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Sample records for babcock and wilcox standard reactor

  1. Standard technical specifications - Babcock and Wilcox Plants: Bases (Sections 2.0-3.3). Volume 2, Revision 1

    SciTech Connect

    1995-04-01

    This NUREG contains the improved Standard Technical Specifications (STS) for Babcock and Wilcox (B&W) plants. Revision 1 incorporates the cumulative changes to Revision 0, which was published in September 1992. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, specifically the B&W Owners Group (BWOG), NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993 (58 FR 39132). Licensees are encouraged to upgrade their technical specifications consistent with those criteria and conforming, to the extent practical and consistent with the licensing basis for the facility, to Revision 1 to the improved STS. The Commission continues to place the highest priority on requests for complete conversions to the improved STS. Licensees adopting portions of the improved STS to existing technical specifications should adopt all related requirements, as applicable, to achieve a high degree of standardization and consistency.

  2. 75 FR 50009 - Babcock & Wilcox Nuclear Operations Group, Inc.; Establishment of Atomic Safety and Licensing Board

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-08-16

    ... COMMISSION Babcock & Wilcox Nuclear Operations Group, Inc.; Establishment of Atomic Safety and Licensing... & Wilcox Nuclear Operations Group, Inc. (Lynchburg, VA Facility). This proceeding concerns an Order Imposing Civil Monetary Penalty served upon the Licensee, Babcock & Wilcox Nuclear Operations Group,...

  3. APPLICATIONS ANALYSIS REPORT: BABCOCK AND WILCOX CYCLONE FURNACE

    EPA Science Inventory

    This document is an evaluation of the performance of the Babcock & Wilcox (B&W) Cyclone Furnace Vitrification Technology and its applicability as a treatment technique for soils contaminated with heavy metals, radionuclides, and organics. oth the technical and economic aspects of...

  4. Babcock and Wilcox assessment of the Pratt and Whitney XNR2000

    NASA Technical Reports Server (NTRS)

    Westerman, Kurt O.; Scoles, Stephen W.; Jensen, R. R.; Rodes, J. R.; Ales, M. W.

    1993-01-01

    Babcock & Wilcox performed four subtasks related to the assessment of the Pratt & Whitney XNR2000 nuclear reactor as follows: (1) cermet fuel element fabricability assessment; (2) mechanical design review of the reactor system; (3) neutronic analysis review; and (4) safety assessment. The results of the mechanical and physics reviews have been integrated into the reactor design. The results of the fuel and safety assessments are presented.

  5. Nuclear criticality safety for drums at Babcock and Wilcox

    SciTech Connect

    Alcorn, F.M.

    1997-12-01

    The Babcock and Wilcox Company (B&W) operates a nuclear fuel facility in Lynchburg, Virginia, processing uranium over the full range of possible enrichments (depleted to 97.65 wt% {sup 235}U). Nuclear fuel is produced for defense programs and for various research and test reactors worldwide. The facility has a uranium recovery operation that handles scrap produced at B&W as well as scrap from other U.S. Department of Energy sites. B&W also has a down-blending operation that is currently completing the down-blending of the fully enriched Project Sapphire Uranium to commercial-grade fuel (4 Wt% {sup 235}U). The facility generates approximately two hundred 55-gal drums of radioactive waste each month. Just a few years ago the number of waste drums on-site stood at {approximately}5000; however, through an aggressive waste reduction program, this number has been reduced to {approximately}2000. B&W strives to avoid storing uranium scrap in 55-gal drums; however, there are approximately sixty-four 55-gal drums of scrap on-site. Scrap is that material from which the uranium is recovered because of financial, contractual, or regulatory considerations; waste is that material destined for disposal. Whether waste or scrap, nuclear criticality safety is of paramount concern in the handling, processing, and storing of uranium-bearing drums at B&W. Any shipment complies with applicable U.S. Nuclear Regulatory Commission and U.S. Department of Transportation regulations.

  6. Radioactive waste shipments to Hanford retrievable storage from Babcock and Wilcox, Leechburg, Pennsylvania

    SciTech Connect

    Duncan, D.R.

    1994-02-14

    This report characterizes, as far as possible, the solid radioactive wastes generated by Babcock and Wilcox`s Park Township Plutonium Facility near Leechburg, Pennsylvania that were sent to retrievable storage at the Hanford Site. Solid waste as defined in this document is any containerized or self-contained material that has been declared waste. The objective is a description of characteristics of solid wastes that are or will be managed by the Restoration and Upgrades Program; gaseous or liquid effluents are discussed only at a summary level This characterization is of particular interest in the planning of transuranic (TRU) waste retrieval operations, including the Waste Receiving and Processing (WRAP) Facility, because Babcock and Wilcox generated greater than 2.5 percent of the total volume of TRU waste currently stored at the Hanford Site.

  7. TECHNOLOGY DEMONSTRATION SUMMARY. BABCOCK AND WILCOX CYCLONE FURNACE VITRIFICATION TECHNOLOGY (EPA/540/SR-92/017)

    EPA Science Inventory

    A Superfund Innovative Technology Evaluation (SITE) Demonstration of the Babcock & Wilcox Cyclone Furnace Vitrification Technology was conducted in November 1991. This Demonstration occurred at the Babcock & Wilcox (B&W) Alliance Research Center (ARC) in Alliance, OH. The B&W cyc...

  8. Conceptual designs and assessments of a coal gasification demonstration plant. Volume IV. Babcock and Wilcox process

    SciTech Connect

    Not Available

    1980-10-01

    This volume of the report contains detailed information on the conceptual design and assessment of the facility required to process approximately 20,000 tons per day of coal to produce medium Btu gas using the Babcock and Wilcox gasification process. The report includes process descriptions, flow diagrams and equipment lists for the various subsystems associated with the gasifiers along with descriptions of the overall facility. The facility is analyzed from both an economic and environmental standpoint. Problems of construction are addressed together with an overall design and construction schedule for the total facility. Resource requirements are summarized along with suggested development areas, both process and environmental.

  9. TVA commercial demonstration plant project. Volume 4. Plant based on Babcock and Wilcox gasifier. Final report

    SciTech Connect

    Not Available

    1980-11-01

    The baseline design of a coal gasification plant producing medium Btu gas, based upon the Babcock and Wilcox gasification process is documented in this report. The coal gasification plant consists of four identical modules, each with a capacity of approximately 5000 tons of coal per day as delivered to the gasifiers. The entire plant (four modules) produces 1205.7 MCFD of gas with a GHV value of approximately 299 Btu/SCF for a total heating value of about 360 billion Btu/day. The plant location is the rural site of Murphy Hill, located along the Tennessee River, some 30 miles east of Huntsville, Alabama. The desired product gas is a clean, medium-Btu gas suitable for pipeline distribution. The coal used for processing and for auxiliary boilers is a Kentucky No. 9 coal. The site is accessible by barge and road, with the plant receiving coal primarily by barge. Water needed for cooling and for process consumption will be drawn from the Tennessee River and will be treated by the plant water treatment facility. A description of the plant by major sections is included as well as flow diagrams, stream balances and lists of major equipment. Estimates of emissions and effluents are presented.

  10. Examination of Babcock and Wilcox tubes after exposure in an industrial waste incinerator

    SciTech Connect

    Keiser, J.R.; Ferber, M.K.; Longmire, H.F.; Walker, L.R.; Hindman, D.L.

    1996-06-01

    Seven ceramic tubes provided by, and in most cases manufactured by, Babcock and Wilcox were exposed in E. I. DuPont`s Wilmington, Delaware, hazardous waste incinerator. These tubes were subsequently examined at Oak Ridge National Laboratory to determine the effect of exposure on the strength and microstructural integrity of the tube materials. An unexposed tube section of one of the materials was also examined. Evaluation methods included c-ring compression tests, light microscopy, and electron microprobe spectroscopy. The c-ring compression tests revealed a very wide range in the strengths of the materials tested; the strongest was DuPont Lanxide Composites (DLC) silicon carbide particulate-strengthened alumina, and the weakest was the DLC Type B mixed-oxide material. The only material for which data on unexposed samples were available showed lower strength than the exposed material. Microstructural examination of the samples yielded minimal evidence of interaction of most of the tube materials with the components of the environment. Microprobe examination showed some segregation of yttrium in the matrix and along the surface of one of the PRD166/zirconia tubes and limited interaction of the fibers in the same tube with the components of the environment.

