Database Demonstration | Department...  

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

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

211

National Hydrogen Learning Demonstration Status | Department...  

Energy Savers (EERE)

Hydrogen Learning Demonstration Status National Hydrogen Learning Demonstration Status Download presentation slides from the Fuel Cell Technologies Program webinar "National...

212

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement (decreased toughness) , as shown in Fig. 1.1, and extending operation from

213

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

214

Diagnostics for hybrid reactors  

SciTech Connect

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

215

Structural materials for fusion reactors  

Science Journals Connector (OSTI)

Fusion Reactors will require specially engineered structural materials, which ... on safety considerations. The fundamental differences between fusion and other nuclear reactors arise due to the 14MeV neutronics ...

P. M. Raole; S. P. Deshpande

2009-04-01T23:59:59.000Z

216

Reactor Materials | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

217

Reactor operation safety information document  

SciTech Connect

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

218

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

219

Transport reactor development status  

SciTech Connect

This project is part of METC`s Power Systems Development Facility (PSDF) located at Wilsonville, Alabama. The primary objective of the Advanced Gasifier module is to produce vitiated gases for intermediate-term testing of Particulate Control Devices (PCDs). The Transport reactor potentially allows particle size distribution, solids loading, and particulate characteristics in the off-gas stream to be varied in a number of ways. Particulates in the hot gases from the Transport reactor will be removed in the PCDs. Two PCDs will be initially installed in the module; one a ceramic candle filter, the other a granular bed filter. After testing of the initial PCDs they will be removed and replaced with PCDs supplied by other vendors. A secondary objective is to verify the performance of a Transport reactor for use in advanced Integrated Gasification Combined Cycle (IGCC), Integrated Gasification Fuel Cell (IG-FC), and Pressurized Combustion Combined Cycle (PCCC) power generation units. This paper discusses the development of the Transport reactor design from bench-scale testing through pilot-scale testing to design of the Process Development Unit (PDU-scale) facility at Wilsonville.

Rush, R.E.; Fankhanel, M.O.; Campbell, W.M.

1994-10-01T23:59:59.000Z

220

Thermal Reactor Safety  

SciTech Connect

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

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

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

222

Direct conversion nuclear reactor space power systems  

SciTech Connect

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

223

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

224

Solution of the space-time reactor kinetics equations using the method of Laplace transforms  

E-Print Network (OSTI)

ANALYTIC CALCULATIONS The analytic solut1on developed in Chapter II is the basis for the computer program ANALYTIC, which calculates the time and spatially dependent solution for the neutron flux and the precursor concentration in a bare slab reactor... case to test the full capabilities of the ANALYTIC computer program is the subcrit1cal reactor. The reactor parameters for the thermal, subcritical and the fast, subcrit1cal reacto~ is listed 1n Table IV. The solution for the neutron flux...

Rottler, Jerry Stephen

2012-06-07T23:59:59.000Z

225

Fusion reactor systems  

Science Journals Connector (OSTI)

In this review we consider deuterium-tritium (D-T) fusion reactors based on four different plasma-confinement and heating approaches: the tokamak, the theta-pinch, the magnetic-mirror, and the laser-pellet system. We begin with a discussion of the dynamics of reacting plasmas and basic considerations of reactor power balance. The essential plasma physical aspects of each system are summarized, and the main characteristics of the corresponding conceptual power plants are described. In tokamak reactors the plasma densities are about 1020 m-3, and the ? values (ratio of plasma pressure to confining magnetic pressure) are approximately 5%. Plasma burning times are of the order of 100-1000 sec. Large superconducting dc magnets furnish the toroidal magnetic field, and 2-m thick blankets and shields prevent heat deposition in the superconductor. Radially diffusing plasma is diverted away from the first wall by means of null singularities in the poloidal (or transverse) component of the confining magnetic field. The toroidal theta-pinch reactor has a much smaller minor diameter and a much larger major diameter, and operates on a 10-sec cycle with 0.1-sec burning pulses. It utilizes shock heating from high-voltage sources and adabatic-compression heating powered by low-voltage, pulsed cryogenic magnetic or inertial energy stores, outside the reactor core. The plasma has a density of about 1022 m-3 and ? values of nearly unity. In the power balance of the reactor, direct-conversion energy obtained by expansion of the burning high-? plasma against the containing magnetic field is an important factor. No divertor is necessary since neutral-gas flow cools and replaces the "spent" plasma between pulses. The open-ended mirror reactor uses both thermal conversion of neutron energy and direct conversion of end-loss plasma energy to dc electrical power. A fraction of this direct-convertor power is then fed back to the ioninjection system to sustain the reaction and maintain the plasma. The average ion energy is 600 keV, plasma diameter 6 m, and the plasma beta 85%. The power levels of the three magnetic-confinement devices are in the 500-2000 MWe range, with the exception of the mirror reactor, for which the output is approximately 200 MWe. In Laser-Pellet reactors, frozen D-T pellets are ignited in a cavity which absorbs the electromagnetic, charged particle, and neutron energy from the fusion reaction. The confinement is "inertial," since the fusion reaction occurs during the disassembly of the heated pellet. A pellet-cavity unit would produce about 200 MWt in pulses with a repetition rate of the order of 10 sec-1. Such units could be clustered to give power plants with outputs in the range of 1000 MWe.

F. L. Ribe

1975-01-01T23:59:59.000Z

226

Independent Activity Report, West Valley Demonstration Project...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

November 2011 Independent Activity Report, West Valley Demonstration Project - November 2011 November 2011 West Valley Demonstration Project Orientation Visit HIAR-WVDP-2011-11-07...

227

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

West Valley Demonstration Project Enforcement Documents - West Valley Demonstration Project December 7, 1999 Preliminary Notice of Violation, West Valley Nuclear Services -...

228

Thanksgiving Goodwill: West Valley Demonstration Project Food...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

229

Independent Activity Report, West Valley Demonstration Project...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

July 2012 Independent Activity Report, West Valley Demonstration Project - July 2012 July 2012 Operational Awareness Oversight of the West Valley Demonstration Project HIAR...

230

Major Demonstrations | netl.doe.gov  

NLE Websites -- All DOE Office Websites (Extended Search)

Major Demonstrations Industrial Carbon Capture and Storage Clean Coal Power Initiative Power Plant Improvement Initiative Clean Coal Technology Demonstration Program FutureGen For...

231

Newberry EGS Demonstration | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Production: Evaluation of Stimulation at the Newberry Volcano EGS Demonstration Site Microearthquake Technology for EGS Fracture Characterization Newberry Volcano EGS Demonstration...

232

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...

233

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...

234

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...

235

Daemen Alternative Energy/Geothermal Technologies Demonstration...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

236

Commercial Building Demonstration and Deployment Overview - 2014...  

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

Demonstration and Deployment Overview - 2014 BTO Peer Review Commercial Building Demonstration and Deployment Overview - 2014 BTO Peer Review Presentation: Kristen Taddonio, U.S....

237

Demonstrating and Deploying Integrated Retrofit Technologies...  

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

Demonstrating and Deploying Integrated Retrofit Technologies and Solutions - 2014 BTO Peer Review Demonstrating and Deploying Integrated Retrofit Technologies and Solutions - 2014...

238

Neutronic analysis of a fusion hybrid reactor  

SciTech Connect

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

239

A future for nuclear energy: pebble bed reactors  

Science Journals Connector (OSTI)

Pebble Bed Reactors could allow nuclear plants to support the goal of reducing global climate change in an energy hungry world. They are small, modular, inherently safe, use a demonstrated nuclear technology and can be competitive with fossil fuels. Pebble bed reactors are helium cooled reactors that use small tennis ball size fuel balls consisting of only 9 grams of uranium per pebble to provide a low power density reactor. The low power density and large graphite core provide inherent safety features such that the peak temperature reached even under the complete loss of coolant accident without any active emergency core cooling system is significantly below the temperature that the fuel melts. This feature should enhance public confidence in this nuclear technology. With advanced modularity principles, it is expected that this type of design and assembly could lower the cost of new nuclear plants removing a major impediment to deployment.

Andrew C. Kadak

2005-01-01T23:59:59.000Z

240

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

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

Nuclear divisional reactor  

SciTech Connect

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

242

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

243

California: Next-Generation Geothermal Demonstration Launched  

Office of Energy Efficiency and Renewable Energy (EERE)

First-of-its-kind achievement successfully demonstrates that EGS technologies are commercially viable.

244

Demonstrating the Safety of Long-Term Dry Storage - 13468  

SciTech Connect

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

245

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

SciTech Connect

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

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

1991-09-01T23:59:59.000Z

246

Combined photonics and MEMs function demonstration  

SciTech Connect

The authors have recently demonstrated two prototypes where photonics and microelectromechanical system (MEMs) technologies have been integrated to show proof-of-principle functionality for weapon surety functions. These activities are part of a program which is exploring the miniaturization of electromechanical components for making weapon systems safer. Such miniaturization can lead to a low-cost, small, high-performance ``systems-on-a-chip``, and have many applications ranging from advanced military systems to large-volume commercial markets like automobiles, rf or land-based communications networks and equipment, or commercial electronics. One of the key challenges in realization of the microsystem is integration of several technologies including digital electronics; analog and rf electronics, optoelectronics (light emitting and detecting devices and circuits), sensors and actuators, and advanced packaging technologies. In this work the authors describe efforts in integrating MEMs and photonic functions and the fabrication constraints on both system components. Here, they discuss two examples of integration of MEMs and a photonic device. In the first instance, a MEMs locking device pin is driven by a voltage generated by photovoltaic cells connected in series, which are driven by a laser. In the second case, a VCSEL emitting at 1.06 {micro}m is packaged together with a metallized MEMs shutter. By appropriate alignment to the opening in the shutter, the VCSEL is turned on and off by the movement of the Si chopper wheel.

Blum, O.; Warren, M.E.; Hou, H.Q.; Choquette, K.D.; Rogers, M.S.; Sniegowski, J.J. [Sandia National Labs., Albuquerque, NM (United States); Carson, R.F. [Microoptical Devices, Inc., Albuquerque, NM (United States)

1998-01-01T23:59:59.000Z

247

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

QUALITY ASSURANCE (QA) QUALITY ASSURANCE (QA) OBJECTIVE QA-1: The RRD QA program has been appropriately modified to reflect the CS modification and its reactor interface, and sufficient numbers of qualified QA personnel are provided to ensure services are adequate to support reactor operation. The QA functions, assignments, responsibilities, and reporting relationships are clearly defined, understood, and effectively implemented with line management control of safety. QA personnel exhibit awareness of the applicable requirements pertaining to reactor operation with the CS and the associated hazards. Through their actions, they have demonstrated a high-priority commitment to comply with these requirements. The level of knowledge of QA personnel related to reactor

248

Concept of nuclear reactor pumped laser for ICF  

SciTech Connect

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

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

1996-05-01T23:59:59.000Z

249

Integral Fast Reactor Program annual progress report, FY 1991  

SciTech Connect

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

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

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

252

Massive Hanford Test Reactor Removed - Plutonium Recycle Test...  

Office of Environmental Management (EM)

Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor removed from Hanford's 300 Area Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor removed...

253

Nuclear Reactor Severe Accident Experiments  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Reactor Severe Accident Experiments Nuclear Reactor Severe Accident Experiments Capabilities Engineering Experimentation Reactor Safety Testing and Analysis Overview Nuclear Reactor Severe Accident Experiments MAX NSTF SNAKE Aerosol Experiments System Components Laser Applications Robots Applications Other Facilities Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Nuclear Reactor Severe Accident Experiments 1 2 3 4 5 6 7 We perform experiments simulating reactor core melt phenomena in which molten core debris ("corium") erodes the concrete floor of a containment building. This occurred during the Fukushima nuclear power plant accident though the extent of concrete damage is yet unknown. This video shows the top view of a churning molten pool of uranium oxide at 2000°C (3600°F) seen during an experiment at Argonne. Corium behaves much like lava.

254

Zircaloy performance in light water reactors  

SciTech Connect

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

255

Nuclear Reactor Materials and Fuels  

Science Journals Connector (OSTI)

Nuclear reactor materials and fuels can be classified into six categories: Nuclear fuel materials Nuclear clad materials Nuclear coolant materials Nuclear poison materials Nuclear moderator materials

Dr. James S. Tulenko

2012-01-01T23:59:59.000Z

256

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

257

Light Water Reactor Sustainability Newsletter  

NLE Websites -- All DOE Office Websites (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)...

258

BURN CONTROL IN FUSION REACTORS VIA NONLINEAR STABILIZATION TECHNIQUES  

E-Print Network (OSTI)

to a decrease of power and quenching. Even during a quenching, a disruptive instability can be reached, causing of auxiliary power and fueling rate are considered as control forces, and the ignition case, where into the plasma center. In this way we make our control system indepen- dent of the fueling system and the reactor

Krstic, Miroslav

259

HTI retrieval demonstration project execution plan  

SciTech Connect

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

260

SPIDERS Joint Capability Technology Demonstration Industry Day...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Technology Demonstration Industry Day May 2, 2014 - 1:15pm Addthis An image of a patch with a spider on it. The Smart Power Infrastructure Demonstration for Energy...

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

Reactor Safety Research Programs Quarterly Report October - December 1981  

SciTech Connect

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

262

Advanced Reactors Thermal Energy Transport for Process Industries  

SciTech Connect

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

263

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

264

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

265

Biparticle fluidized bed reactor  

DOE Patents (OSTI)

A fluidized bed reactor system is described which 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 and tertiary particulate phases, continuously introduced and removed simultaneously in the cocurrent and countercurrent mode, act in a role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Means for introducing and removing the sorbent phases include feed screw mechanisms and multivane slurry valves. 3 figs.

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

1996-02-27T23:59:59.000Z

266

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

267

Ground Source Heat Pump Demonstration Projects  

Energy.gov (U.S. Department of Energy (DOE))

Below are the project presentations and respective peer review results for Ground Source Heat Pump Demonstration Projects.