  11. DEMONSTRATION BULLETIN: CYCLONE FURNACE SOIL VITRI- FICATION TECHNOLOGY - BABCOCK & WILCOX

    EPA Science Inventory

    Babcock and Wilcox's (B&W) cyclone furnace is an innovative thermal technology which may offer advantages in treating soils containing organics, heavy metals, and/or radionuclide contaminants. The furnace used in the SITE demonstration was a 4- to 6-million Btu/hr pilot system....

  12. BABCOCK & WILCOX CYCLONE VITRIFICATION TECHNOLOGY FOR CONTAMINATED SOIL

    EPA Science Inventory

    The Babcock & Wilcox 6 million Btu/hr pilot cyclone furnace was successfully used in a 2-yr Superfund Innovative Technology Evaluation (SITE) Emerging Technology project to melt and vitrify an EPA Synthetic Soil Matrix (SSM) spiked with 7,000 ppm lead, 1,000 ppm cadmium, and 1,5...

  13. Modeling operator actions during a small break loss-of-coolant accident in a Babcock and Wilcox nuclear power plant

    SciTech Connect

    Ghan, L.S.; Ortiz, M.G.

    1991-01-01

    A small break loss-of-accident (SBLOCA) in a typical Babcock and Wilcox (B W) nuclear power plant was modeled using RELAP5/MOD3. This work was performed as part of the United States Regulatory Commission's (USNRC) Code, Scaling, Applicability and Uncertainty (CSAU) study. The break was initiated by severing one high pressure injection (HPI) line at the cold leg. Thus, the small break was further aggravated by reduced HPI flow. Comparisons between scoping runs with minimal operator action, and full operator action, clearly showed that the operator plays a key role in recovering the plant. Operator actions were modeled based on the emergency operating procedures (EOPs) and the Technical Bases Document for the EOPs. The sequence of operator actions modeled here is only one of several possibilities. Different sequences of operator actions are possible for a given accident because of the subjective decisions the operator must make when determining the status of the plant, hence, which branch of the EOP to follow. To assess the credibility of the modeled operator actions, these actions and results of the simulated accident scenario were presented to operator examiners who are familiar with B W nuclear power plants. They agreed that, in general, the modeled operator actions conform to the requirements set forth in the EOPs and are therefore plausible. This paper presents the method for modeling the operator actions and discusses the simulated accident scenario from the viewpoint of operator actions.

  14. Modeling operator actions during a small break loss-of-coolant accident in a Babcock and Wilcox nuclear power plant

    SciTech Connect

    Ghan, L.S.; Ortiz, M.G.

    1991-12-31

    A small break loss-of-accident (SBLOCA) in a typical Babcock and Wilcox (B&W) nuclear power plant was modeled using RELAP5/MOD3. This work was performed as part of the United States Regulatory Commission`s (USNRC) Code, Scaling, Applicability and Uncertainty (CSAU) study. The break was initiated by severing one high pressure injection (HPI) line at the cold leg. Thus, the small break was further aggravated by reduced HPI flow. Comparisons between scoping runs with minimal operator action, and full operator action, clearly showed that the operator plays a key role in recovering the plant. Operator actions were modeled based on the emergency operating procedures (EOPs) and the Technical Bases Document for the EOPs. The sequence of operator actions modeled here is only one of several possibilities. Different sequences of operator actions are possible for a given accident because of the subjective decisions the operator must make when determining the status of the plant, hence, which branch of the EOP to follow. To assess the credibility of the modeled operator actions, these actions and results of the simulated accident scenario were presented to operator examiners who are familiar with B&W nuclear power plants. They agreed that, in general, the modeled operator actions conform to the requirements set forth in the EOPs and are therefore plausible. This paper presents the method for modeling the operator actions and discusses the simulated accident scenario from the viewpoint of operator actions.

  15. Effect of flow leakage on the benchmarking of FLOWTRAN with Mark-22 mockup flow excursion test data from Babcock and Wilcox

    SciTech Connect

    Chen, Kuo-Fu.

    1992-10-01

    This report presents a revised analysis of the Babcock and Wilcox (B and W) downflow flow excursion tests that accounts for leakage between flow channels in the test assembly. Leak rates were estimated by comparing results from the downflow tests with those for upflow tests conducted using an identical assembly with some minor modifications. The upflow test assembly did not contain leaks. This revised analyses shows that FLOWTRAN with the SRS working criterion conservatively predicts onset of flow instability without using a local peaking factor to model heat transfer variations near the ribs.

  16. Compact Process Development at Babcock & Wilcox

    SciTech Connect

    Eric Shaber; Jeffrey Phillips

    2012-03-01

    Multiple process approaches have been used historically to manufacture cylindrical nuclear fuel compacts. Scale-up of fuel compacting was required for the Next Generation Nuclear Plant (NGNP) project to achieve an economically viable automated production process capable of providing a minimum of 10 compacts/minute with high production yields. In addition, the scale-up effort was required to achieve matrix density equivalent to baseline historical production processes, and allow compacting at fuel packing fractions up to 46% by volume. The scale-up approach of jet milling, fluid-bed overcoating, and hot-press compacting adopted in the U.S. Advanced Gas Reactor (AGR) Fuel Development Program involves significant paradigm shifts to capitalize on distinct advantages in simplicity, yield, and elimination of mixed waste. A series of compaction trials have been completed to optimize compaction conditions of time, temperature, and forming pressure using natural uranium oxycarbide (NUCO) fuel at packing fractions exceeding 46% by volume. Results from these trials are included. The scale-up effort is nearing completion with the process installed and operable using nuclear fuel materials. Final process testing is in progress to certify the process for manufacture of qualification test fuel compacts in 2012.

  17. Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    SciTech Connect

    1990-02-01

    This Preliminary Design Report (PDR) provides a detailed description of the design, analyses, and testing programs for the BR-100 cask. The BR-100 is a Type B(U) cask designed for transport by rail or barge. This report presents the preliminary analyses and tests which have been performed for the BR-100 and outlines the confirmatory analyses and tests which will be performed.

  18. STIRLING BOILER BY BABCOCK & WILCOX CO. (45,000 LB/HR CAPACITY), ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    STIRLING BOILER BY BABCOCK & WILCOX CO. (45,000 LB/HR CAPACITY), INSIDE BOILER HOUSE NO. 2. - Pittsburgh Steel Company, Monessen Works, Open Hearth Plant, Donner Avenue, Monessen, Westmoreland County, PA

  19. 75 FR 35846 - In the Matter of Babcock & Wilcox Nuclear Operations Group, Inc., Lynchburg, VA; Order Imposing...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-06-23

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION In the Matter of Babcock & Wilcox Nuclear Operations Group, Inc., Lynchburg, VA; Order Imposing Civil Monetary Penalty I Babcock & Wilcox Nuclear Operations Group, Inc., (Licensee) is the holder...

  20. Preliminary design report: Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    SciTech Connect

    1990-02-01

    The purpose of this document is to provide information on burnup credit as applied to the preliminary design of the BR-100 shipping cask. There is a brief description of the preliminary basket design and the features used to maintain a critically safe system. Following the basket description is a discussion of various criticality analyses used to evaluate burnup credit. The results from these analyses are then reviewed in the perspective of fuel burnups expected to be shipped to either the final repository or a Monitored Retrievable Storage (MRS) facility. The hurdles to employing burnup credit in the certification of any cask are then outlines and reviewed. the last section gives conclusions reached as to burnup credit for the BR-100 cask, based on our analyses and experience. All information in this study refers to the cask configured to transport PWR fuel. Boiling Water Reactor (BWR) fuel satisfies the criticality requirements so that burnup credit is not needed. All calculations generated in the preparation of this report were based upon the preliminary design which will be optimized during the final design. 8 refs., 19 figs., 16 tabs.