268

Webinar: National Hydrogen Learning Demonstration Status  

Energy.gov (U.S. Department of Energy (DOE))

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

269

Webinar: Hydrogen Storage Materials Database Demonstration  

Energy.gov (U.S. Department of Energy (DOE))

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

270

Systems Integration Research, Development, and Demonstration...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

activities are focused on these key research, development, and demonstration areas: Distribution Grid Integration Transmission Grid Integration Solar Resource Assessment...

271

Sapphire Energy, Inc. Demonstration-Scale Project  

Energy.gov (U.S. Department of Energy (DOE))

Sapphire Energy, Inc. is scaling up an operational facility to demonstrate conversion of algal carbon dioxide to green crude oil.

272

Proliferation resistance of small modular reactors fuels  

SciTech Connect

In this paper the proliferation resistance of different types of Small Modular Reactors (SMRs) has been examined and classified with criteria available in the literature. In the first part of the study, the level of proliferation attractiveness of traditional low-enriched UO{sub 2} and MOX fuels to be used in SMRs based on pressurized water technology has been analyzed. On the basis of numerical simulations both cores show significant proliferation risks. Although the MOX core is less proliferation prone in comparison to the UO{sub 2} core, it still can be highly attractive for diversion or undeclared production of nuclear material. In the second part of the paper, calculations to assess the proliferation attractiveness of fuel in typical small sodium cooled fast reactor show that proliferation risks from spent fuel cannot be neglected. The core contains a highly attractive plutonium composition during the whole life cycle. Despite some aspects of the design like the sealed core that enables easy detection of unauthorized withdrawal of fissile material and enhances proliferation resistance, in case of open Non-Proliferation Treaty break-out, weapon-grade plutonium in sufficient quantities could be extracted from the reactor core.

Polidoro, F.; Parozzi, F. [RSE - Ricerca sul Sistema Energetico,Via Rubattino 54, 20134, Milano (Italy); Fassnacht, F.; Kuett, M.; Englert, M. [IANUS, Darmstadt University of Technology, Alexanderstr. 35, D-64283 Darmstadt (Germany)

2013-07-01T23:59:59.000Z

273

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

274

Hybrid Plasma Reactor/Filter for Transportable Collective Protection Systems  

SciTech Connect

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

275

An improved weighted average reactor temperature estimation for simulation of adiabatic industrial hydrotreaters  

Science Journals Connector (OSTI)

A study on the improvement of the representative operating temperature from the temperature profile of an industrial adiabatic reactor is presented. This temperature is used to simulate the reactor performance by small scale laboratory isothermal reactors. An improved methodology for the estimation of a Weighted Average Bed Temperature (WABT) was elaborated to simulate an industrial multi-bed HDS reactor. The improved WABT, so called Weighted Average Reactor Temperature (WART), was compared with the most usually used WABT in a wide range of operational conditions as well as of kinetic parameters. In case of a multi-bed industrial hydrotreater, where quench zones are located between the beds and the H2 flow rate, which enters each bed, is different, the optimal gas to oil ratio was estimated for the laboratory-scale reactor.

G.D. Stefanidis; G.D. Bellos; N.G. Papayannakos

2005-01-01T23:59:59.000Z

276

Kimberlina: a zero-emissions demonstration plant  

SciTech Connect

FutureGen may be getting the headlines, but it is not the only superclean demonstration plant in town. In fact, you could argue that other technologies are further down the evolutionary timeline. Case in point: Clean Energy Systems' adaptation of rocket engine technology to radically change the way fuel is burned. The result is a true zero-emissions power plant. Its most distinctive element is an oxy-combustor, similar to one used in rocket engines, that generates steam by burning clean, gaseous fuel in the presence of gaseous oxygen and water. The clean fuel is prepared by processing a conventional fossil fuel such as coal-derived syngas, refinery residues, biomass or biodigester gas, or natural or landfill gas. Combustion takes place at near-stoichiometric conditions to produce a mixture of steam and CO{sub 2} at high temperature and pressure. The steam conditions are suitable for driving a conventional or advanced steam turbine-generator, or a gas turbine modified to be driven by high-temperature steam or to do work as an expansion unit at intermediate pressure. After pressure through the turbine(s), the steam/CO{sub 2} mixture is condensed, cooled, and separated into water and CO{sub 2}. The CO{sub 2} can be sequestered and/or purified and sold for commercial use. Durability and performance tests carried out between March 2005 and March 2006 produced excellent results. CO and NOx emissions are considerably low than those of combined-cycle power plants fuelled by natural gas and using selective catalytic reduction for NOx control. Work is continuing under an NETL grant. Progress and plans are reported in the article. 7 figs.

Pronske, K. [Clean Energy Systems Inc. (USA)

2007-06-15T23:59:59.000Z

277

Independent Oversight Review, West Valley Demonstration Project  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Independent Oversight Review, West Valley Demonstration Project Independent Oversight Review, West Valley Demonstration Project Transportation - September 2000 Independent Oversight Review, West Valley Demonstration Project Transportation - September 2000 September 2000 Transportation Emergency Management Review of the West Valley Demonstration Project (WVDP) and National Transportation Program (NTP)/Transportation Compliance Evaluation/Assistance Program (TCEAP) The U.S. Department of Energy (DOE) Office of Emergency Management Oversight, within the Secretary of Energy's Office of Independent Oversight and Performance Assurance, conducted a transportation emergency management review of the West Valley Demonstration Project (WVDP) and National Transportation Program (NTP)/Transportation Compliance Evaluation/Assistance Program (TCEAP) in September 2000.

278

DOE Drops Plan to Restart Reactor  

Science Journals Connector (OSTI)

...longer in flux. Hanford research reactor...decision to scrap the Hanford reactor, which...research. At public meetings, however...decision to scrap the Hanford reactor, which...research. At public meetings, however, FFTF...

Robert F. Service

2000-12-01T23:59:59.000Z

279

Operational Analysis of Multiregional Nuclear Reactor Kinetics  

Science Journals Connector (OSTI)

......Operational Analysis of Multiregional Nuclear Reactor Kinetics NASSAR H. S. HAIDAR...analytically for a multiregional nuclear reactor whose subregions are of arbitrary...Operational Analysis of Multiregional Nuclear Reactor Kinetics NASSAU H. S. HAIDAR......

NASSAR H. S. HAIDAR

1983-05-01T23:59:59.000Z

280

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

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

Dynamics of reverse osmosis in a standalone cogenerative nuclear reactor (Part I: reactivity changes)  

Science Journals Connector (OSTI)

The present study considers the dynamic behaviour of the pressurised water reactor safety features, represented by the integrity of the fuel cladding, under some transient cases. A cosine-shaped heating through the fuel is taken with the corresponding coolant lumps, to simulate realistic cases encountered in nuclear reactors. A mathematical model was developed for the Westinghouse 3411 MWth pressurised water reactor, as an example of a familiar design with predominantly published data design. The model consists of two parts. The first part is concerned with the dynamics of the primary side of the reactor, which is described in this paper. The second part is concerned with the secondary side of the plant, which is described elsewhere in this issue. To study the dynamics of the reactor, a model of 17 lumped parameters was used, consisting of first-order differential equations deduced from the first principles considering six groups of delayed neutrons. A computer program was developed using the Runge-Kutta method to solve these equations and to predict the behaviour of the state variables with time. Two case studies were considered as examples for normal transients. The first case study, which represents Part 1 of this study, considers the effect of primary side transient on the system as the reactivity changes. Reactor reactivity changes, including movements of the reactor control rods, which are taken as an example for the effect of the reactor primary side conditions. These reactivity changes vary from 0.0005 up to 0.003, both for positive and negative reactivity. The results of the developed model, which describe the dynamic response of the reactor primary circuit, have been analysed and verified with the relevant models. These results indicate that the reactor components and the integrity of the fuel cladding were attained during different step changes of reactivity.

Aly Karameldin; M.M. Shamloul; M.R. Shaalan; M.H. Esawy

2007-01-01T23:59:59.000Z

282

Implications of Fast Reactor Transuranic Conversion Ratio  

SciTech Connect

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

283

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

284

Nuclear reactors in the United States  

Science Journals Connector (OSTI)

Nuclear reactors in the United States ... A chart listing the operating and planned nuclear reactors in the United States. ... Nuclear / Radiochemistry ...

Hubert N. Alyea

1956-01-01T23:59:59.000Z

285

Advanced Reactor Research and Development Funding Opportunity...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

286

LIGHT WATER REACTOR SUSTAINABILITY PROGRAM: INTRODUCTION  

NLE Websites -- All DOE Office Websites (Extended Search)

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

287

MOOSE simulating nuclear reactor CRUD buildup  

SciTech Connect

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

288

Advanced Reactor Technologies | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

289

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

290

F Reactor Inspection | Department of Energy  

NLE Websites -- All DOE Office Websites (Extended Search)

Inspection F Reactor Inspection Addthis Description Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor last week before...

291

Physics of nuclear reactor safety  

Science Journals Connector (OSTI)

Provides a concise review of the physical aspects of safety of nuclear fission reactors. It covers the developments of roughly the last decade. The introductory chapter contains an analysis of the changes in safety philosophy that are characteristic of the last decade and that have given rise to an increased importance of physical aspects because of the emphasis on passive or natural safety. The second chapter focuses on the basics of reactor safety, identifying the main risk sources and the main principles for a safe design. The third chapter concerns a systematic treatment of the physical processes that are fundamental for the properties of fission chain reacting processes and the control of those processes. Because of the rather specialized nature of the field of reactor physics, each paragraph contains a very concise description of the theory of the phenomenon under consideration, before presenting a review of the developments. Chapter 4 contains a short review of the thermal aspects of reactor safety, restricted to those aspects that are characteristic of the nuclear reactor field, because thermal hydraulics of fission reactors is not principally different from that of other physical systems. In chapter 5 the consequences of the physics treated in the preceding chapters for the dynamics and safety of actual reactors are reviewed. The systematics of the treatment is mainly based on a division of reactors into three categories according to the type of coolant, which to a large extent determines the safety properties of the reactors. The last chapter contains a physical analysis of the Chernobyl accident that occurred in 1986. The reason for an attempt to give a review of this accident, as complete as possible within the space limits set by the editors, is twofold: the Chernobyl accident is the most severe accident in history and physical properties of the reactor played a decisive role, thereby serving as an illustration of the material of the preceding chapters.

H van Dam

1992-01-01T23:59:59.000Z

292

Axial xenon stability considerations in VVER-1000 reactors  

SciTech Connect

Frequent problems experienced in VVER-1000 reactors with xenon oscillation control have been reported. Modern Pressurized Water Reactors (PWRs) are designed with operational strategies to help avoid such axial xenon oscillations. Uncontrolled xenon oscillations can cause operational problems, requiring frequent operator intervention and leading to power reductions (or plant trips) due to high core peaking factors, thereby reducing overall plant capacity factors. In the worst case, an uncontrolled xenon oscillation can lead to violations of safety requirements on pellet clad interaction (PCI), departure from nucleate boiling or fuel centerline melting if the plant does not have adequate safety protection systems that account for limiting axial power distributions.

Doshi, P.K.; Miller, R.W.

1993-12-31T23:59:59.000Z

293

Categorical Exclusion Determinations: West Valley Demonstration Project |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Valley Demonstration Valley Demonstration Project Categorical Exclusion Determinations: West Valley Demonstration Project Categorical Exclusion Determinations issued by West Valley Demonstration Project. DOCUMENTS AVAILABLE FOR DOWNLOAD July 11, 2013 CX-010718: Categorical Exclusion Determination Replacement Ventilation System for the Main Plant Process Building CX(s) Applied: B6.3 Date: 07/11/2013 Location(s): New York Offices(s): West Valley Demonstration Project December 20, 2012 CX-009527: Categorical Exclusion Determination WVDP-2012-02 Routine Maintenance CX(s) Applied: B1.3 Date: 12/20/2012 Location(s): New York Offices(s): West Valley Demonstration Project August 2, 2012 CX-009528: Categorical Exclusion Determination WVDP-2012-01 WVDP Reservoir Interconnecting Canal Maintenance Activities

294

13 - Generation IV reactor designs, operation and fuel cycle  

Science Journals Connector (OSTI)

Abstract: This chapter looks at Generation IV nuclear reactors, such as the very high-temperature reactor (VHTR), the supercritical water reactor (SCWR), the molten salt reactor (MSR), the sodium-cooled fast reactor (SFR), the lead-cooled fast reactor (LFR) and the gas-cooled fast reactor (GFR). Reactor designs and fuel cycles are also described.

N. Cerullo; G. Lomonaco

2012-01-01T23:59:59.000Z

295

Frontiers in Reactor Engineering  

Science Journals Connector (OSTI)

...supercritical water, chemical looping combustion, carbon dioxide sequestration...adsorption, 100% conversion of syngas to methanol has been demonstrated...reaction (e.g., methane combustion) with an endothermic one...energy storage and clean coal combustion. Successful periodic operation...

Milorad P. Dudukovic

2009-08-07T23:59:59.000Z

296

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...

297

Oak Ridge City Center Technology Demonstration Project  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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...

298

Oak Ridge National Laboratory Manufacturing Demonstration Facility  

NLE Websites -- All DOE Office Websites (Extended Search)

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

299

NASA's Laser Communications Relay Demonstration Project  

Science Journals Connector (OSTI)

This paper provides an overview of NASA's Laser Communications Relay Demonstration Project (LCRD). LCRD will provide two years of continuous high data rate optical communications as a...

Edwards, Bernard L; Fletcher, Andrew

300

Pacific Northwest Smart Grid Demonstration Project  

NLE Websites -- All DOE Office Websites (Extended Search)

5 megawatt lithium-ion battery * Intelligent distribution management * Commercial demand response * Demonstrates renewable integration For More inForMation: Kevin Whitener...

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

Electric Drive Vehicle Demonstration and Vehicle Infrastructure...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

vs. Utility Meter Utilize communication strategies to alter EVSE operation - Demand Response demonstration Approach EVSE Utility HARDWARE DEPLOYMENT 7,871 Level 2...

302

The Smithsonian American Art Museum GATEWAY Demonstration  

Energy.gov (U.S. Department of Energy (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.