  1. An Analysis of the Corporate Merger between the Babcock & Wilcox Co. and J. Ray Mcdermott & Co., Inc.

    DTIC Science & Technology

    1980-09-01

    Convertible Preferred Stock and one share of $2.50 Preferred Stock. In addition, any B&W stock options still outstanding became options to purchase two shares... stock options ........... 2,392 332,940 Elimination of Baw’s Equity Accounts 14. Common Stock...in calculating the pro forms earnings per share, the weighted average number of shares outstanding has been adjusted to assume that B&W stock options exercisable

  2. 60. View from stock bin trestle looking northeast at Babcock ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    60. View from stock bin trestle looking northeast at Babcock & Wilcox type boilers (manufactured by Casey-Hedges Co., Chattanooga, TN) where washed furnace gas is burned with natural gas and coal to generate heat for steam. - Sloss-Sheffield Steel & Iron, First Avenue North Viaduct at Thirty-second Street, Birmingham, Jefferson County, AL

  3. Lithostratigraphic framework and production history of Wilcox in central Louisiana

    SciTech Connect

    Tye, R.S.; Wheeler, C.W.; Kimbrell, W.C.; Moslow, T.F.

    1988-02-01

    Complex fluvial, deltaic, and marine sedimentary processes active during deposition of the Wilcox Group in central Louisiana created multiple, discontinuous sandstones. Prolific hydrocarbon reservoirs developed in association with positive structural features or where sedimentary characteristics were favorable. Because the Wilcox Group contains numerous complex depocenters, it does not lend itself to easy regional correlation. Therefore, to better delineate the occurrence of hydrocarbon-bearing sediments and to promote further exploration, a five-fold lithostratigraphic framework is proposed for the Wilcox Group in Louisiana.

  4. Advanced reactors transition FY 1997 multi-year work plan WBS 7.3

    SciTech Connect

    Hulvey, R.K.

    1996-09-27

    This document describes in detail the work to be accomplised in FY 1997 and the out-years for the Advanced Reactors Transition (WBS 7.3) under the management of the Babcock & Wilcox Hanford Company. This document also includes specific milestones and funding profiles. Based upon the Fiscal Year 1997 Multi-Year Work Plan, the Department of Energy will provide authorization to perform the work described.

  5. Lithostratigraphic framework and production history of Wilcox in central Louisiana

    SciTech Connect

    Tye, R.S.; Wheeler, C.W.; Kimbrell, W.C.; Moslow, T.F.

    1988-01-01

    Complex fluvial, deltaic, and marine sedimentary processes active during deposition of the Wilcox Group in central Louisiana created multiple, discontinuous sandstones. Prolific hydrocarbon reservoirs developed in association with positive structural features or where sedimentary characteristics were favorable. Because the Wilcox Group contains numerous complex depocenters, it does not lend itself to easy regional correlation. Therefore, to better delineate the occurrence of hydrocarbon-bearing sediments and to promote further exploration, five-fold lithostratigraphic framework is proposed for the Wilcox Group in Louisiana. Five lithostratigraphic zones were defined on the basis of sedimentary processes and resistivity-log character. Their thickness and sand content were mapped within 21 parishes. These zones vary from 115 to 1,000 ft thick and the sand content in each ranges from 25 to 60%. All zones produce hydrocarbons, although production is geographically variable. Production in the updip Wilcox is by far the greatest in Zone III, whereas Zone I and Zone II are most productive in the downdip deep Wilcox shelf-margin trend, a paleo-shelf margin. Isopach and isolith maps indicate that the Wilcox was sourced from the northeast and northwest. All zoness display a strong north-south isopach grain in the northern two-thirds of the study area. East-west-oriented sand packages are presented southward along the paleo-shelf margin. Regional mapping facilitated the extension of presently existing production from fluvial/deltaic depocenters by following the sand packages to the west and southwest. Where these sandstones are associated with favorable structural/stratigraphic trapping conditions, new and lucrative Wilcox fields should be found.

  6. Preliminary Gulf Coast Coalbed Methane Exploration Maps: Depth to Wilcox, Apparent Wilcox Thickness and Vitrinite Reflectance

    USGS Publications Warehouse

    Barker, Charles E.; Biewick, Laura R.; Warwick, Peter D.; SanFilipo, John R.

    2000-01-01

    Strong economic controls on the viability of coalbed methane (CBM) prospects make coal geometry and coal property maps key elements in identifying sweet spots and production fairways. Therefore, this study seeks to identify the apparent prospective areas for CBM exploration in the Wilcox Group (Paleocene-Eocene) lignite and coalbeds by mapping net coal thickness, depth to coal, and coal rank (vitrinite reflectance). Economic factors are not considered in this CBM prospects study. Given the comparatively extensive gas pipeline and other production infrastructure development in the Gulf Coast Region, these factors seem less a control compared to other areas. However, open leasable public lands are minimal or nonexistent in the Gulf Coast region and access to the CBM prospects could be a problem.

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

    SciTech Connect

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

    2012-07-01

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

  8. Temperature and petroleum generation history of the Wilcox Formation, Louisiana

    USGS Publications Warehouse

    Pitman, Janet K.; Rowan, Elisabeth Rowan

    2012-01-01

    A one-dimensional petroleum system modeling study of Paleogene source rocks in Louisiana was undertaken in order to characterize their thermal history and to establish the timing and extent of petroleum generation. The focus of the modeling study was the Paleocene and Eocene Wilcox Formation, which contains the youngest source rock interval in the Gulf Coast Province. Stratigraphic input to the models included thicknesses and ages of deposition, lithologies, amounts and ages of erosion, and ages for periods of nondeposition. Oil-generation potential of the Wilcox Formation was modeled using an initial total organic carbon of 2 weight percent and an initial hydrogen index of 261 milligrams of hydrocarbon per grams of total organic carbon. Isothermal, hydrous-pyrolysis kinetics determined experimentally was used to simulate oil generation from coal, which is the primary source of oil in Eocene rocks. Model simulations indicate that generation of oil commenced in the Wilcox Formation during a fairly wide age range, from 37 million years ago to the present day. Differences in maturity with respect to oil generation occur across the Lower Cretaceous shelf edge. Source rocks that are thermally immature and have not generated oil (depths less than about 5,000 feet) lie updip and north of the shelf edge; source rocks that have generated all of their oil and are overmature (depths greater than about 13,000 feet) are present downdip and south of the shelf edge. High rates of sediment deposition coupled with increased accommodation space at the Cretaceous shelf margin led to deep burial of Cretaceous and Tertiary source rocks and, in turn, rapid generation of petroleum and, ultimately, cracking of oil to gas.

  9. CSRL-V ENDF/B-V library and thermal reactor and criticality safety benchmarks

    SciTech Connect

    Ford, W.E. III; Diggs, B.R.; Knight, J.R.; Greene, N.M.; Petrie, L.M.; Williams, M.L.

    1982-01-01

    CSRL-V, an ENDF/B-V 227-group neutron cross-section library which has recently been expanded to include Bondarenko factor data for unresolved resonance processing, was used to calculate performance parameters for a series of thermal reactor and criticality safety benchmarks. Among the thermal benchmarks calculated were the Babcock and Wilcox lattice critical experiments B and W-XIII and B and W-XX. These two slightly-enriched (2.46%) UO/sub 2/, water-moderated, tight-pitch lattice experiments were chosen because (a) they have similar U/sup 238/ resonance shielding characteristics as power reactor cores, and (b) they provide benchmark results representative of high-leakage and low-leakage lattices, respectively. Among the criticality safety benchmarks calculated were homogeneous, highly-enriched (93.2%) uranyl fluoride spheres with hydrogen-to-uranium ratios varying from 76 to 972.

  10. Babcock-Leighton Solar Dynamo: The Role of Downward Pumping and the Equatorward Propagation of Activity

    NASA Astrophysics Data System (ADS)

    Karak, Bidya Binay; Cameron, Robert

    2016-11-01

    The key elements of the Babcock-Leighton dynamos are the generation of poloidal field through decay and the dispersal of tilted bipolar active regions and the generation of toroidal field through the observed differential rotation. These models are traditionally known as flux transport dynamo models as the equatorward propagations of the butterfly wings in these models are produced due to an equatorward flow at the bottom of the convection zone. Here we investigate the role of downward magnetic pumping near the surface using a kinematic Babcock-Leighton model. We find that the pumping causes the poloidal field to become predominately radial in the near-surface shear layer, which allows the negative radial shear to effectively act on the radial field to produce a toroidal field. We observe a clear equatorward migration of the toroidal field at low latitudes as a consequence of the dynamo wave even when there is no meridional flow in the deep convection zone. Both the dynamo wave and the flux transport type solutions are thus able to reproduce some of the observed features of the solar cycle including the 11-year periodicity. The main difference between the two types of solutions is the strength of the Babcock-Leighton source required to produce the dynamo action. A second consequence of the magnetic pumping is that it suppresses the diffusion of fields through the surface, which helps to allow an 11-year cycle at (moderately) larger values of magnetic diffusivity than have previously been used.

  11. Standard Designations of Alloys for Aircraft and Missiles

    DTIC Science & Technology

    1961-05-24

    B & W Babcock and Wilcox Company Bethlehem Bethlehem Steel Company B-K Blaw - Knox Company Braeburn’ Braeburn Alloy Steel Corporation C-M Cannon...PennsylvaniaPi 3,enslaa Firth Vickers Stainless Steels, Ltd. Blaw - Knox Co. Sheffield, England National Alloy Division Pittsburgh 38, Pennsylvania General

  12. Magnetic flux transport and the sun's dipole moment - New twists to the Babcock-Leighton model

    NASA Technical Reports Server (NTRS)

    Wang, Y.-M.; Sheeley, N. R., Jr.

    1991-01-01

    The mechanisms that give rise to the sun's large-scale poloidal magnetic field are explored in the framework of the Babcock-Leighton (BL) model. It is shown that there are in general two quite distinct contributions to the generation of the 'alpha effect': the first is associated with the axial tilts of the bipolar magnetic regions as they erupt at the surface, while the second arises through the interaction between diffusion and flow as the magnetic flux is dispersed over the surface. The general relationship between flux transport and the BL dynamo is discussed.

  13. SAFT inspections for developing empirical database of fabrication flaws in nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Doctor, Steven R.; Schuster, George J.; Pardini, Allan F.

    1998-03-01

    The Pacific Northwest National Laboratory (PNNL) is developing a methodology for estimating the size and density distribution of fabrication flaws in U.S. nuclear reactor pressure vessels. This involves the nondestructive evaluation (NDE) of reactor pressure vessel materials and the destructive validation of the flaws found. NDE has been performed on reactor pressure vessel material made by Babcock & Wilcox and Combustion Engineering. A metallographic analysis is being performed to validate the flaw density and size distributions estimated from the 2500 indications of fabrication flaws that were detected and characterized in the very sensitive SAFT-UT (synthetic aperture focusing technique for ultrasonic testing) inspection data from the Pressure Vessel Research User Facility (PVRUF) vessel at Oak Ridge National Laboratory. Research plans are also described for expanding the work to include other reactor pressure vessel materials.

  14. Time-stratigraphic correlation of lower Wilcox valley-fill sequences, Colorado and Lavaca Counties, Texas

    SciTech Connect

    Devine, P.E.; Wheeler, D.M. )

    1989-09-01

    Late Paleocene and early Eocene lower Wilcox strata in southeast Texas are characterized principally by sandstone-rich deposits of fluvial and deltaic systems that prograded from a stable platform area into an unstable growth-faulted shelf margin setting. In contrast, incised valley systems, initiated during episodes of sea level lowering and filled dominantly with mud during subsequent transgressions, punctuate several lower Wilcox intervals. Further, valley-fill sequences are known to provide seals and/or reservoirs for a number of stratigraphically trapped hydrocarbon accumulations. Time-stratigraphic correlation of lower Wilcox strata provides improved differentiation of stacked valley-fill sequences and thereby more refined interpretation of depositional history and more accurate mapping for exploration purposes.

  15. Potentiometric surfaces in the Cockfield and Wilcox aquifers of southern and northeastern Arkansas, 2003

    USGS Publications Warehouse

    Yeatts, Daniel S.

    2004-01-01

    This report presents the results of water-level measurements made at wells in the Cockfield Formation and Wilcox Group of southern and northeastern Arkansas during 2003, and the water levels are displayed in potentiometric-surface maps and hydrographs. During March and April 2003, the water level was measured at 55 wells completed in the Cockfield aquifer, 13 wells completed in the Wilcox aquifer of southern Arkansas, and 43 wells completed in the Wilcox aquifer of northeastern Arkansas. The Cockfield Formation generally consists of discontinuous sand units interbedded with silt, clay, and lignite in southeastern Arkansas. Sand beds near the base of the Cockfield Formation constitute most of the Cockfield aquifer. Withdrawals from the Cockfield aquifer in the study area during 2000 totaled about 9 million gallons per day. The potentiometric surface of the Cockfield aquifer constructed from the 2003 water levels shows that regional direction of ground-water flow generally is towards the east and southeast, away from the outcrop, except in areas of intense ground-water withdrawals. Some local ground-water flow in the outcrop area is toward rivers that have eroded into the Cockfield Formation and deposited alluvium in south Bradley and Calhoun Counties (Ouachita River), and in north Dallas County (Saline River). An evaluation of 20 wells with water-level data from 1983 to 2003 shows that water levels in 15 wells have declined at a rate of -0.04 to -0.97 feet per year, and water levels in 5 wells have risen at a rate of 0.07 to 0.32 feet per year. An evaluation of the same 20 wells from 2000 to 2003 shows that water levels have declined in only 8 wells, and water levels have risen in 12 wells. The Wilcox Group is distributed throughout most of southern and eastern Arkansas. There are two study areas in southern and northeastern Arkansas. The Wilcox Group of the southern study area consists of interbedded clay, sandy clay, sand, and lignite. Thin discontinuous sand

  16. Origin of crude oil in the Wilcox trend of Louisiana and Mississippi: evidence of long-range migration

    SciTech Connect

    Sassen, R.; Tye, R.S.; Chinn, E.W.; Lemoine, R.C.

    1988-09-01

    Geochemical characterization of crude oils from Wilcox reservoirs in central Louisiana and southwestern Mississippi suggests they represent a single crude oil family that is distinct when compared with crude oils in deeper Tuscaloosa and Smackover reservoirs. This observation is consistent with geologic constraints that suggest an origin of crude oil from within the Wilcox Group itself. Although shales of the shallow Wilcox Group in central Louisiana and southwestern Mississippi contain gas-prone kerogen and are thermally immature, a more oil-prone source facies is present in marine shales of the deep Wilcox Group in south-central Louisiana. Thermal maturity measurements based on pyrolysis suggest a broad area of effective Wilcox source rock in southcentral Louisiana. Migration distances from source to reservoir rocks of the downdip Wilcox Trend of south-central Louisiana appear to be relatively short. However, long-range updip migration (sometimes > 100 km) from deeply buried Wilcox source facies provides the best explanation for emplacement of crude oil in the shallow Wilcox Trend of central Louisiana and southwestern Mississippi.

  17. Depositional framework and genesis of Wilcox Submarine Canyon systems, Northwest Gulf Coast

    SciTech Connect

    Galloway, W.F.; Dinqus, W.F.; Paige, R.E.