303

Learning Demonstration Progress Report -- Spring 2008  

SciTech Connect

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

304

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

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

Capability Technology Demonstration Industry Day Presentations Partnering with Utilities for Energy Efficiency & Security 2010 Smart Grid Peer Review Day Two Morning Presentations...

305

Daemen Alternative Energy/Geothermal Technologies Demonstration...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

306

Technical Demonstration and Economic Validation of Geothermal...  

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

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

307

University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor  

SciTech Connect

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

308

NEUTRON RADIOGRAPHY (NRAD) REACTOR 64-ELEMENT CORE UPGRADE  

SciTech Connect

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

309

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

SciTech Connect

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

310

Assessment of torsatrons as reactors  

SciTech Connect

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R[sub 0] = 6.6-8.8 m, on-axis magnetic field B[sup 0] = 4.8-7.5 T, B[sub max] (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions.

Lyon, J.F. (Oak Ridge National Lab., TN (United States)) [Oak Ridge National Lab., TN (United States); Painter, S.L. (Australian National Univ., Canberra, ACT (Australia)) [Australian National Univ., Canberra, ACT (Australia)

1992-12-01T23:59:59.000Z

311

Assessment of torsatrons as reactors  

SciTech Connect

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R{sub 0} = 6.6-8.8 m, on-axis magnetic field B{sup 0} = 4.8-7.5 T, B{sub max} (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions.

Lyon, J.F. [Oak Ridge National Lab., TN (United States); Painter, S.L. [Australian National Univ., Canberra, ACT (Australia)

1992-12-01T23:59:59.000Z

312

Model-to-model transformations by demonstration  

Science Journals Connector (OSTI)

During the last decade several approaches have been proposed for easing the burden of writing model transformation rules manually. Among them are Model Transformation By-Demonstration (MTBD) approaches which record actions performed on example models ... Keywords: by-demonstration, by-example, model transformations

Philip Langer; Manuel Wimmer; Gerti Kappel

2010-06-01T23:59:59.000Z

313

Oak Ridge Manufacturing Demonstration Facility (MDF)  

Office of Energy Efficiency and Renewable Energy (EERE)

The Manufacturing Demonstration Facility (MDF) is a collabora­tive manufacturing community that shares a common RD&D infrastructure. This shared infrastructure provides affordable access to advanced physical and virtual tools for rapidly demonstrating new manufacturing technologies and optimizing critical processes. Oak Ridge National Laboratory is home to AMO's MDF focused on Additive Manufacturing and Low-cost Carbon Fiber.

314

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.

315

Manufacturing Demonstration Facility (MDF) | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Manufacturing Demonstration Facility (MDF) Manufacturing Demonstration Facility (MDF) Manufacturing Demonstration Facility (MDF) October 11, 2013 - 9:44am Addthis The Manufacturing Demonstration Facility (MDF) is a collabora-tive manufacturing community that shares a common RD&D infrastructure. This shared infrastructure provides affordable access to advanced physical and virtual tools for rapidly demonstrating new manufacturing technologies and optimizing critical processes. Oak Ridge National Laboratory is home to AMO's MDF focused on Additive Manufacturing and Low-cost Carbon Fiber. Fostering Collaboration to Accelerate Progress Work conducted by MDF partners and users provides real data that is used to reduce the technical risk associated with full commercialization of promising foundational manufacturing process and materials innovations. The

316

Milestone Project Demonstrates Innovative Mercury Emissions Reduction  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Milestone Project Demonstrates Innovative Mercury Emissions Milestone Project Demonstrates Innovative Mercury Emissions Reduction Technology Milestone Project Demonstrates Innovative Mercury Emissions Reduction Technology January 12, 2010 - 12:00pm Addthis Washington, DC - An innovative technology that could potentially help some coal-based power generation facilities comply with anticipated new mercury emissions standards was successfully demonstrated in a recently concluded milestone project at a Michigan power plant. Under a cooperative agreement with the U.S. Department of Energy's (DOE's) National Energy Technology Laboratory (NETL), WE Energies demonstrated the TOXECON(TM) process in a $52.9million project at the Presque Isle Power Plant in Marquette, Mich. TOXECON is a relatively cost-effective option for achieving significant reductions in mercury emissions and increasing the

317

Software demonstration: Demand Response Quick Assessment Tool  

NLE Websites -- All DOE Office Websites (Extended Search)

Software demonstration: Demand Response Quick Assessment Tool Software demonstration: Demand Response Quick Assessment Tool Speaker(s): Peng Xu Date: February 4, 2008 - 12:00pm Location: 90-3122 The potential for utilizing building thermal mass for load shifting and peak demand reduction has been demonstrated in a number of simulation, laboratory, and field studies. The Demand Response Quick Assessment Tools developed at LBNL will be demonstrated. The tool is built on EnergyPlus simulation and is able to evaluate and compare different DR strategies, such as global temperature reset, chiller cycling, supply air temperature reset, etc. A separate EnergyPlus plotting tool will also be demonstrated during this seminar. Users can use the tool to test EnergyPlus models, conduct parametric analysis, or compare multiple EnergyPlus simulation

318

Buried waste integrated demonstration technology integration process  

SciTech Connect

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

319

Buried waste integrated demonstration technology integration process  

SciTech Connect

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

320

CALDERON COKEMAKING PROCESS/DEMONSTRATION PROJECT  

SciTech Connect

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 studies of LTV's site for the installation of the commercial Demonstration Unit with site specific layouts; Environmental Work; Firm commitments for funding from the private sector; and Federal funding to complement the private contribution.

Albert Calderon

1999-06-23T23:59:59.000Z

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 Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Initial Assessment of Thermal Annealing Needs and Challenges Initial Assessment of Thermal Annealing Needs and Challenges Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement (decreased toughness) , as shown in Fig. 1.1, and extending operation from 40y to 80y implies a doubling of the neutron exposure for the RPV. Thus,

322

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

323

Environmentally assisted cracking of light-water reactor materials  

SciTech Connect

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

324

2012 Annual Report Research Reactor Infrastructure Program  

SciTech Connect

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

Douglas Morrell

2012-11-01T23:59:59.000Z

325

Tritium diagnostics in a fusion reactor  

Science Journals Connector (OSTI)

Methods for controlling tritium in a fusion reactor are reviewed. The characteristic features of the...

A. I. Markin; N. I. Syromyatnikov; A. M. Belov

2010-05-01T23:59:59.000Z

326

Operational Demonstration Program | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Operational Demonstration Program Operational Demonstration Program Operational Demonstration Program < Back Eligibility Commercial Savings Category Heating & Cooling Commercial Heating & Cooling Heating Home Weatherization Commercial Weatherization Sealing Your Home Cooling Construction Design & Remodeling Windows, Doors, & Skylights Ventilation Heat Pumps Appliances & Electronics Commercial Lighting Lighting Manufacturing Buying & Making Electricity Bioenergy Alternative Fuel Vehicles Hydrogen & Fuel Cells Water Energy Sources Solar Water Heating Wind Maximum Rebate $500,000 Program Info Funding Source CEFIA Start Date 2005 State Connecticut Program Type Industry Recruitment/Support Rebate Amount $150,000 - $500,000 Provider Clean Energy Finance and Investment Authority This program is currently closed. Applications were due in February 2012.

327

Combustion synthesis continuous flow reactor  

DOE Patents (OSTI)

The present invention is a reactor for combustion synthesis of inorganic powders. The reactor includes a reaction vessel having a length and a first end and a second end. The reaction vessel further has a solution inlet and a carrier gas inlet. The reactor further has a heater for heating both the solution and the carrier gas. In a preferred embodiment, the reaction vessel is heated and the solution is in contact with the heated reaction vessel. It is further preferred that the reaction vessel be cylindrical and that the carrier gas is introduced tangentially into the reaction vessel so that the solution flows helically along the interior wall of the reaction vessel. As the solution evaporates and combustion produces inorganic material powder, the carrier gas entrains the powder and carries it out of the reactor.

Maupin, Gary D. (Richland, WA); Chick, Lawrence A. (West Richland, WA); Kurosky, Randal P. (Maple Valley, WA)

1998-01-01T23:59:59.000Z

328

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

329

Unique features of space reactors  

SciTech Connect

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

330

Nuclear Reactors and Technology; (USA)  

SciTech Connect

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

331

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

332

Alternate-fuel reactor studies  

SciTech Connect

A number of studies related to improvements and/or greater understanding of alternate-fueled reactors is presented. These studies cover the areas of non-Maxwellian distributions, materials and lifetime analysis, a /sup 3/He-breeding blanket, tritium-rich startup effects, high field magnet support, and reactor operation spanning the range from full D-T operation to operation with no tritium breeding.

Evans, K. Jr.; Ehst, D.A.; Gohar, Y.; Jung, J.; Mattas, R.F.; Turner, L.R.

1983-02-01T23:59:59.000Z

333

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

334

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

335

Hywind 2 Demonstration | Open Energy Information  

Open Energy Info (EERE)

Hywind 2 Demonstration Hywind 2 Demonstration Jump to: navigation, search Name Hywind 2 Demonstration Facility Hywind 2 Demonstration Sector Wind energy Facility Type Offshore Wind Facility Status Proposed Owner Statoil North America Developer Statoil North America Location Atlantic Ocean ME Coordinates 43.524°, -69.534° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":43.524,"lon":-69.534,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

336

Registration, Force Protection Equipment Demonstration - May 2009 |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Registration, Force Protection Equipment Demonstration - May 2009 Registration, Force Protection Equipment Demonstration - May 2009 Registration, Force Protection Equipment Demonstration - May 2009 May 2009 Demonstrating commercially availale physical security/force protection soultions around the world The bombing of Khobar Towers in Saudi Arabia on 25 June 1996 revealed the need for continal vigilance and protection againist terrorist forces intent on harming US personnel and interests. The Chairman if the Joint Chiefs of Staff directed the Services to investigate COTS equipments solutions for physical security/force protection needs. The Office of the Under Secretary of Defense for Acquistion, Technology, and Logistics (OUSD {at&l}) tasked the Office of the US Army Product Manager, force Protection Systems (PM-FPS), to coordiante and facilitate a Force Protection Equipment

337

Hampton Roads Demonstration Project | Open Energy Information  

Open Energy Info (EERE)

Roads Demonstration Project Roads Demonstration Project Jump to: navigation, search Name Hampton Roads Demonstration Project Facility Hampton Roads Demonstration Project Sector Wind energy Facility Type Offshore Wind Facility Status Proposed Developer Virginia State Government Location Chesapeake Bay VA Coordinates 36.965°, -76.289° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":36.965,"lon":-76.289,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

338

The Lunar Laser Communications Demonstration (LLCD)  

E-Print Network (OSTI)

NASA is presently overseeing a project to create the world's first free-space laser communications system that can be operated over a range ten times larger than the near-earth ranges that have been demonstrated to date. ...

Boroson, Don M.

339

Demonstration and Deployment Strategy Workshop: Summary  

Energy.gov (U.S. Department of Energy (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.

340

Government-sponsored demonstrations of new technologies  

Science Journals Connector (OSTI)

...435959 CONTR . MANSFIELD, E, ECONOMICS TECHNOLOGI ( 1968 ). NELSON...North Carolina; a water desalination demonstration plant built...D Chicago expressway o Desalination (Pt. Loma) LO o Fish protein...1974); E. Mansfield, The Economics of Technological Change...

WS Baer; LL Johnson; EW Merrow

1977-05-27T23:59:59.000Z

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

National Hydrogen Learning Demonstration Status (Presentation)  

SciTech Connect

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

342

"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

343

Status of the MAJORANA DEMONSTRATOR experiment  

SciTech Connect

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

344

Calderon Cokemaking Process/Demonstration Project  

SciTech Connect

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

345

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

346

Technoclimat- Green Technologies Demonstration Program (Quebec, Canada)  

Energy.gov (U.S. Department of Energy (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...

347

Grays Harbor Demonstration Project | Open Energy Information  

Open Energy Info (EERE)

Demonstration Project Demonstration Project Jump to: navigation, search Name Grays Harbor Demonstration Project Facility Grays Harbor Demonstration Project Sector Wind energy Facility Type Offshore Wind Facility Status Proposed Owner Grays Harbor Ocean Energy Company LLC Developer Grays Harbor Ocean Energy Company LLC Location Pacific Ocean Coordinates 46.858°, -124.187° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":46.858,"lon":-124.187,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

348

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

349

Actinide burning in the integral fast reactor  

SciTech Connect

During the past few years, Argonne National Laboratory has been developing the integral fast reactor (IFR), an advanced liquid-metal reactor concept. In the IFR, the inherent properties of liquid-metal cooling are combined with a new metallic fuel and a radically different refining process to allow breakthroughs in passive safety, fuel cycle economics, and waste management. A key feature of the IFR concept is its unique pyroprocessing. Pyroprocessing has the potential to radically improve long-term waste management strategies by exploiting the following attributes: 1. Minor actinides accompany plutonium product stream; therefore, actinide recycling occurs naturally. Actinides, the primary source of long-term radiological toxicity, are removed from the waste stream and returned to the reactor for in situ burning, generating useful energy. 2. High-level waste volume from pyroprocessing call be reduced substantially as compared with direct disposal of spent fuel. 3. Decay heat loading in the repository can be reduced by a large factor, especially for the long-term burden. 4. Low-level waste generation is minimal. 5. Troublesome fission products, such as [sup 99]Tc, [sup 129]I, and [sup 14]C, are contained and immobilized. Singly or in combination, the foregoing attributes provide important improvements in long-term waste management in terms of the ease in meeting technical performance requirements (perhaps even the feasibility of demonstrating that technical performance requirements can be met) and perhaps also in ultimate public acceptance. Actinide recycling, if successfully developed, could well help the current repository program by providing an opportunity to enhance capacity utilization and by deferring the need for future repositories. It also represents a viable technical backup option in the event unforeseen difficulties arise in the repository licensing process.