    1988-01-01

    Wilcox (late Paleocene-early Eocene) slope systems of the Texas coastal plain contain two families of paleosubmarine canyons that exhibit distinctly different characteristics and stratigraphic settings: Yoakum and Lavaca type canyons occur as widely separated features within the generally retrogradational middle Wilcox interval. Four such canyons exhibit high length to width ratios, extend far updip of the contemporaneous shelf edge, were excavated deeply into paralic and coastal-plain deposits, and were filled primarily by mud. Fills consist of a lower onlapping unit and capping progradational deposits that are genetically related to deposition of the upper Wilcox fluvial-deltaic sequence. Significantly, the canyon fills correlate with widespread transgressive marine mudstones (the Yoakum shale-Sabinetown Formation and ''Big Shale''). In contrast, Lavaca-type canyons form a system of erosional features created along the rapidly prograding, unstable lower Wilcox continental margin. Comparative analysis of the two canyon system suggests a general process model for submarine canyon formation on prograding basin margins. Key elements are depositional loading of the continental margin creating instability, initiation of a large-scale slump, family of slumps, or listric bedding-plane fault creating a depression or indentation in the margin, and headward and lateral expansion of the depression by slumping and density-underflow erosion. Extent of canyon evolution varies according to time and submerged space available for maturation; short, broad canyons form on narrow shelves of actively prograding margins, and elongate mature canyons form in retrogradational or transgressive settings.

  18. WILCOX COUNTY, ALABAMA--A STUDY OF SOCIAL, ECONOMIC, AND EDUCATIONAL BANKRUPTCY. REPORT OF AN INVESTIGATION.

    ERIC Educational Resources Information Center

    BROADUS, JAMES; AND OTHERS

    THE REQUEST FOR THIS INVESTIGATION BY THE SPECIAL COMMITTEE OF THE NATIONAL EDUCATION ASSOCIATION COMMISSION ON PROFESSIONAL RIGHTS AND RESPONSIBILITIES RESULTED FROM THE FIRING OF NINE NEGRO TEACHERS IN WILCOX COUNTY. THE STUDY ITSELF IS MORE INCLUSIVE, INCORPORATING THE FINDINGS AND CONCLUSIONS OF SEPARATE STUDIES IN POVERTY, SCHOOL FINANCE,…

  19. Qualification of a Tritium-Producing Target for the light water reactor application

    SciTech Connect

    Apley, W.J.; Beeman, G.H.; Ethridge, J.L.

    1992-06-01

    The Pacific Northwest Laboratory (PNL) currently manages the Light Water Reactor (LWR) Tritium Target Development Project (TTDP) for the Office of New Production Reactors (NP), US Department of Energy. The project`s objective is to demonstrate and qualify a high-temperature LWR tritium target system with fabrication and extraction processes sufficiently confirmed to ensure a deployable system consistent with variable tritium production demands. The project has also examined and reported on technical and institutional issues associated with acquisition and conversion of the yet uncompleted Washington Public Power Supply System Unit I (WNP-1) for tritium production purposes. WNP-1 is a 63% complete, Babcock and Wilcox, 3800 MW thermal, 205 assembly, pressurized water reactor located at Hanford, Washington. WNP-1 has served as the reference LWR plant for the technical evaluation and target development activities. This report discussed the evaluation and development necessary to provide a complete LWR target qualification package.

  20. Structural and hydrological parameters modeling of Wilcox group, Central Louisiana and Mississippi

    SciTech Connect

    Zuo, H. )

    1993-09-01

    In a 900 mi[sup 2] area between Louisiana and Mississippi, 300 well logs were studied. Three major sand formations in middle Wilcox and five target log picks within upper and middle Wilcox were correlated. The lithofacies variation appear to be related to the sequence and eustatic boundaries. The temperature and salinity values modeled were also obtained from electric-log and BHT (bottom hole temperature) data. The relationships among the temperature and salinity distributions, topography, stratigraphy, and subsurface structure were studied. Several real images of three-dimensional topography, formation structure, and stratigraphic boundary were reconstructed by using the well log data. This detailed study suggests that the productive oil fields along the east and west flank of Mississippi Embayment are related to the salinity anomalies, and are controlled by structural and stratigraphic features. From the three-dimensional maps, one can confidently interpret the structural and stratigraphic changes with time and basin growth history. For instance, it appears that the Holocene Mississippi River course in the study area is nearly coincident with the main depositional axis of middle and late Wilcox deposition. The current Mississippi River channel, as well as the basin's depositional axis of 20-50 m.y. ago, appear to have been affected by some regional structural events. Also, a north-south salinity anomaly in the east part of study area is a possible indicator of a northeast-trending wrench fault along the Mississippi River.

  1. 78 FR 59981 - Proposed Revision to Physical Security-Standard Design Certification and Operating Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-09-30

    ... COMMISSION Proposed Revision to Physical Security--Standard Design Certification and Operating Reactors...: Section 13.6.2 ``Physical Security--Standard Design Certification and Operating Reactors.'' The NRC seeks... security reviews of design certification applications. DATES: Comments must be filed no later than...

  2. Babcock-Leighton solar dynamo: the role of downward pumping and the equatorward propagation of activity

    NASA Astrophysics Data System (ADS)

    Karak, Bidya Binay; Cameron, Robert

    2016-05-01

    We investigate the role of downward magnetic pumping near the surface using a kinematic Babcock-Leighton model. We find that the pumping causes the poloidal field to become predominately radial in the near-surface shear layer. This allows the negative radial shear in the near-surface layer to effectively act on the radial field to produce a toroidal field. Consequently, we observe a clear equatorward migration of the toroidal field at low latitudes even when there is no meridional flow in the deep CZ. We show a case where the period of a dynamo wave solution is approximately 11 years. Flux transport models are also shown with periods close to 11 years. Both the dynamo wave and flux transport dynamo are thus able to reproduce some of the observed features of solar cycle. The main difference between the two types of dynamo is the value of $\\alpha$ required to produce dynamo action. In both types of dynamo, the surface meridional flow helps to advect and build the polar field in high latitudes, while in flux transport dynamo the equatorward flow near the bottom of CZ advects toroidal field to cause the equatorward migration in butterfly wings and this advection makes the dynamo easier by transporting strong toroidal field to low latitudes where $\\alpha$ effect works. Another conclusion of our study is that the magnetic pumping suppresses the diffusion of fields through the photospheric surface which helps to achieve the 11-year dynamo cycle at a moderately larger value of magnetic diffusivity than has previously been used.

  3. Stratigraphic relationship between Odontogryphaea Thirsae beds and big shale of Wilcox (Paleocene-Eocene) in Louisiana

    SciTech Connect

    Glawe, L.N. )

    1989-09-01

    The relationship between beds of the oyster Odontogryphaea thirsae, of the surface Wilcox Group exposures in the Sabine uplift area of northwestern Louisiana and the Big Shale, a well-known subsurface mapping unit in central Louisiana, has been unclear. In a paleontological study of 2,158 ft of continuous Wilcox core from Sabine Parish, planktonic foraminiferal biostratigraphic zones were determined that include: Globorotalia angulata and G. pusilla (P3), G. Pseudomenardii (P4), and G. velascoensis (P5). The O. thirsae bed was assigned to the G. Pseudomenardii zone. Electrical-log cross sections were used to demonstrate (1) the correlation between the O. thirsae bed in the core hole and the O. thirsae bed exposed at the surface at Many, Louisiana, a distance of 6 mi and (2) the correlation between the base of the 10-ft thick shale beneath the O. thirsae bed in the core hole and the base of the Big Shale as recognized in the central Louisiana subsurface. Based on its assigned biostratigraphic zone, The O. thirsae bed in the updip Wilcox core hole is equivalent, in part, to the Big Shale of the basinal Wilcox of east-central Louisiana, which previously has been included in the G. pseudomenardii zone. Paleontological recognition of a thin marine unit 1 ft above the O. thirsae bed in the core hole on the southern flank of the Sabine uplift provides documentation of the transgressive character of the Big Shale - O. thirsae sequence and is evidence of the TP 2.1 cycle in Louisiana.