Chang, Y.I. (Argonne National Lab., IL (United States))

1993-01-01T23:59:59.000Z

350

What is the Federal Demonstration Project  

SciTech Connect

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

351

Integrated powerhead demonstration full flow cycle development  

Science Journals Connector (OSTI)

The Integrated Powerhead Demonstration (IPD) is a 1 112 000 N (250 000? lb f ) thrust (at sea level) LOX/LH2 demonstration of a full flow cycle in an integrated system configuration. Aerojet and Rocketdyne are on contract to the Air Force Research Laboratory to design develop and deliver the required components and to provide test support to accomplish the demonstration. Rocketdyne is on contract to provide a fuel and oxygen turbopump a gas-gas injector and system engineering and integration. Aerojet is on contract to provide a fuel and oxygen preburner a main combustion chamber and a nozzle. The IPD components are being designed with Military Spaceplane (MSP) performance and operability requirements in mind. These requirements include: lifetime ?200 missions mean time between overhauls ?100 cycles and a capability to throttle from 20% to 100% of full power. These requirements bring new challenges both in designing and testing the components. This paper will provide some insight into these issues. Lessons learned from operating and supporting the space shuttle main engine (SSME) have been reviewed and incorporated where applicable. The IPD program will demonstrate phase I goals of the Integrated High Payoff Rocket Propulsion Technology (IHPRPT) program while demonstrating key propulsion technologies that will be available for MSP concepts. The demonstration will take place on Test Stand 2A at the Air Force Research Laboratory at Edwards AFB. The component tests will begin in 1999 and the integrated system tests will be completed in 2002.

J. Mathew Jones; James T. Nichols; William F. Sack; William D. Boyce; William A. Hayes

1998-01-01T23:59:59.000Z

352

Role of fast reactor and its cycle to reduce nuclear waste burden  

SciTech Connect

The role of the metal fuel fast reactor with recycling of actinides and the five long-lived fission products based on the concept of the Self-Consistent Nuclear Energy System has been examined by evaluating the reduction of nuclear wastes during the transition period to this reactor system. The evaluation was done in comparison to an LWR once-through case and a conventional actinide recycling oxide fast reactor. As a result, it is quantitatively clarified that a metal fuel fast reactor with actinide and the five long-lived fission products (I{sup 129}, Tc{sup 99}, Zr{sup 93}, Cs{sup 135} and Sn{sup 126}) recycling could play a significant role in reducing the nuclear waste burden including the current LWR wastes. This can be achieved by using a fast neutron spectrum reactor enhanced with metal fuel that brings high capability as a 'waste burner'. (authors)

Arie, Kazuo; Oomori, Takashi; Okita, Takeshi [Toshiba Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan); Kawashima, Masatoshi [Toshiba Nuclear Engineering Services Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan); Kotake, Shoji [The Japan Atomic Power Company, 1-1, Kanda-Mitoshiro-cho, Chiyoda-ku, Tokyo 101-0053 (Japan); Fuji-ie, Yoichi [Nuclear Salon Fuji-ie, 1-11-10, Yushima, Bunkyo-ku, Tokyo 113-0034 (Japan)

2013-07-01T23:59:59.000Z

353

Reliability analysis of a passive cooling system using a response surface with an application to the Flexible Conversion Ratio Reactor  

E-Print Network (OSTI)

A comprehensive risk-informed methodology for passive safety system design and performance assessment is presented and demonstrated on the Flexible Conversion Ratio Reactor (FCRR). First, the methodology provides a framework ...

Fong, Christopher J. (Christopher Joseph)

2008-01-01T23:59:59.000Z

354

Reactor Pressure Vessel Head Packaging & Disposal  

SciTech Connect

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

355

Decision support software technology demonstration plan  

SciTech Connect

The performance evaluation of innovative and alternative environmental technologies is an integral part of the US Environmental Protection Agency's (EPA) mission. Early efforts focused on evaluating technologies that supported the implementation of the Clean Air and Clean Water Acts. In 1986 the Agency began to demonstrate and evaluate the cost and performance of remediation and monitoring technologies under the Superfund Innovative Technology Evaluation (SITE) program (in response to the mandate in the Superfund Amendments and Reauthorization Act of 1986 (SARA)). In 1990, the US Technology Policy was announced. This policy placed a renewed emphasis on making the best use of technology in achieving the national goals of improved quality of life for all Americans, continued economic growth, and national security. In the spirit of the technology policy, the Agency began to direct a portion of its resources toward the promotion, recognition, acceptance, and use of US-developed innovative environmental technologies both domestically and abroad. Decision Support Software (DSS) packages integrate environmental data and simulation models into a framework for making site characterization, monitoring, and cleanup decisions. To limit the scope which will be addressed in this demonstration, three endpoints have been selected for evaluation: Visualization; Sample Optimization; and Cost/Benefit Analysis. Five topics are covered in this report: the objectives of the demonstration; the elements of the demonstration plan; an overview of the Site Characterization and Monitoring Technology Pilot; an overview of the technology verification process; and the purpose of this demonstration plan.

SULLIVAN,T.; ARMSTRONG,A.

1998-09-01T23:59:59.000Z

356

Buried Waste Integrated Demonstration Plan. Revision 1  

SciTech Connect

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

357

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

SciTech Connect

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

358

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

EMERGENCY PREPAREDNESS (EP) EMERGENCY PREPAREDNESS (EP) OBJECTIVE EP-1: A routine drill program and emergency operations drill program, including program records, have been established and implemented. (Core Requirement 11) Criteria * Reactor operation with the CS has been appropriately incorporated into the emergency preparedness hazards analysis and emergency response procedures. * The implemented routine and emergency operations drill program, including program records, have incorporated the CS SSCs and the CS's operation, hazards, and reactor interface. * Proficiency to appropriately respond to incidents and accidents associated with reactor operation has been demonstrated through the implemented routine and emergency operations drill program. Approach Record Review: Examine ORNL/RRD/INT-114, HFIR Emergency Planning Hazards

359

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

ENGINEERING (ENG) ENGINEERING (ENG) OBJECTIVE ENG-1: The engineering program has been appropriately modified to reflect the CS modification and its reactor interface, sufficient numbers of qualified engineering personnel are provided, and adequate facilities and equipment are available to ensure engineering services are adequate to support reactor and CS operations. The engineering functions, assignments, responsibilities, and reporting relationships are clearly defined, understood, and effectively implemented with line management control of safety. Engineering personnel exhibit awareness of the applicable requirements pertaining to reactor operation with the CS and with CS operations and hazards. Through their actions, they have demonstrated a high-priority commitment

360

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

Energy.gov (U.S. Department of Energy (DOE))

Honeywell’s Smart Grid Investment Grant (SGIG) project demonstrates utility-scale performance of a hardware/software platform for automated demand response (ADR) for utility, commercial, and industrial customers. The case study is now available for downloading.

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

Natural Fueling of a Tokamak Fusion Reactor  

E-Print Network (OSTI)

A natural fueling mechanism that helps to maintain the main core deuterium and tritium (DT) density profiles in a tokamak fusion reactor is discussed. In H-mode plasmas dominated by ion- temperature gradient (ITG) driven turbulence, cold DT ions near the edge will naturally pinch radially inward towards the core. This mechanism is due to the quasi-neutral heat flux dominated nature of ITG turbulence and still applies when trapped and passing kinetic electron effects are included. Fueling using shallow pellet injection or supersonic gas jets is augmented by an inward pinch of could DT fuel. The natural fueling mechanism is demonstrated using the three-dimensional toroidal electromagnetic gyrokinetic turbulence code GEM and is analyzed using quasilinear theory. Profiles similar to those used for conservative ITER transport modeling that have a completely flat density profile are examined and it is found that natural fueling actually reduces the linear growth rates and energy transport.

Wan, Weigang; Chen, Yang; Perkins, Francis W

2009-01-01T23:59:59.000Z

362

Safeguards Licensing Aspects of a Future Generation IV Demonstration Facility.  

E-Print Network (OSTI)

?? Generation IV (Gen IV) is a developing new generation of nuclear power reactors which is foreseen to bring about a safer and more sustainable… (more)

Åberg Lindell, Matilda

2010-01-01T23:59:59.000Z

363

NETL: CCPI/Clean Coal Demonstrations  

NLE Websites -- All DOE Office Websites (Extended Search)

Topical Reports Topical Reports CCPI/Clean Coal Demonstrations Topical Reports General Topical Report #18: Environmental Benefits of Clean Coal Technologies[PDF-2MB] (Apr 2001) This report describes a variety of processes that are capable of meeting existing and emerging environmental regulations and competing economically in a deregulated electric power marketplace. Topical Report #17: Software Systems in Clean Coal Demonstration Projects [PDF-650KB] (Dec 2001) This report describes computer software systems used to optimize coal utilization technologies. Environmental Control Technologies Sulfur Dioxide Control Technologies Topical Report #12: Advanced Technologies for the Control of Sulfur Dioxide Emissions from Coal-Fired Boilers [PDF-1.6MB] (June 1999) A discussion of three CCT projects that demonstrate innovative wet flue gas desulfurization technologies to remove greater than 90% SO2.

364

Students, Professors Demonstrating Virginia's Potential | Department of  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Students, Professors Demonstrating Virginia's Potential Students, Professors Demonstrating Virginia's Potential Students, Professors Demonstrating Virginia's Potential April 23, 2010 - 10:37am Addthis Joshua DeLung What does this project do? James Madison University is spearheading an effort to make at least 25 percent of the energy used in the Shenandoah Valley come from renewable sources as soon as possible. Peter Denbigh's real interest is in finding ways to help Americans save energy - especially in terms of the transportation industry - so it's appropriate that he says "the key is to address the situation effectively so that we don't just spin our wheels and get nowhere." Peter enjoys solving complex problems, including one thing he sees as an urgent situation that is vastly important - energy creation. He earned a

365

TRW advanced slagging coal combustor utility demonstration  

SciTech Connect

The TRW Advanced Entrained Coal Combustor Demonstration Project consists of retrofitting Orange and Rockland (O R) Utility Corporation's Lovett Plant Unit No. 3 with four (4) slagging combustors which will allow the gas/oil unit to fire 2.5% sulfur coal. The slagging combustor process will provide NO{sub x} and SO{sub x} emissions that meet NSPS and New York State Environmental Standards. The TRW-Utility Demonstration Unit (UDU) is responsible for the implementation of program policies and overall direction of the project. The following projects will be carried out: process and design development of clean coal technology CCT-1 the development and operation of the entrained coal combustor will enable the boiler to burn low and medium sulfur coal while meeting all the Federal/State emission requirements; demonstrate sulfur dioxide emissions control by pulverized limestone injection into the entrained coal combustor system.

Not Available

1990-01-01T23:59:59.000Z

366

Heat Pump Water Heaters Demonstration Project  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Heat Pump Water Heaters Heat Pump Water Heaters Demonstration Project Building America Stakeholder Meeting Ron Domitrovic Ammi Amarnath 3/1/2012 Austin, TX 2 © 2011 Electric Power Research Institute, Inc. All rights reserved. HPWH Field Demonstration: Research Objectives * Assess heat pump water heater technology by measuring efficiency. * Provide credible data on the performance and reliability of heat pump water heaters. * Assess user satisfaction in a residential setting. 3 © 2011 Electric Power Research Institute, Inc. All rights reserved. Demonstration Host Utilities Target: 40 Units per Utility Installed and Potential Sites by Climate Zone Source: Department of Energy (DOE), Building America climate regions 4 © 2011 Electric Power Research Institute, Inc. All rights reserved. Installation Locations-Southern Company Region

367

Airborne Process Commercial Scale Demonstration Program  

NLE Websites -- All DOE Office Websites (Extended Search)

CCPI 2) CCPI 2) contacts Brad tomer Director Office of Major Demonstrations National Energy Technology Laboratory 3610 Collins Ferry Road P.O. Box 880 Morgantown, WV 26507-0880 304-285-4692 brad.tomer@netl.doe.gov PaRtIcIPant Mustang Clean Energy, LLC, a subsidiary of Peabody Energy St. Louis, MO. Airborne Process(tm) commerciAl scAle DemonstrAtion ProgrAm (withDrAwn Prior to AwArD) Project Description Mustang Clean Energy will design, construct, and operate a full scale sodium-based multi-pollutant scrubber in conjunction with a revenue-generating fertilizer by-product processing plant at Mustang Energy Company, LLC's Mustang Generating Station. Both Mustang Clean Energy and Mustang Energy Company are subsidiaries of Peabody Energy, the world's largest coal company. The 300 MW (net) station will

368

Advanced Nuclear Reactors | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

369

Advanced Reactor Technology Documents | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

370

Microsoft Word - power_reactors_briggs.doc  

NLE Websites -- All DOE Office Websites (Extended Search)

Most common - Boiling Water and Pressurized Most common - Boiling Water and Pressurized Water Reactors About 80% of the world's nuclear reactors used for generating electricity are either boiling water reactors or pressurized water reactors. Of these, about 30% are boiling water reactors and 70% are pressurized water reactors. All power reactors currently in use in the United States are of these two types. Both types of reactors have been very successfully used for reliable, on-demand, emissions-free electricity generation for decades. How does a boiling water reactor work? Water flows from the bottom of the fuel to the top of the fuel, and as it moves past the fuel, it carries away the heat produced by the

371

A Demonstration System for Capturing Geothermal Energy from Mine Waters  

Open Energy Info (EERE)

System for Capturing Geothermal Energy from Mine Waters System for Capturing Geothermal Energy from Mine Waters beneath Butte, MT Geothermal Project Jump to: navigation, search Last modified on July 22, 2011. Project Title A Demonstration System for Capturing Geothermal Energy from Mine Waters beneath Butte, MT Project Type / Topic 1 Recovery Act - Geothermal Technologies Program: Ground Source Heat Pumps Project Type / Topic 2 Topic Area 1: Technology Demonstration Projects Project Description Butte, Montana, like many other mining towns that developed because of either hard-rock minerals or coal, is underlain by now-inactive water-filled mines. In Butte's case, over 10,000 miles of underground workings have been documented, but as in many other mining communities these waters are regarded as more of a liability than asset. Mine waters offer several advantages:

372

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

Energy Savers (EERE)

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

373

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

374

Novel Composite Materials Demonstrate Ultra-sensitivity  

NLE Websites -- All DOE Office Websites (Extended Search)

2, Issue 25 2, Issue 25 Novel Composite Materials Demonstrate Ultra-sensitivity Gold nanowires on graphite templates used in gas sensing application page 2 Coronary Stent Wins Technology Transfer Award page 4 University of Oregon Team Wins Competition for Commercializing NETL Technology page 5 the ENERGY lab NATIONAL ENERGY TECHNOLOGY LABORATORY 2 Novel Composite Materials Demonstrate Ultra-sensitivity-Gold nanowires on graphite templates used in gas sensing applications ____________________2 Coronary Stent Wins Technology Transfer Award ________4 University of Oregon Team Wins Competition for Commercializing NETL Technology __________________5 NETL & WVU Researchers Design New Catalysts for CO 2 Management ___________________________________6 Structurally Dynamic MOF Sorbent Selectively Adsorbs

375

Power Plant Optimization Demonstration Projects Cover Photos:  

NLE Websites -- All DOE Office Websites (Extended Search)

5 SEPTEMBER 2007 5 SEPTEMBER 2007 Power Plant Optimization Demonstration Projects Cover Photos: * Top left: Coal Creek Station * Top right: Big Bend Power Station * Bottom left: Baldwin Energy Complex * Bottom right: Limestone Power Plant A report on four projects conducted under separate cooperative agreements between the U.S. Department of Energy and: * Great River Energy * Tampa Electric Company * Pegasus Technologies * NeuCo. , Inc.  Power Plant Optimization Demonstration Projects Executive Summary .......................................................................................4 Background: Power Plant Optimization ......................................................5 Lignite Fuel Enhancement Project ...............................................................8

376

JEA successfully completes world's largest CFB demonstration  

SciTech Connect

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

377

UCLA program in reactor studies: The ARIES tokamak reactor study  

SciTech Connect

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

378

Eddy current NDE performance demonstrations using simulation tools  

SciTech Connect

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

379

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

380

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

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

Spectral Effects on Stress Relaxation of Inconel X-750 Springs in CANDU Reactors  

SciTech Connect

CANDU reactors have been operating for periods up to about 25 years. During this time there are changes to the nuclear reactor core components that are a function of operating environment and time. It is important to know how the properties of critical core components are likely to change over the life of a reactor and therefore their behaviours are characterised long before the end of the reactor design life. Tests are typically conducted in materials test reactors. The behaviour of a material is often characterised as a function of fast neutron fluence and the expected effect of operating time is established by simply extrapolating as a function of fluence. This may be appropriate when the neutron energy spectrum for the materials test reactor matches closely the neutron spectrum where the component resides in the power reactor. However, in cases where the spectrum is very different one has to convert the accumulated dose into a unit that is common in its effect on the material properties. For many property changes in nuclear reactor cores this unit is displacements per atom (dpa).

Griffiths, M.; Butcher, F. J.; Ariani, I.; Douglas, S.; Garner, Francis A.; Greenwood, Lawrence R.

2008-11-16T23:59:59.000Z

382

Demand Response Projects: Technical and Market Demonstrations  

E-Print Network (OSTI)

Demand Response Projects: Technical and Market Demonstrations Philip D. Lusk Deputy Director Energy Analyst #12;PLACE CAPTION HERE. #12;#12;#12;#12;City of Port Angeles Demand Response History energy charges · Demand charges during peak period only ­ Reduced demand charges for demand response

383

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

384

Webinar: EISPC Energy Zones Mapping Tool Demonstration  

Energy.gov (U.S. Department of Energy (DOE))

Argonne National Laboratory has organized a webinar demonstration of the Eastern Interconnection States’ Planning Council (EISPC) Energy Zones (EZ) Mapping Tool, a free online mapping tool for identifying areas in all 39 EISPC states that may be suitable for new clean power generation. The tool is maintained by Argonne National Laboratory with funding from the U.S. Department of Energy.

385

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

386

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

Science Journals Connector (OSTI)

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

Fernando L.P. Resende

2014-01-01T23:59:59.000Z

387

PINON PINE: An advanced IGCC demonstration  

SciTech Connect

The Pinon Pine Power Project is a second generation integrated gasification combined cycle (IGCC) power plant, located at Sierra Pacific Power Company`s (SPPC) Tracy Station, 17 miles east of Reno, Nevada. The project is being partially funded under the Department of Energy`s (DOE`s) Clean Coal Technology Program (CCT). SPPC intends to operate the plant in base-load mode to supply approximately 100 megawatts electric (MWe) to the transmission grid. This plant will be the first full-scale integration of several advanced technologies: an air-blown KRW gasifier; full-stream hot gas desulfurization using a transport reactor system with a zinc-based sorbent; full-stream, high-temperature ceramic filters for particulate removal; the General Electric Model MS6001FA (617A) Gas Turbine Engine/generator, and a 950 pound per square inch absolute (psia), 950{degrees}F steam turbine generator. This paper reviews the overall configuration and integration of the gasification and power islands components, which yield the plant`s high efficiency. Current status of the project is addressed.

Freier, M.D.; Jewell, D.M. [Morgantown Energy Technology Center, WV (United States); Motter, J.W. [Sierra Pacific Power Co., Reno, NV (United States)

1996-04-01T23:59:59.000Z

388

N Reactor Placed In Interim Safe Storage: Largest Hanford Reactor Cocooning  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

N Reactor Placed In Interim Safe Storage: Largest Hanford Reactor N Reactor Placed In Interim Safe Storage: Largest Hanford Reactor Cocooning Project Now Complete N Reactor Placed In Interim Safe Storage: Largest Hanford Reactor Cocooning Project Now Complete June 14, 2012 - 12:00pm Addthis Media Contacts Cameron Hardy Cameron.Hardy@rl.doe.gov 509-376-5365 Mark McKenna mmckenna@wch-rcc.com 509-372-9032 RICHLAND, WASH. - The U.S. Department of Energy's (DOE's) River Corridor contractor, Washington Closure Hanford, has completed placing N Reactor in interim safe storage, a process also known as "cocooning." N Reactor was the last of nine plutonium production reactors to be shut down at DOE's Hanford Site in southeastern Washington state. It was Hanford's longest-running reactor, operating from 1963 to 1987. "In the 1960's, N Reactor represented the future of energy in America.

389

Graphite Reactor | ornl.gov  

NLE Websites -- All DOE Office Websites (Extended Search)

Graphite Reactor Graphite Reactor 'In the early, desperate days of World War II, the United States launched the top-secret, top-priority Manhattan Project...' In the early, desperate days of U.S. involvement in World War II, American scientists began to fear that the German discovery of uranium fission in 1939 might enable the Nazis to develop a super bomb. Afraid of losing this crucial race, the United States launched the top-secret, top-priority Manhattan Project. The plan was to create two atomic weapons-one fueled by plutonium, the other by enriched uranium. Hanford, Washington, was selected as the site for plutonium production, but before large reactors could be built there, a pilot plant was necessary to prove the feasibility of scaling up from laboratory experiments. A secluded, rural area near Clinton, Tennessee, was

390

Actinide Burning in CANDU Reactors  

SciTech Connect

Actinide burning in CANDU reactors has been studied as a method of reducing the actinide content of spent nuclear fuel from light water reactors, and thereby decreasing the associated long term decay heat load. In this work simulations were performed of actinides mixed with natural uranium to form a mixed oxide (MOX) fuel, and also mixed with silicon carbide to form an inert matrix (IMF) fuel. Both of these fuels were taken to a higher burnup than has previously been studied. The total transuranic element destruction calculated was 40% for the MOX fuel and 71% for the IMF. (authors)

Hyland, B.; Dyck, G.R. [Atomic Energy of Canada Limited, Chalk River, Ontario, K0J 1J0 (Canada)

2007-07-01T23:59:59.000Z

391

Commercial-Scale Demonstration of the Liquid Phase Methanol (LPMEOH) Process  

SciTech Connect

he Liquid Phase Methanol (LPMEOW) Demonstration Project at Kingsport Tennessee, is a $213.7 million cooperative agreement between the U.S. Department of Energy (DOE) and Air Products Liquid Phase Conversion Company, L.P. (the Partnership) to produce methanol from coal-derived synthesis gas (syngas). Air Products and Chemicals, Inc. (Air Products) and Eastman Chemical Company (Eastman) formed the Partnership to execute the Demonstration Project. The LPMEOEP Process Demonstration Unit was built at a site located at the Eastman coal-to-chemicals complex in Kingsport. The LPMEOHW Demonstration Facility completed its first year of operation on 02 April 1998. The LPMEOW Demonstration Facility also completed the longest continuous operating run (65 days) on 21 April 1998. Catalyst activity, as defined by the ratio of the rate constant at any point in time to the rate constant for freshly reduced catalyst (as determined in the laboratory autoclave), was monitored throughout the reporting period. During a six-week test at a reactor temperature of 225oC and Balanced Gas flowrate of 700 KSCFH, the rate of decline in catalyst activity was steady at 0.29-0.36% per day. During a second one-month test at a reactor temperature of 220oC and a Balanced Gas flowrate of 550-600 KSCFH, the rate of decline in catalyst activity was 0.4% per day, which matched the pefiorrnance at 225"C, as well as the 4-month proof-of-concept run at the LaPorte AFDU in 1988/89. Beginning on 08 May 1998, the LPMEOW Reactor temperature was increased to 235oC, which was the operating temperature tier the December 1997 restart with the fresh charge of catalyst (50'Yo of design loading). The flowrate of the primary syngas feed stream (Balanced Gas) was also increased to 700-750 KSCFH. During two stable operating periods between 08 May and 09 June 1998, the average catalyst deactivation rate was 0.8% per day. Due to the scatter of the statistical analysis of the results, this test was extended to better quanti& the catalyst aging behavior. During the reporting perio~ two batches of fresh catalyst were activated and transferred to the reactor (on 02 April and 20 June 1998). The weight of catalyst in the LPMEOW Reactor has reached 80% of the design value. At the end of the reporting period, a step-change in the pressure-drop profile within the LPMEOW Reactor and an increase in the pressure of the steam system which provides cooling to the LPMEOW Reactor were observed. No change in the calculated activity of the catalyst was detected during either of these transients. These parameters will be monitored closely for any additional changes.

None

1998-12-21T23:59:59.000Z

392

Western Greenbrier Co-Production Demonstration Project  

NLE Websites -- All DOE Office Websites (Extended Search)

contacts contacts Brad tomer Director Office of Major Demonstrations National Energy Technology Laboratory 3610 Collins Ferry Road P.O. Box 880 Morgantown, WV 26507-0880 304-285-4692 brad.tomer@netl.doe.gov nelson Rekos Project Manager National Energy Technology Laboratory 3610 Collins Ferry Road P.O. Box 880 Morgantown, WV 26507-0880 304-285-4066 nelson.rekos@netl.doe.gov PaRtIcIPant Western Greenbrier Co-Generation, LLC Lewisburg, WV Western Greenbrier Co-ProduCtion demonstration ProjeCt (disContinued) Project Description The Western Greenbrier Co-Production (WGC) project will generate about 100 megawatts of electricity and commercial quantities of salable ash by-products by burning waste coal presently contained in numerous coal refuse dumps in the vicinity of the plant. These refuse dumps, created by coal cleaning operations over

393

Environmental management technology demonstration and commercialization  

SciTech Connect

The Energy & Environmental Research Center (EERC), a contract-supported organization focused on technology research, development, demonstration, and commercialization (RDD&C), is entering its second year of a Cooperative Agreement with the U.S. Department of Energy (DOE) Morgantown Energy Technology Center (METC) to facilitate the development, demonstration, and commercialization of innovative environmental management (EM) technologies in support of the activities of DOE`s Office of Environmental Science and Technology (EM-50) under DOE`s EM Program. This paper reviews the concept and approach of the program under the METC-EERC EM Cooperative Agreement and profiles the role the program is playing in the commercialization of five EM technologies.

Daly, D.J.; Erickson, T.A.; Groenewold, G.H. [and others

1995-12-31T23:59:59.000Z

394

Implementation of the buried waste integrated demonstration  

SciTech Connect

The Department of Energy (DOE), Office of Technology Development (OTD) has initiated the Buried Waste Integrated Demonstration (BWID) to resolve technological deficiencies associated with the remediation of radioactive and hazardous buried waste. The BWID mission is to identify, demonstrate, and transfer innovative technologies for the remediation of DOE buried waste. To accomplish the mission, BWID is using a systems approach which supports the development of a suite of advanced and innovative technologies for the effective and efficient remediation of buried waste. This systems approach includes technologies for theentire remediation cycle. Specifically, BWID sponsors technology development in the following technology categories: site and waste characterization, retrieval, preprocessing, ex situ treatment, packaging, transportation, storage, disposal, and post-disposal monitoring.

Kostelnik, K.M.; Merrill, S.K.

1992-09-01T23:59:59.000Z

395

Implementation of the buried waste integrated demonstration  

SciTech Connect

The Department of Energy (DOE), Office of Technology Development (OTD) has initiated the Buried Waste Integrated Demonstration (BWID) to resolve technological deficiencies associated with the remediation of radioactive and hazardous buried waste. The BWID mission is to identify, demonstrate, and transfer innovative technologies for the remediation of DOE buried waste. To accomplish the mission, BWID is using a systems approach which supports the development of a suite of advanced and innovative technologies for the effective and efficient remediation of buried waste. This systems approach includes technologies for theentire remediation cycle. Specifically, BWID sponsors technology development in the following technology categories: site and waste characterization, retrieval, preprocessing, ex situ treatment, packaging, transportation, storage, disposal, and post-disposal monitoring.

Kostelnik, K.M.; Merrill, S.K.