  4. Potentiometric surfaces and water-level trends in the Cockfield (upper Claiborne) aquifer in southern Arkansas and the Wilcox (lower Wilcox) aquifer of northeastern and southern Arkansas, 2012

    USGS Publications Warehouse

    Rodgers, Kirk D.

    2015-01-01

    Linear regression analysis of long-term hydrographs was used to determine the mean annual water-level rise and decline in the Wilcox aquifer in the northeastern and southern areas of Arkansas. In the northeastern area, the mean annual water level declined in all seven counties. The mean annual declines ranged from -0.55 ft/yr in Craighead County to -1.46 ft/yr in St. Francis County. In the southern area, the annual rise and decline calculations for wells with over 20 years of records indicate rising and declining water levels in Clark, Hot Spring, and Nevada Counties. The mean annual water level declined in all counties except Hot Spring County.

  5. Potentiometric Surfaces and Water-Level Trends in the Cockfield (Upper Claiborne) and Wilcox (Lower Wilcox) Aquifers of Southern and Northeastern Arkansas, 2009

    USGS Publications Warehouse

    Pugh, Aaron L.

    2010-01-01

    Eocene-age sand beds near the base of the Cockfield Formation of Claiborne Group constitute the aquifer known locally as the Cockfield aquifer. Upper-Paleocene age sand beds within the lower parts of the Wilcox Group constitute the aquifer known locally as the Wilcox aquifer. In 2005, reported water withdrawals from the Cockfield aquifer in Arkansas totaled 16.1 million gallons per day, while reported water withdrawals from the Wilcox aquifer in Arkansas totaled 27.0 million gallons per day. Major withdrawals from these units were for industrial and public water supplies with lesser but locally important withdrawals for commercial, domestic, and agricultural uses. During February 2009, 56 water-level measurements were made in wells completed in the Cockfield aquifer and 57 water-level measurements were made in wells completed in the Wilcox aquifer. The results from the 2009 water-level measurements are presented in potentiometric-surface maps and in combination with previous water-level measurements. Trends in water-level change over time within the two aquifers are investigated using water-level difference maps and well hydrographs. Water-level difference maps were constructed for each aquifer using the difference between depth to water measurements made in 2003 to 2009. Well hydrographs for each aquifer were constructed for wells with 20 or more years of historical water-level data. The hydrographs were evaluated individually using linear regression to calculate the annual rise or decline in water levels, and by aggregating the regression results by county and statistically summarizing for the range, mean, and median water-level change in each county. The 2009 potentiometric surface of the Cockfield aquifer map indicates the regional direction of groundwater flow generally towards the east and southeast, except in two areas of intense groundwater withdrawals that have developed into cones of depression. The lowest water-level altitude measured was 43 feet and the

  6. Conversion and standardization of US university reactor fuels using LEU, status 1989

    SciTech Connect

    Brown, K.R.; Matos, J.E.; Argonne National Lab., IL )

    1989-01-01

    In 1986, the US Department of Energy initiated a program to change the fuel used in most of the US university research reactors using HEU (93%) to LEU({lt}20{percent}) in order to minimize the risk of theft or diversion of this weapons-useable material. An important consideration in the LEU conversion planning process has been the desire to standardize the fuels that are used and to enhance the performance and utilization of the reactors. This paper describes the current status of this conversion process and the plans and schedules to complete an orderly transition from HEU to LEU fuel in most of these reactors. To date, three university reactors have been converted to LEU fuel, completed safety documentation for three reactors is being evaluated by the USNRC, and work on the safety documentation for six reactors is in progress. 13 refs., 9 tabs.

  7. Conversion and standardization of university reactor fuels using low-enrichment uranium: Plans and schedules

    SciTech Connect

    Young, H.H.; Brown, K.R.; Matos, J.E.

    1986-01-01

    The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. To minimize this risk, the US Nuclear Regulatory Commission issued its final rule on ''Limiting the Use of Highly Enriched Uranium in Domestically Licensed Research and Test Reactors,'' in February 1986. This paper describes the plans and schedules developed by the US Department of Energy to coordinate an orderly transition from HEU to LEU fuel in most of these reactors. An important element in the planning process has been the desire to standardize the LEU fuels used in US university reactors and to enhance the performance and utilization of a number of these reactors. The program is estimated to cost about $10 million and to last about five years.

  8. Depositional environments of the Wilcox Group, Texas Gulf Coast: Stratigraphic and early diagenetic signatures

    SciTech Connect

    May, J.A.; Stonecipher, S.A. )

    1990-09-01

    Deposition of the late Paleocene-early Eocene Wilcox Group is controversial. Are Wilcox reservoirs entirely of shallow-marine origin, or are basinal turbidites also present The authors analyzed over 5,000 ft of core from 15 wells along the Texas Gulf Coast to constrain the environments of deposition. They attribute all cores examined to date to 12 subenvironments of the delta plain to continental shelf. These include distributary channel, lake, marine bay, crevasse-splay delta, shoreface, lagoon, tidal flat, tidal channel, distributary-mouth bar, distal bar, prodelta, and shelf. Each subenvironment displays a characteristic well-log signature. Criteria for recognition in core include grain-size variations, physical sedimentary structures, trace fossils, mineralogy, bedding styles, and vertical sequences, all resulting from the interplay of specific physical, biological, and chemical processes operative in each subenvironment. They did not identify any submarine-fan deposits. They also attempted to determine the importance of depositional facies and provenance on diagenetic trends. Early diagenetic patterns appear to be related to factors such as sediment texture, detrital composition, organic content, and original water chemistry, which were, in turn, controlled directly or indirectly by depositional environment. Rapid lateral and vertical changes in depositional environments produced markedly different early diagenetic patterns in sand units only a few feet or even inches apart. Thus, diagenetic facies defined on the basis of texture, composition, and cements can be used to complement, and test, their interpretations of depositional environments based solely on traditional sedimentologic and genetic-sequence criteria.

  9. Well log and 2D seismic data character of the Wilcox Group in south-central Louisiana

    USGS Publications Warehouse

    Enomoto, Catherine B.

    2014-01-01

    The Wilcox Group is productive in updip areas of Texas and Louisiana from fluvial, deltaic, and near-shore marine shelf sandstones. The reported presence of porous sandstones at 29,000 feet within the Wilcox Group containing about 200 feet of gas in the Davy Jones 1 discovery well in the offshore Louisiana South Marsh Island area illustrates a sand-rich system developed during the Paleocene and early Eocene. This study describes some of the well log and reflection seismic data characteristics of the slope and basin-floor reservoirs with gas-discovery potential that may be in the area between the producing trend onshore Louisiana and the offshore discovery.

  10. Assessement of Codes and Standards Applicable to a Hydrogen Production Plant Coupled to a Nuclear Reactor

    SciTech Connect

    M. J. Russell

    2006-06-01

    This is an assessment of codes and standards applicable to a hydrogen production plant to be coupled to a nuclear reactor. The result of the assessment is a list of codes and standards that are expected to be applicable to the plant during its design and construction.

  11. On the Meaning of Formative Measurement and How It Differs from Reflective Measurement: Comment on Howell, Breivik, and Wilcox (2007)

    ERIC Educational Resources Information Center

    Bagozzi, Richard P.