1992-01-01T23:59:59.000Z

396

A transmission electron microscope for lecture demonstrations  

Science Journals Connector (OSTI)

A simple transmission electron microscope (TEM) suitable for lecture demonstrations is described. In this TEM electrons are created in a glow discharge between two parallel electrodes in air at a reduced pressure. The electrons are collimated by a small hole in the anode focused by a solenoid that acts as an electromagnetic lens and imaged on a thin layer of phosphor deposited inside an Erlenmeyer flask. An image of a biological sample placed between the source and the lens can be magnified about 20 times. The microscope uses inexpensive components that can be quickly assembled during a demonstration. The TEM provides a visual and memorable display that highlights phenomena such as mean-free-path charged particle optics electrical discharges and cathodoluminescence.

J. A. Panitz; Gertrude Rempfer

2006-01-01T23:59:59.000Z

397

Pit Disassembly and Conversion Demonstration Environmental Ass  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

August 1998 August 1998 i TABLE OF CONTENTS 1.0 INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.1 Related National Environmental Policy Act Reviews . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.0 PURPOSE AND NEED FOR ACTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.1 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.2 Purpose and Need for Action . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.0 PROPOSED PIT DISASSEMBLY AND CONVERSION DEMONSTRATION . . . . . . . . . . . . . . . . 6 4.0 NO ACTION ALTERNATIVE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 5.0 AFFECTED ENVIRONMENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 5.1 History and Current Mission of Los Alamos National Laboratory

398

Los Alamos Team Demonstrates Bottle Scanner Technology  

SciTech Connect

Los Alamos scientists are demonstrating a Nuclear Magnetic Resonance Imaging (NMR) technology that may provide a breakthrough for screening liquids at airport security. By adding low-power X-ray data to the NMR mix, scientists believe they have unlocked a new detection technology. Funded in part by the Department of Homeland Security's Science and Technology Directorate, the new technology is called MagRay.

Espy, Michelle; Schultz, Larry

2014-05-06T23:59:59.000Z

399

Scintigraphic demonstration of tracheo-esophageal fistula  

SciTech Connect

A tracheo-esophageal fistula, developed following radiotherapy for an esophageal carcinoma, was vividly demonstrated by radionuclide imaging. The abnormality was later confirmed by a barium esophagram and endoscopic examinations. The scintigraphic procedure, making use of a Tc-99m sulfur colloid swallow, appears to be a simple alternative method use of a Tc-99m sulfur colloid swallow, appears to be a simple alternative method that may be clinically useful for the diagnosis of such a condition.

Dunn, E.K.; Man, A.C.; Lin, K.J.; Kaufman, H.D.; Solomon, N.A.

1983-12-01T23:59:59.000Z

400

Lithium-Polysulfide Flow Battery Demonstration  

SciTech Connect

In this video, Stanford graduate student Wesley Zheng demonstrates the new low-cost, long-lived flow battery he helped create. The researchers created this miniature system using simple glassware. Adding a lithium polysulfide solution to the flask immediately produces electricity that lights an LED. A utility version of the new battery would be scaled up to store many megawatt-hours of energy.

Zheng, Wesley

2014-06-30T23:59:59.000Z

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.


401

Los Alamos Team Demonstrates Bottle Scanner Technology  

ScienceCinema (OSTI)

Los Alamos scientists are demonstrating a Nuclear Magnetic Resonance Imaging (NMR) technology that may provide a breakthrough for screening liquids at airport security. By adding low-power X-ray data to the NMR mix, scientists believe they have unlocked a new detection technology. Funded in part by the Department of Homeland Security's Science and Technology Directorate, the new technology is called MagRay.

Espy, Michelle; Schultz, Larry

2014-06-02T23:59:59.000Z

402

BNL | Our History: Reactors as Research Tools  

NLE Websites -- All DOE Office Websites (Extended Search)

> See also: Accelerators > See also: Accelerators Brookhaven History: Using Reactors as Research Tools BGRR Brookhaven Graphite Research Reactor The Brookhaven Graphite Research Reactor (BGRR) was the Laboratory's first big machine and the first peace-time reactor built in the United States following World War II. The reactor's primary mission was to produce neutrons for scientific experimentation and to refine reactor technology. At the time, the BGRR could accommodate more simultaneous experiments than any other reactor. Scientists and engineers from every corner of the U.S. came to use the reactor, which was not only a source of neutrons for experiments, but also an excellent training facility. Researchers used the BGRR's neutrons as tools for studying atomic nuclei and the structure of solids, and to investigate many physical, chemical and

403

New fast-reactor approach. [LMFBR  

SciTech Connect

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

404

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

405

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

406

Inherent safety concepts in nuclear power reactors  

Science Journals Connector (OSTI)

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

O M Pal Singh; R Shankar Singh

1989-06-01T23:59:59.000Z

407

Choice of coils for a fusion reactor  

Science Journals Connector (OSTI)

...configurations. The most ambitious is the International Thermonuclear Experimental Reactor, a large tokamak planned for construction...configuration has features in common with the International Thermonuclear Experimental Reactor experiment. Mathematical Model We...

Romeo Alexander; Paul R. Garabedian

2007-01-01T23:59:59.000Z

408

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

409

The development of structural materials for fusion reactors  

Science Journals Connector (OSTI)

...severely exposed parts of future fusion reactors and pose key problems...successful implementation of fusion reactors as an efficient source...conditions in the International Thermonuclear Experimental Reactor (ITER...environmental attractiveness of fusion reactors. In this paper...

1999-01-01T23:59:59.000Z

410

Utilization of Refractory Metals and Alloys in Fusion Reactor Structures  

Science Journals Connector (OSTI)

In design of fusion reactors, structural material selection is very crucial to improve reactor’s performance. Different types of materials have been proposed for use in fusion reactor structures. Among these mate...

Mustafa Übeyli; ?enay Yalç?n

2006-12-01T23:59:59.000Z

411

Buffalo river dredging demonstration. Final report  

SciTech Connect

The Corps of Engineers Buffalo District conducted a demonstration of equipment for dredging contaminated sediments. Several thousand cubic yards of sediment were removed from outside the Buffalo River Federal navigation channel limits using three dredge types: (1) open bucket, (2) enclosed bucket, and (3) submersible pump. The effectiveness of a silt screen deployed downstream of the dredge to reduce suspended sediment transport was also evaluated. Extensive sediment and water column monitoring and sampling were conducted during the 2-week demonstration as part of the effort to determine sediment resuspension rates and contaminant releases associated with the dredging operations. Water column samples were analyzed for total suspended solids, total organic carbon, PCBs, PAHs, metals, ammonia, and pH. A water column bioassay test using Daphnia magna was also performed to assess toxicity effects of the dredging operation. Results of this study were used to assess and refine techniques and laboratory tests that have been previously developed by the Corps of Engineers to predict sediment resuspension rates and contaminant releases. In another phase of the study, the Bureau of Mines demonstrated the use of polyelectrolytes for rapid removal of suspended solids from a dilute dredged material slurry.

Averett, D.E.; Zappi, P.A.; Tatem, H.E.; Gibson, A.C.; Tominey, E.A.

1996-02-01T23:59:59.000Z

412

CHP Fuel Cell Durability Demonstration - Final Report  

SciTech Connect

Plug Power has managed a demonstration project that has tested multiple units of its high-temperature, PEM fuel cell system in micro-combined heat and power (?-CHP) applications in California. The specific objective of the demonstration project was to substantiate the durability of GenSys Blue, and, thereby, verify its technology and commercial readiness for the marketplace. In the demonstration project, Plug Power, in partnership with the National Fuel Cell Research Center (NFCRC) at the University of California, Irvine (UCI), and Sempra, will execute two major tasks: • Task 1: Internal durability/reliability fleet testing. Six GenSys Blue units will be built and will undergo an internal test regimen to estimate failure rates. This task was modified to include 3 GenSys Blue units installed in a lab at UCI. • Task 2: External customer testing. Combined heat and power units will be installed and tested in real-world residential and/or light commercial end user locations in California.

Petrecky, James; Ashley, Christopher J

2014-07-21T23:59:59.000Z

413

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

SciTech Connect

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

414

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

415

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

416

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

417

Light Water Reactor Sustainability (LWRS) Program  

NLE Websites -- All DOE Office Websites (Extended Search)

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

418

High-Fidelity Light Water Reactor Analysis with the Numerical Nuclear Reactor  

Science Journals Connector (OSTI)

Technical Paper / Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications

David P. Weber; Tanju Sofu; Won Sik Yang; Thomas J. Downar; Justin W. Thomas; Zhaopeng Zhong; Jin Young Cho; Kang Seog Kim; Tae Hyun Chun; Han Gyu Joo; Chang Hyo Kim

419

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

420

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

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

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

422

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

423

DOSE RATES FROM NEUTRON ACTIVATION OF FUSION REACTOR COMPONENTS  

E-Print Network (OSTI)

NEUTRON ACTIVATION OF FUSION REACTOR C01WONENTS LawrenceNeutron Activation of Fusion Reactor Components Lawrence

Ruby, Lawrence

2014-01-01T23:59:59.000Z

424

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

SciTech Connect

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

425

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

SciTech Connect

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

426

Nuclear Reactor Safety Design Criteria  

Directives, Delegations, and Requirements

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

427

Computer aided nuclear reactor modeling  

E-Print Network (OSTI)

Nuclear reactor modeling is an important activity that lets us analyze existing as well as proposed systems for safety, correct operation, etc. The quality of a analysis is directly proportional to the quality of the model used. In this work we look...

Warraich, Khalid Sarwar

2012-06-07T23:59:59.000Z

428

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

429

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

430

Laminar Entrained Flow Reactor (Fact Sheet)  

SciTech Connect

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

431

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

432

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

SciTech Connect

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

433

The Integral Fast Reactor (IFR) - Reactors designed/built by Argonne  

NLE Websites -- All DOE Office Websites (Extended Search)

Integral Fast Reactor Integral Fast Reactor About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

434

Integrated gasification fuel cell (IGFC) demonstration test  

SciTech Connect

As concern about the environment generates interest in ultra-clean energy plants, fuel cell power plants can respond to the challenge. Fuel cells convert hydrocarbon fuels to electricity at efficiencies exceeding conventional heat engine technologies while generating extremely low emissions. Emissions of SOx and NOx are expected to be well below current and anticipated future standards. Nitrogen oxides, a product of combustion, will be extremely low in this power plant because power is produced electrochemically rather than by combustion. Due to its higher efficiencies, a fuel cell power plant also produces less carbon dioxide. Fuel cells in combination with coal gasification, are an efficient and environmentally acceptable means to utilize the abundant coal reserves both in the US and around the world. To demonstrate this technology, FuelCell Energy, Inc. (FCE), is planning to build and test a 2-MW Fuel Cell Power Plant for operation on coal derived gas. This power plant is based on Direct Fuel Cell (DFC{trademark}) technology and will be part of a Clean Coal V IGCC project supported by the US DOE. A British Gas Lurgi (BGL) slagging fixed-bed gasification system with cold gas clean up is planned as part of a 400 MW IGCC power plant to provide a fuel gas slip stream to the fuel cell. The IGFC power plant will be built by Kentucky Pioneer Energy, A subsidiary of Global Energy, in Clark County, KY. This demonstration will result in the world's largest fuel cell power plant operating on coal derived gas. The objective of this test is to demonstrate fuel cell operation on coal derived gas at a commercial scale and to verify the efficiency and environmental benefits.

Steinfeld, G.; Ghezel-Ayagh, H.; Sanderson, R.; Abens, S.

2000-07-01T23:59:59.000Z

435

Heterogeneous Recycling in Fast Reactors  

SciTech Connect

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

436

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

SciTech Connect

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

Shamsuddin Illias

2002-06-10T23:59:59.000Z

437

Design and operation of a rotating drum radio frequency plasma reactor for the modification of free nanoparticles  

SciTech Connect

A rotating drum rf plasma reactor was designed to functionalize the surface of nanoparticles and other unusually shaped substrates through plasma polymerization and surface modification. This proof-of-concept reactor design utilizes plasma polymerized allyl alcohol to add OH functionality to Fe{sub 2}O{sub 3} nanoparticles. The reactor design is adaptable to current plasma hardware, eliminating the need for an independent reactor setup. Plasma polymerization performed on Si wafers, Fe{sub 2}O{sub 3} nanoparticles supported on Si wafers, and freely rotating Fe{sub 2}O{sub 3} nanoparticles demonstrated the utility of the reactor for a multitude of processes. X-ray photoelectron spectroscopy and Fourier transform infrared spectroscopy were used to characterize the surface of the substrates prior to and after plasma deposition, and scanning electron microscopy was used to verify that no extensive change in the size or shape of the nanoparticles occurred because of the rotating motion of the reactor. The reactor design was also extended to a non-depositing NH{sub 3} plasma modification system to demonstrate the reactor design is effective for multiple plasma processes.

Shearer, Jeffrey C.; Fisher, Ellen R. [Department of Chemistry, Colorado State University, Fort Collins, Colorado 80523-1872 (United States)

2013-06-15T23:59:59.000Z

438

Better Buildings Residential Program Solution Center Demonstration  

NLE Websites -- All DOE Office Websites (Extended Search)

Danielle Sass Byrnett Better Buildings Residential Building Technologies Office Program Solution Center Demonstration Outline * Goals, History, Content Sources * Tour: Organization - Program Components - Handbooks * Tour: Navigation Options * Tour: Examples * Next Steps * Questions & Feedback 2 eere.energy.gov Overview 3 eere.energy.gov Purpose: Support Residential Energy Efficiency Upgrade Programs & Partners * Provide an easily accessed repository for key lessons, resources, and knowledge collected from the experience of past programs. * Help programs and their partners plan, implement, manage, and evaluate better * Help stakeholders leapfrog past missteps en route to a larger and more successful industry. 4 eere.energy.gov Intended Audiences

439

Newberry EGS Demonstration Project Environmental Analysis (EA)  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Newberry Volcano Enhanced Geothermal System (EGS) Demonstration Project UNITED STATES DEPARTMENT OF THE INTERIOR BUREAU OF LAND MANAGEMENT (BLM) DOI-BLM-OR-P000-2011-0003-EA DOE/EA-1897 ENVIRONMENTAL ASSESSMENT DECEMBER 2011 Location: Federal Geothermal Leases on the West Flank of Newberry Volcano, Deschutes County, 22 miles south of Bend, Oregon Applicant: Davenport Newberry Holdings LLC and AltaRock Energy, Inc. 225 NW Franklin Avenue, Suite 1 Bend, OR 97701 Tel: 541-323-1190 Lead Agency: U.S. Department of the Interior,

440

Trial Demonstration of Area Lighting Retrofit  

Energy.gov (U.S. Department of Energy (DOE))

LED Area Lighting Retrofit: Yuma Border Patrol Along the Yuma Sector Border Patrol Area in Yuma, Arizona, the GATEWAY program conducted a trial demonstration in which the incumbent quartz metal halide area lighting was replaced with LED at three pole locations. The LED system was found to equal or better the incumbent system in terms of both illuminance and uniformity, and an advanced optical system and lower pole height improved the illuminance uniformity, reduced stray light, and increased projected energy and maintenance cost savings. This high luminous flux and high temperature application is not unique and similar applications can benefit from the findings of this installation.

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441

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

442

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

SciTech Connect

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

443

THOR Bench-Scale Steam Reforming Demonstration  

SciTech Connect

The Idaho Nuclear Technology and Engineering Center (INTEC) was home to nuclear fuel reprocessing activities for decades at the Idaho National Engineering and Environmental Laboratory. As a result of the reprocessing activities, INTEC has accumulated approximately one million gallons of acidic, radioactive, sodium-bearing waste (SBW). The purpose of this demonstration was to investigate a reforming technology, offered by THORsm Treatment Technologies, LLC, for treatment of SBW into a ''road ready'' waste form that would meet the waste acceptance criteria for the Waste Isolation Pilot Plant (WIPP). A non-radioactive simulated SBW was used based on the known composition of waste tank WM-180 at INTEC. Rhenium was included as a non-radioactive surrogate for technetium. Data was collected to determine the nature and characteristics of the product, the operability of the technology, the composition of the off-gases, and the fate of key radionuclides (cesium and technetium) and volatile mercury compounds. The product contained a low fraction of elemental carbon residues in the cyclone and filter vessel catches. Mercury was quantitatively stripped from the product but cesium, rhenium (Tc surrogate), and the heavy metals were retained. Nitrates were not detected in the product and NOx destruction exceeded 98%. The demonstration was successful.

Marshall, D.W.; Soelberg, N.R.; Shaber, K.M.

2003-05-21T23:59:59.000Z

444

THOR Bench-Scale Steam Reforming Demonstration  

SciTech Connect

The Idaho Nuclear Technology and Engineering Center (INTEC) was home to nuclear fuel reprocessing activities for decades at the Idaho National Engineering and Environmental Laboratory. As a result of the reprocessing activities, INTEC has accumulated approximately one million gallons of acidic, radioactive, sodium-bearing waste (SBW). The purpose of this demonstration was to investigate a reforming technology, offered by THORsm Treatment Technologies, LLC, for treatment of SBW into a "road ready" waste form that would meet the waste acceptance criteria for the Waste Isolation Pilot Plant (WIPP). A non-radioactive simulated SBW was used based on the known composition of waste tank WM-180 at INTEC. Rhenium was included as a non-radioactive surrogate for technetium. Data was collected to determine the nature and characteristics of the product, the operability of the technology, the composition of the off-gases, and the fate of key radionuclides (cesium and technetium) and volatile mercury compounds. The product contained a low fraction of elemental carbon residues in the cyclone and filter vessel catches. Mercury was quantitatively stripped from the product but cesium, rhenium (Tc surrogate), and the heavy metals were retained. Nitrates were not detected in the product and NOx destruction exceeded 98%. The demonstration was successful.

D. W. Marshall; N. R. Soelberg; K. M. Shaber

2003-05-01T23:59:59.000Z

445

Robotics for mixed waste operations, demonstration description  

SciTech Connect

The Department of Energy (DOE) Office of Technology Development (OTD) is developing technology to aid in the cleanup of DOE sites. Included in the OTD program are the Robotics Technology Development Program and the Mixed Waste Integrated Program. These two programs are working together to provide technology for the cleanup of mixed waste, which is waste that has both radioactive and hazardous constituents. There are over 240,000 cubic meters of mixed low level waste accumulated at DOE sites and the cleanup is expected to generate about 900,000 cubic meters of mixed low level waste over the next five years. This waste must be monitored during storage and then treated and disposed of in a cost effective manner acceptable to regulators and the states involved. The Robotics Technology Development Program is developing robotics technology to make these tasks safer, better, faster and cheaper through the Mixed Waste Operations team. This technology will also apply to treatment of transuranic waste. The demonstration at the Savannah River Site on November 2-4, 1993, showed the progress of this technology by DOE, universities and industry over the previous year. Robotics technology for the handling, characterization and treatment of mixed waste as well robotics technology for monitoring of stored waste was demonstrated. It was shown that robotics technology can make future waste storage and waste treatment facilities better, faster, safer and cheaper.

Ward, C.R.

1993-11-01T23:59:59.000Z

446

LIFAC sorbent injection desulfurization demonstration project  

SciTech Connect

In December 1990, the US Department of Energy selected 13 projects for funding under the Federal Clean Coal Technology Program (Round 3). One of the projects selected was the project sponsored by LIFAC North America, (LIFAC NA), titled LIFAC Sorbent Injection Desulfurization Demonstration Project.'' The host site for this $17 million, three-phase project is Richmond Power and Light's Whitewater Valley Unit No. 2 in Richmond, Indiana. The LIFAC technology uses upper-furnace limestone injection with patented humidification of the flue gas to remove 75--80% of the sulfur dioxide (SO{sub 2}) in the flue gas. In November 1990, after a ten (10) month negotiation period, LIFAC NA and the US DOE entered into a Cooperative Agreement for the design, construction, and demonstration of the LIFAC system. This report is the first Technical Progress Report covering the period from project execution through the end of December 1990. Due to the power plant's planned outage schedule, and the time needed for engineering, design and procurement of critical equipment, DOE and LIFAC NA agreed to execute the Design Phase of the project in August 1990, with DOE funding contingent upon final signing of the Cooperative Agreement.

Not Available

1991-01-01T23:59:59.000Z

447

Off site demonstrations for MWLID technologies  

SciTech Connect

Open demonstrations of technologies developed by the Office of Technology Development`s (QTD`s) Mixed Waste Landfill Integrated Demonstration (MWLID) should facilitate regulatory acceptance and speed the transfer and commercialization of these technologies. The purpose of the present project is to identify the environmental restoration needs of hazardous waste and/or mixed waste landfill owners within a 25-mile radius of Sandia National Laboratories (SNL). Most municipal landfills that operated prior to the mid-1980s accepted household/commercial hazardous waste and medical waste that included low-level radioactive waste. The locations of hazardous and/or mixed waste landfills within the State of New Mexico were. identified using federal, state, municipal and Native American tribal environmental records. The records reviewed included the US Environmental Protection Agency (EPA) Superfund Program CERCLIS Event/Site listing (which includes tribal records), the New Mexico Environment Department (NMED), Solid Waste Bureau mixed waste landfill database, and the City of Albuquerque Environmental Health Department landfill database. Tribal envirorunental records are controlled by each tribal government, so each tribal environmental officer and governor was contacted to obtain release of specific site data beyond what is available in the CERCLIS listings.

Williams, C. [Sandia National Labs., Albuquerque, NM (United States); Gruebel, R. [Tech Reps, Inc., Albuquerque, NM (United States)

1995-04-01T23:59:59.000Z

448

Atmospheric Aerosol Chemistry Analyzer: Demonstration of feasibility  

SciTech Connect

This is the final report of a three-year, Laboratory-Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The project objective was to demonstrate the technical feasibility of an Atmospheric Aerosol Chemistry Analyzer (AACA) that will provide a continuous, real-time analysis of the elemental (major, minor and trace) composition of atmospheric aerosols. The AACA concept is based on sampling the atmospheric aerosol through a wet cyclone scrubber that produces an aqueous suspension of the particles. This suspension can then be analyzed for elemental composition by ICP/MS or collected for subsequent analysis by other methods. The key technical challenge was to develop a wet cyclone aerosol sampler suitable for respirable particles found in ambient aerosols. We adapted an ultrasonic nebulizer to a conventional, commercially available, cyclone aerosol sampler and completed collection efficiency tests for the unit, which was shown to efficiently collect particles as small as 0.2 microns. We have completed the necessary basic research and have demonstrated the feasibility of the AACA concept.

Mroz, E.J.; Olivares, J.; Kok, G.

1996-04-01T23:59:59.000Z

449

Solid SCR Demonstration Truck Application | Department of Energy  

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

SCR Demonstration Truck Application Solid SCR Demonstration Truck Application Demonstrate the feasibility and performance of the FEV Solid SCR (Ammonium Carbamate) Technology...

450

U.S. Offshore Wind Advanced Technology Demonstration Projects...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Offshore Wind Advanced Technology Demonstration Projects Public Meeting Transcript for Offshore Wind Demonstrations U.S. Offshore Wind Advanced Technology Demonstration Projects...

451

West Valley Demonstration Project Low-Level Waste Shipment |...  

Office of Environmental Management (EM)

West Valley Demonstration Project Low-Level Waste Shipment West Valley Demonstration Project Low-Level Waste Shipment West Valley Demonstration Project Low-Level Waste Shipment...

452

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

SciTech Connect

We present the design and performance of a home-built high-pressure and high-temperature reactor equipped with a high-resolution scanning tunneling microscope (STM) for catalytic studies. In this design, the STM body, sample, and tip are placed in a small high pressure reactor ({approx}19 cm{sup 3}) located within an ultrahigh vacuum (UHV) chamber. A sealable port on the wall of the reactor separates the high pressure environment in the reactor from the vacuum environment of the STM chamber and permits sample transfer and tip change in UHV. A combination of a sample transfer arm, wobble stick, and sample load-lock system allows fast transfer of samples and tips between the preparation chamber, high pressure reactor, and ambient environment. This STM reactor can work as a batch or flowing reactor at a pressure range of 10{sup -13} to several bars and a temperature range of 300-700 K. Experiments performed on two samples both in vacuum and in high pressure conditions demonstrate the capability of in situ investigations of heterogeneous catalysis and surface chemistry at atomic resolution at a wide pressure range from UHV to a pressure higher than 1 atm.

Tao, Feng; Tang, David C.; Salmeron, Miquel; Somorjai, Gabor A.

2008-05-12T23:59:59.000Z

453

Heat pipe cooled reactors for multi-kilowatt space power supplies  

SciTech Connect

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

454

Reactor and Nuclear Systems Division (RNSD)  

NLE Websites -- All DOE Office Websites (Extended Search)

RNSD Home RNSD Home Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation Transport Reactor Physics Thermal Hydraulics & Irradiation Engineering Used Fuel Systems Staff Details (CV/Bios) Publications Org Chart Contact Us ORNL Staff Only Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation Transport Reactor Physics Thermal Hydraulics & Irradiation Engineering Used Fuel Systems Reactor and Nuclear Systems Division News Highlights U.S. Rep. Fleischmann touts ORNL as national energy treasure Martin Peng wins Fusion Power Associates Leadership Award

455

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.

Groh, Edward F. (Naperville, IL); Olson, Arne P. (Western Springs, IL); Wade, David C. (Naperville, IL); Robinson, Bryan W. (Oak Lawn, IL)

1984-01-01T23:59:59.000Z

456

Reactor monitoring and safeguards using antineutrino detectors  

Science Journals Connector (OSTI)

Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these very weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore orer an alternative means for verifying the power history and fissile inventory of a reactors, as part of International Atomic Energy Agency (IAEA) and other reactor safeguards regimes. Several erorts to develop this monitoring technique are underway across the globe.

N S Bowden

2008-01-01T23:59:59.000Z

457

Self isolating high frequency saturable reactor  

DOE Patents (OSTI)

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

Moore, James A. (Powell, TN)

1998-06-23T23:59:59.000Z

458

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

459

Dynamics of reverse osmosis in a standalone cogenerative nuclear reactor (Part II: load changes)  

Science Journals Connector (OSTI)

The coupling of desalination units with nuclear power plants has been studied in the present work â?? in this respect, the dynamic behaviour of the pressurised water reactor safety features, represented by the integrity of the fuel cladding, under some transient cases caused by secondary circuit load changes. A cosine-shaped heating through the reactor fuel is taken with its corresponding coolant lumps, to simulate realistic cases encountered in nuclear reactors. As an example, the mathematical model for the Westinghouse 3411 MWth pressurised water reactor, a familiar design with widely published design data was developed. The model consists of two parts; the first one is concerned with the dynamics of the primary side of the reactor; and the second, with the secondary side of the plant. To study the dynamics of the reactor, a 17-lumped parameters model was used. This is a first-order differential equation deduced from the first principles considering six groups of delayed neutrons. A computer program was developed using the Runge-Kutta method to solve these equations and to predict the behaviour of the state variables with time. Two case studies were considered as examples for normal transients. The developed model, which describes the dynamic response of the reactor, primary circuit and secondary circuit, has been analysed and verified with the relevant models. The first case, represented in Part 1 of this study, is concerned with the effect of changes of primary side transient reactivity, including the movement of the reactor control rods. The second one, represented in Part 2 of this study, considers the effect of the secondary side transient reactivity, as the load changes, on the system behaviour. As an example for the secondary side transient, load perturbations, such as load variations in standalone RO desalination units, are selected to study the effect of changing the secondary side conditions on the plant behaviour. The results showed that the reactor components and the fuel matrix should not be affected, and the fuel-cladding integrity is maintained within the safe limits, in all scram cases of RO desalination units coupled with nuclear power plants in the case of either a planned shutdown or an accidental shutdown.