    2007-01-01

    D. Howell, E. Breivik, and J. B. Wilcox (2007) have presented an important and interesting analysis of formative measurement and have recommended that researchers abandon such an approach in favor of reflective measurement. The author agrees with their recommendations but disagrees with some of the bases for their conclusions. He suggests that…

  12. Obituary: Horace Welcome Babcock, 1912-2003

    NASA Astrophysics Data System (ADS)

    Vaughan, Arthur Harris

    2003-12-01

    sunspot cycles. Until about 1957 this work had been done at the Hale Solar Laboratory on Holladay Road in Pasadena. Improved models of the magnetograph developed by Robert F. Howard, in collaboration with Horace, went into operation in the 150-foot solar tower telescope at Mount Wilson in 1959 and later, and similar instruments are now employed at many other solar observatories. In 1961 Horace proposed an explanation of the Sun's 22-year magnetic cycle that contained many of the features still embodied in contemporary theoretical models of the phenomenon. The advance in our understanding of solar and stellar magnetism brought forth by Horace Babcock is a worthy sequel to the pioneering efforts initiated by George E. Hale early in the twentieth century. Faced with the growing obsolescence of the Carnegie Institution of Washington's facilities at Mount Wilson along with the competition from Caltech's 200-inch telescope, the Carnegie Trustees in 1963 adopted the idea of founding a major observatory in the Southern Hemisphere as its master plan for modernizing the astronomical facilities of the Institution. Upon becoming Director of the Mount Wilson and Palomar Observatories in 1964, Horace Babcock embraced the job of carrying out this plan, although it meant giving up his own science. Beginning in 1963, and with his usual ingenuity, Horace developed apparatus for measuring astronomical ``seeing." In collaboration with John Irwin and others, he carried out site surveys in Chile, Australia and New Zealand with the aim of selecting the best available location for the anticipated array of large telescopes. Some five years of exploration led, in 1968, to the selection and purchase of a 276 square-kilometer tract on Cerro Las Campanas in north central Chile as the site for the new observatory. Babcock and Irwin had first climbed to its summit, on foot, in October 1966. The team Horace assembled to build the observatory and its infrastructure proved equal to the high standards he

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

    SciTech Connect

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

    2011-10-31

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

  14. Compiled reports on the applicability of selected codes and standards to advanced reactors

    SciTech Connect

    Benjamin, E.L.; Hoopingarner, K.R.; Markowski, F.J.; Mitts, T.M.; Nickolaus, J.R.; Vo, T.V.

    1994-08-01

    The following papers were prepared for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission under contract DE-AC06-76RLO-1830 NRC FIN L2207. This project, Applicability of Codes and Standards to Advance Reactors, reviewed selected mechanical and electrical codes and standards to determine their applicability to the construction, qualification, and testing of advanced reactors and to develop recommendations as to where it might be useful and practical to revise them to suit the (design certification) needs of the NRC.

  15. Interpretational Confounding Is Due to Misspecification, Not to Type of Indicator: Comment on Howell, Breivik, and Wilcox (2007)

    ERIC Educational Resources Information Center

    Bollen, Kenneth A.

    2007-01-01

    R. D. Howell, E. Breivik, and J. B. Wilcox (2007) have argued that causal (formative) indicators are inherently subject to interpretational confounding. That is, they have argued that using causal (formative) indicators leads the empirical meaning of a latent variable to be other than that assigned to it by a researcher. Their critique of causal…

  16. The simulation of thermohydraulic phenomena in a pressurized water reactor primary loop

    SciTech Connect

    Popp, M

    1987-01-01

    Several important fluid flow and heat transfer phenomena essential to nuclear power reactor safety were investigated. Scaling and modeling laws for pressurized water reactors are reviewed and a new scaling approach focusing on the overall loop behavior is presented. Scaling criteria for one- and two-phase natural circulation are developed, as well as a simplified model describing the first phase of a small break loss of coolant accident. Reactor vessel vent valve effects are included in the analysis of steady one-phase natural circulation flow. Two new dimensionless numbers, which uniquely describe one-phase flow in natural circulation loops, were deduced and are discussed. A scaled model of the primary loop of a typical Babcock and Wilcox reactor was designed, built, and tested. The particular prototype modeled was the TMI unit 2 reactor. The electrically heated, stainless steel model operates at a maximum pressure of 300 psig and has a maximum heat input of 188 kW. The model is about 4 times smaller in height than the prototype reactor, with a nominal volume scale of 1:500. Experiments were conducted establishing subcooled natural circulation in the model loop. Both steady flow and power transients were investigated.

  17. Geopressured Geothermal Resource and Recoverable Energy Estimate for the Wilcox and Frio Formations, Texas (Presentation)

    SciTech Connect

    Esposito, A.; Augustine, C.

    2011-10-01

    An estimate of the total and recoverable geopressured geothermal resource of the fairways in the Wilcox and Frio formations is made using the current data available. The flow rate of water and methane for wells located in the geopressured geothermal fairways is simulated over a 20-year period utilizing the TOUGH2 Reservoir Simulator and research data. The model incorporates relative permeability, capillary pressure, rock compressibility, and leakage from the bounding shale layers. The simulations show that permeability, porosity, pressure, sandstone thickness, well spacing, and gas saturation in the sandstone have a significant impact on the percent of energy recovered. The results also predict lower average well production flow rates and a significantly higher production of natural gas relative to water than in previous studies done from 1975 to 1980. Previous studies underestimate the amount of methane produced with hot brine. Based on the work completed in this study, multiphase flow processes and reservoir boundary conditions greatly influence the total quantity of the fluid produced as well as the ratio of gas and water in the produced fluid.

  18. Steps towards verification and validation of the Fetch code for Level 2 analysis, design, and optimization of aqueous homogeneous reactors

    SciTech Connect

    Nygaard, E. T.; Pain, C. C.; Eaton, M. D.; Gomes, J. L. M. A.; Goddard, A. J. H.; Gorman, G.; Tollit, B.; Buchan, A. G.; Cooling, C. M.; Angelo, P. L.

    2012-07-01

    Babcock and Wilcox Technical Services Group (B and W) has identified aqueous homogeneous reactors (AHRs) as a technology well suited to produce the medical isotope molybdenum 99 (Mo-99). AHRs have never been specifically designed or built for this specialized purpose. However, AHRs have a proven history of being safe research reactors. In fact, in 1958, AHRs had 'a longer history of operation than any other type of research reactor using enriched fuel' and had 'experimentally demonstrated to be among the safest of all various type of research reactor now in use [1].' While AHRs have been modeled effectively using simplified 'Level 1' tools, the complex interactions between fluids, neutronics, and solid structures are important (but not necessarily safety significant). These interactions require a 'Level 2' modeling tool. Imperial College London (ICL) has developed such a tool: Finite Element Transient Criticality (FETCH). FETCH couples the radiation transport code EVENT with the computational fluid dynamics code (Fluidity), the result is a code capable of modeling sub-critical, critical, and super-critical solutions in both two-and three-dimensions. Using FETCH, ICL researchers and B and W engineers have studied many fissioning solution systems include the Tokaimura criticality accident, the Y12 accident, SILENE, TRACY, and SUPO. These modeling efforts will ultimately be incorporated into FETCH'S extensive automated verification and validation (V and V) test suite expanding FETCH'S area of applicability to include all relevant physics associated with AHRs. These efforts parallel B and W's engineering effort to design and optimize an AHR to produce Mo99. (authors)

  19. Pore water salinity as a tool for evaluating reservoir continuity and fluid migration pathways in the Wilcox Group of central Louisiana

    SciTech Connect

    Funayama, M.; Hanor, J.S.

    1995-10-01

    Spatial variations in pore water salinity derived from SP logs can provide useful information on the degree of hydrologic continuity and compartmentalization of sedimentary sequences and can thus aid in evaluating both local and regional reservoir continuity and in determining possible migration pathways for hydrocarbons. Applying these techniques in highly mature hydrocarbon-producing districts, such as the Wilcox, however, is complicated by the fact that many of the older SP logs lack header information, such as bottom hole temperatures, BHT, and values for R{sub mf}, required to make salinity calculations from SP response. We have determined the statistical relation between R{sub mf} and R{sub m}, mud weight, and year of measurement for wells having complete header information in the Wilcox Group of central Louisiana. We have also determined BHT-depth relations in this region from nearly 700 paired measurements. The correlation equations thus generated can be used to calculate formation water salinities in key areas of the Wilcox where only older logs lacking complete header information are available. Formation water salinities in the shallow, updip Wilcox in central Louisiana vary from less than 35g/L to over 100 g/L. Salinities generally increase with stratigraphic depth and with distance downdip to the south, until the Wilcox starts to become overpresured. There is no evidence for significant regional vertical or laterral hydrologic compartmentalization in the shallow updip Wilcox in spite of the presence of numerous shaly interbeds. Some of the saline water in the shallow Wilcox mat have been derived from issolution of salt domes 100 km or more to the north. Our work supports the concept that much of the shallow Wilcox is sufficiently continuous hydrologically to have permitted large-scale vertical and lateral fluid migration and solute transport.