Aly Karameldin; M.M. Shamloul; M.R. Shaalan; M.H. Esawy

2007-01-01T23:59:59.000Z

460

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

OPERATIONS OPERATIONS OBJECTIVE OP-1: Operations staff and management exhibit awareness of applicable requirements pertaining to CS operation, hazards, and reactor operations with the hydrogen-moderated CS. Through their actions, they have demonstrated a high-priority commitment to comply with these requirements. The level of knowledge of reactor operations and CS system operations managers and staff related to CS operations, hazards, and reactor operations with the hydrogen-moderated CS is adequate based on interviews. Sufficient numbers of qualified reactor operations and CS system operations staff and management are available to conduct and support safe operations with the hydrogen-moderated CS. (CR - 1, CR - 4, CR - 6) Criteria * Minimum staffing requirements have been established for operations and support

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

Physics-based multiscale coupling for full core nuclear reactor simulation  

Science Journals Connector (OSTI)

Abstract Numerical simulation of nuclear reactors is a key technology in the quest for improvements in efficiency, safety, and reliability of both existing and future reactor designs. Historically, simulation of an entire reactor was accomplished by linking together multiple existing codes that each simulated a subset of the relevant multiphysics phenomena. Recent advances in the MOOSE (Multiphysics Object Oriented Simulation Environment) framework have enabled a new approach: multiple domain-specific applications, all built on the same software framework, are efficiently linked to create a cohesive application. This is accomplished with a flexible coupling capability that allows for a variety of different data exchanges to occur simultaneously on high performance parallel computational hardware. Examples based on the KAIST-3A benchmark core, as well as a simplified Westinghouse AP-1000 configuration, demonstrate the power of this new framework for tackling—in a coupled, multiscale manner—crucial reactor phenomena such as CRUD-induced power shift and fuel shuffle.

Derek R. Gaston; Cody J. Permann; John W. Peterson; Andrew E. Slaughter; David Andrš; Yaqi Wang; Michael P. Short; Danielle M. Perez; Michael R. Tonks; Javier Ortensi; Ling Zou; Richard C. Martineau

2014-01-01T23:59:59.000Z

462

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

SciTech Connect

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

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

1994-06-01T23:59:59.000Z

463

Underground nuclear power station using self-regulating heat-pipe controlled reactors  

DOE Patents (OSTI)

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

Hampel, Viktor E. (Pleasanton, CA)

1989-01-01T23:59:59.000Z

464

An underground nuclear power station using self-regulating heat-pipe controlled reactors  

DOE Patents (OSTI)

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

Hampel, V.E.

1988-05-17T23:59:59.000Z

465

Research Program of a Super Fast Reactor  

SciTech Connect

Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is not breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie; Terai, Takayuki; Nagasaki, Shinya; Muroya, Yusa; Abe, Hiroaki [Nuclear Professional School / Department of Nuclear Engineering and Management, The University of Tokyo, Tokaimura, Naka-gun, Ibaraki, 319-1188 (Japan); Mori, Hideo [Department of Mechanical Engineering, Kyushu University (Japan); Akiba, Masato; Akimoto, Hajime; Okumura, Keisuke; Akasaka, Naoaki [Japan Atomic Energy Agency (Japan); GOTO, Shoji [Tokyo Electric Power Company (Japan)

2006-07-01T23:59:59.000Z

466

Nuclear reactor vessel fuel thermal insulating barrier  

DOE Patents (OSTI)

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

2013-03-19T23:59:59.000Z

467

History of Research Reactors at Brookhaven  

NLE Websites -- All DOE Office Websites (Extended Search)

History of Research Reactors at Brookhaven History of Research Reactors at Brookhaven Brookhaven National Laboratory has three nuclear reactors on its site that were used for scientific research. The reactors are all shut down, and the Laboratory is addressing environmental issues associated with their operations. photo of BGRR Brookhaven Graphite Research Reactor - Beginning operations in 1950, the graphite reactor was used for research in medicine, biology, chemistry, physics and nuclear engineering. One of the most significant achievements at this facility was the development of technetium-99m, a radiopharmaceutical widely used to image almost any organ in the body. The graphite reactor was shut down in 1969. Parts of it have been decommissioned, with the remainder to be addressed by 2011. More history

468

High-Temperature Superconductivity Cable Demonstration Projects  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Temperature Temperature Superconductivity Cable Demonstration Projects Superconductivity Power Equipment www.oe.energy.gov Phone: 202-586-1411 Office of Electricity Delivery and Energy Reliability, OE-1 U.S. Department of Energy - 1000 Independence Avenue, SW - Washington, DC 20585. Plugging America Into the Future of Power "A National Effort to Introduce New Technology into the Power Delivery Infrastructure" "In order to meet President Obama's ambitious energy goals, we must modernize the nation's electrical grid to improve the transmission, storage and reliability of clean energy across the country and help to move renewable energy from the places it can be produced to the places it can be used. The Department of Energy is working with industry partners to develop the

469

LIRR High-Speed Flywheel Demonstration  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

LIRR High-Speed Flywheel LIRR High-Speed Flywheel Demonstration Guy Sliker Program Manager Research & Technology Development New York Power Authority This project is part of the Joint Energy Storage Initiative between the New York State Energy Research and Development Authority (NYSERDA) and the Energy Storage Systems Program of the U.S. Department of Energy (DOE/ESS), and managed by Sandia National Laboratories (SNL). Presentation Layout Introduction Flywheel Description Project Reasoning Project Objective Project Participants Expected Benefits New York Power Authority Highlights A public benefit energy corporation founded 1931 Largest non-federal public electric utility in United States Wholesale power supplier throughout New York State and neighboring states as required by law

470

Visualizations for Real-time Pricing Demonstration  

SciTech Connect

In this paper, the visualization tools created for monitoring the operations of a real-time pricing demonstration system that runs at a distribution feeder level are presented. The information these tools provide gives insights into demand behavior from automated price responsive devices, distribution feeder characteristics, impact of weather on system’s development, and other significant dynamics. Given the large number of devices that bid into a feeder-level real-time electricity market, new techniques are explored to summarize the present state of the system and contrast that with previous trends as well as future projections. To better understand the system behavior and correctly inform decision-making procedures, effective visualization of the data is imperative.

Marinovici, Maria C.; Hammerstrom, Janelle L.; Widergren, Steven E.; Dayley, Greg K.

2014-10-13T23:59:59.000Z

471

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

SciTech Connect

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

Buden, D.

1993-08-11T23:59:59.000Z

472

Optimizing Neutron Thermal Scattering Effects in very High Temperature Reactors  

SciTech Connect

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

Hawari, Ayman; Ougouag, Abderrafi

2014-07-08T23:59:59.000Z

473

On the operator action analysis to reduce operational risk in research reactors  

Science Journals Connector (OSTI)

Abstract Human errors during operation and the resulting increase in operational risk are major concerns for nuclear reactors, just as they are for all industries. Additionally, human reliability analysis together with probabilistic risk analysis is a key element in reducing operational risk. The purpose of this paper is to analyze human reliability using appropriate methods for the probabilistic representation and calculation of human error to be used alongside probabilistic risk analysis in order to reduce the operational risk of the reactor operation. We present a technique for human error rate prediction and standardized plant analysis risk. Human reliability methods have been utilized to quantify different categories of human errors, which have been applied extensively to nuclear power plants. The Tehran research reactor is selected here as a case study, and after consultation with reactor operators and engineers human errors have been identified and adequate performance shaping factors assigned in order to calculate accurate probabilities of human failure.

Ramin Barati; Saeed Setayeshi

2014-01-01T23:59:59.000Z

474

In-reactor corrosion performance of ZIRLO{trademark}4 and Zircaloy-4  

SciTech Connect

In-reactor and long-term autoclave corrosion data have been obtained on ZIRCLO and three variants of Zircaloy-4: conventional (1.5% tin), low-tin, and beta-treated. In-reactor data from demonstration assemblies irradiated in the Virginia Power Company`s North Anna Unit 1 reactor demonstrate the superiority of ZIRLO and, to a lesser extent, low-tin Zircaloy-4 over conventional Zircaloy-4. After two cycles of irradiation to an assembly burnup of 37.8 GWD/MTU, the average axial peak corrosion of ZIRLO was 32% that of conventional Zircaloy-4. Low-tin and beta-treated materials displayed average peak oxides 76% and 150% of that formed on conventional Zircaloy-4, respectively. Autoclave corrosion tests of archive tubing have been performed in 633 K water, 672 K, 700 K, and 727 K steam, and in 633 K water containing 70 and 210 ppm lithium as the hydroxide. Correlation of the in-reactor data with the autoclave data indicates that the 633 K pure water test is the best qualitative indicator of in-reactor corrosion performance, and the 672 K steam test the poorest. Differences in in-reactor corrosion between ZIRLO and the Zircaloy-4 materials are consistent with the relative behavior of these materials in lithium hydroxide solutions. The relationships among the in-reactor and autoclave corrosion data, the microstructures, and the processing are discussed. In addition to improved corrosion resistance, ZIRLO exhibits improved dimensional stability over Zircaloy-4. The in-reactor creep of ZIRLO is confirmed to be about 80% of that of Zircaloy-4, and irradiation growth is observed to be about 50% that of Zircaloy-4. These data are also presented and discussed.

Sabol, G.P.; Weiner, R.A.; Stanutz, R.N. [Westinghouse Electric Corp., Pittsburgh, PA (United States). Nuclear Manufacturing Division; Comstock, R.J.; Larouere, P. [Virginia Power, Glen Allen, VA (United States). Innsbrook Technical Center

1994-12-31T23:59:59.000Z

475

Category:Smart Grid Projects - Regional Demonstrations | Open Energy  

Open Energy Info (EERE)

Demonstrations Demonstrations Jump to: navigation, search Smart Grid Regional Demonstrations Projects category. Pages in category "Smart Grid Projects - Regional Demonstrations" The following 16 pages are in this category, out of 16 total. B Battelle Memorial Institute, Pacific Northwest Division Smart Grid Demonstration Project C Center for the Commercialization of Electric Technologies Smart Grid Demonstration Project Columbus Southern Power Company (doing business as AEP Ohio) Smart Grid Demonstration Project Consolidated Edison Company of New York, Inc. Smart Grid Demonstration Project K Kansas City Power & Light Company Smart Grid Demonstration Project L Long Island Power Authority Smart Grid Demonstration Project L cont. Los Angeles Department of Water and Power Smart Grid Demonstration Project

476

Residential Energy Efficiency Demonstration: Hawaii and Guam Energy Improvement Technology Demonstration Project  

SciTech Connect

In order to meet its energy goals, the Department of Defense (DOD) has partnered with the Department of Energy (DOE) to rapidly demonstrate and deploy cost-effective renewable energy and energy-efficiency technologies. The scope of this project was to demonstrate tools and technologies to reduce energy use in military housing, with particular emphasis on measuring and reducing loads related to consumer electronics (commonly referred to as 'plug loads'), hot water, and whole-house cooling.

Earle, L.; Sparn, B.; Rutter, A.; Briggs, D.

2014-03-01T23:59:59.000Z

477

Nuclear reactor alignment plate configuration  

DOE Patents (OSTI)

An alignment plate that is attached to a core barrel of a pressurized water reactor and fits within slots within a top plate of a lower core shroud and upper core plate to maintain lateral alignment of the reactor internals. The alignment plate is connected to the core barrel through two vertically-spaced dowel pins that extend from the outside surface of the core barrel through a reinforcement pad and into corresponding holes in the alignment plate. Additionally, threaded fasteners are inserted around the perimeter of the reinforcement pad and into the alignment plate to further secure the alignment plate to the core barrel. A fillet weld also is deposited around the perimeter of the reinforcement pad. To accomodate thermal growth between the alignment plate and the core barrel, a gap is left above, below and at both sides of one of the dowel pins in the alignment plate holes through with the dowel pins pass.

Altman, David A; Forsyth, David R; Smith, Richard E; Singleton, Norman R

2014-01-28T23:59:59.000Z

478

Parallel Monte Carlo reactor neutronics  

SciTech Connect

The issues affecting implementation of parallel algorithms for large-scale engineering Monte Carlo neutron transport simulations are discussed. For nuclear reactor calculations, these include load balancing, recoding effort, reproducibility, domain decomposition techniques, I/O minimization, and strategies for different parallel architectures. Two codes were parallelized and tested for performance. The architectures employed include SIMD, MIMD-distributed memory, and workstation network with uneven interactive load. Speedups linear with the number of nodes were achieved.

Blomquist, R.N.; Brown, F.B.

1994-03-01T23:59:59.000Z

479

A novel reactor configuration for packed bed chemical-looping combustion of syngas  

Science Journals Connector (OSTI)

Abstract This study reports on the application of chemical looping combustion (CLC) in pressurized packed bed reactors using syngas as a fuel. High pressure operation of CLC in packed bed has a different set of challenges in terms of material properties, cycle and reactor design compared to fluidized bed operation. However, high pressure operation allows the use of inherently more efficient power cycles than low pressure fluidized bed solutions. This paper quantifies the challenges in high pressure operation and introduces a novel reactor concept with which those challenges can be addressed. Continuous cyclic operation of a packed bed CLC system is simulated in a 1D numerical reactor model. Importantly, it is demonstrated that the temperature profiles that can occur in a packed bed reactor as a result of the different process steps do not accumulate, and have a negligible effect on the overall performance of the system. Moreover, it has been shown that an even higher energy efficiency can be achieved by feeding the syngas from the opposite direction during the reduction step (i.e. countercurrent operation). Unfortunately, in this configuration mode, more severe temperature fluctuations occur in the reactor exhaust, which is disadvantageous for the operation of a downstream gas turbine. Finally, a novel reactor configuration is introduced in which the desired temperature rise for obtained hot pressured air suitable for a gas turbine is obtained by carrying out the process with two packed bed reactor in series (two-stage CLC). This is shown to be a good alternative to the single bed configuration, and has the added advantage of decreasing the demands on both the oxygen carrier and the reactor materials and design specification.

H.P. Hamers; F. Gallucci; P.D. Cobden; E. Kimball; M. van Sint Annaland

2013-01-01T23:59:59.000Z

480

The ARIES tokamak reactor study  

SciTech Connect

The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. 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. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

Not Available

1989-10-01T23:59:59.000Z