  20. Organic petrology and coalbed gas content, Wilcox Group (Paleocene-Eocene), northern Louisiana

    USGS Publications Warehouse

    Hackley, P.C.; Warwick, P.D.; Breland, F.C.

    2007-01-01

    Wilcox Group (Paleocene-Eocene) coal and carbonaceous shale samples collected from four coalbed methane test wells in northern Louisiana were characterized through an integrated analytical program. Organic petrographic analyses, gas desorption and adsorption isotherm measurements, and proximate-ultimate analyses were conducted to provide insight into conditions of peat deposition and the relationships between coal composition, rank, and coalbed gas storage characteristics. The results of petrographic analyses indicate that woody precursor materials were more abundant in stratigraphically higher coal zones in one of the CBM wells, consistent with progradation of a deltaic depositional system (Holly Springs delta complex) into the Gulf of Mexico during the Paleocene-Eocene. Comparison of petrographic analyses with gas desorption measurements suggests that there is not a direct relationship between coal type (sensu maceral composition) and coalbed gas storage. Moisture, as a function of coal rank (lignite-subbituminous A), exhibits an inverse relationship with measured gas content. This result may be due to higher moisture content competing for adsorption space with coalbed gas in shallower, lower rank samples. Shallower ( 600??m) coal samples containing less moisture range from under- to oversaturated with respect to their CH4 adsorption capacity.

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

    SciTech Connect

    Scarangella, M. J.

    2012-07-01

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

  2. Potentiometric Surfaces and Water-Level Trends in the Cockfield and Wilcox Aquifers of Southern and Northeastern Arkansas, 2006

    USGS Publications Warehouse

    Schrader, T.P.

    2007-01-01

    The Cockfield Formation of Claiborne Group and the Wilcox Group contain aquifers that provide sources of ground water in southern and northeastern Arkansas. In 2000, about 9.9 million gallons per day was withdrawn from the Cockfield Formation of Claiborne Group and about 22.2 million gallons per day was withdrawn from the Wilcox Group. Major withdrawals from the aquifers were for industrial and public water supplies. A study was conducted by the U.S. Geological Survey in cooperation with the Arkansas Natural Resources Commission and the Arkansas Geological Survey to determine the water level associated with the aquifers in the Cockfield Formation of Claiborne Group and the Wilcox Group in southern and northeastern Arkansas. During February and March 2006, 56 water-level measurements were made in wells completed in the Cockfield aquifer and 59 water-level measurements were made in wells completed in the Wilcox aquifer, 16 in southwestern and 43 in northeastern Arkansas. This report presents the results as potentiometric-surface maps and as long-term water-level hydrographs. The regional direction of ground-water flow in the Cockfield Formation of Claiborne Group generally is towards the east and southeast, away from the outcrop, except in areas of intense ground-water withdrawals, such as western Drew County, southeastern Lincoln County, southwestern Calhoun County, and near Crossett in Ashley County. There are three cones of depression indicated by relatively low water-level altitudes in southeastern Lincoln County, southwestern Calhoun County, and near Crossett in Ashley County. The lowest water-level altitude measured was 44 feet above the National Geodetic Vertical Datum of 1929 in Lincoln County; the highest water-level altitude measured was 346 feet above the National Geodetic Vertical Datum of 1929 in Columbia County at the outcrop area. Hydrographs from 40 wells with historical water levels from 1986 to 2006 were evaluated using linear regression to

  3. Structural Styles of the Wilcox and Frio Growth-Fault Trends in Texas: Constraints on Geopressured Reservoirs

    SciTech Connect

    Ewing, Thomas E.; Anderson, R. G.; Babalola, O.; Hubby, K.; Padilla y Sanchez, R.; Reed, R. S.

    1986-01-01

    The wide variability in structural styles within the growth-faulted, geopressured trends of the Texas Gulf Coast is illustrated by detailed structural maps of selected areas of the Wilcox and Frio growth-fault trends and quantified by statistical analysis of fault compartment geometries. Structural variability is a key determinant of the size of geopressured aquifers in the deep subsurface. Two major structural styles exist in the Wilcox trend. (1) In southeast and Central Texas, the trend consists of continuous, closely spaced faults that have little associated rollover despite moderate expansion of section; the fault plane flattens little with depth. Where the trend crosses the Houston Diapir Province, growth faults are localized by preexisting salt pillows; however, the salt diapirs pierced the growth-faulted horizons after the main phase of faulting, so that salt-related movement deforms the growth faults. (2) By contrast, in South Texas a narrow band of growth faults having high expansion and moderate rollover lies above and downdip of a ridge of deformed, overpressured shale but updip of a deep basin formed by withdrawal of overpressured shale. Large antithetic faults are associated with this band of faults. Frio fault systems generally display greater rollover and wider spacing than do Wilcox fault systems; however, the Frio trend displays distinctive features in each study area. At Sarita in South Texas, shale mobilization produced shale ridges, one of which localized a low-angle growth fault. Shale mobilization at Corpus Christi produced a series of large growth faults, shale-cored domal anticlines, and shale-withdrawal basins, which become younger to the east. At Blessing, major growth faults show little progradation seaward. A major early-formed growth-fault system was deformed by later salt tectonism at Pleasant Bayou. At Port Arthur, low-displacement, long-lived faults formed on a sand-poor shelf margin contemporaneously with broad salt uplifts and

  4. Application of the failure assessment diagram to the evaluation of pressure-temperature limits for a pressurized water reactor

    SciTech Connect

    Yoon, K.K.; Bloom, J.M.; Pavinich, W.A.; Slager, H.W.

    1984-06-01

    The failure assessment diagram approach, an elastic-plastic fracture mechanics procedure based on the J-integral concept, was used in the evaluation of pressure-temperature (P-T) limits for the beltline region of the vessel of a pressurized water reactor. The main objective of this paper is to illustrate the application of an alternate fracture mechanics method for the evaluation of pressure-temperature limits, as allowed by Title 10, Code of Federal Regulation Part 50 (10 CFR 50), Appendix G. The evaluation of P-T limits for the beltline region of a pressurized water reactor vessel was based on the following assumptions: ASME Pressure Vessel and Piping Code, Section III, Appendix G reference flaw End-of-life fluence level in the beltline region Longitudinal flaw in the beltline weld J-resistance material toughness curves obtained from the U.S. Nuclear Regulatory Commission's Heavy Section Steel Technology (HSST) program Other material properties obtained from the Babcock and Wilcox Integrated Reactor Vessel Material Surveillance Program The maximum allowable pressure levels were calculated at 33 time points along the given bulk coolant temperature history representing the normal operation of a pressurized water reactor. The results of the calculations showed that adequate margins of safety on operating pressure for the critical weld in the beltline of the pressurized water reactor vessel are assured.

  5. Potentiometric surface, 2013, and water-level differences, 1991-2013, of the Carrizo-Wilcox aquifer in northwest Louisiana

    USGS Publications Warehouse

    Fendick, Robert B.; Carter, Kayla

    2015-01-01

    This report presents data and maps that illustrate the potentiometric surface of the Carrizo-Wilcox aquifer during March–May 2013 and water-level differences from 1991 to 2013. The potentiometric surface map can be used for determining the direction of groundwater flow, hydraulic gradients, and effects of withdrawals on the groundwater resource. The rate of groundwater movement also can be estimated from the gradient when the hydraulic conductivity is applied. Water-level data collected for this study are stored in the USGS National Water Information System (NWIS) (http://waterdata.usgs.gov/nwis) and are on file at the USGS office in Baton Rouge, La.

  6. Coal gasification systems engineering and analysis. Appendix G: Commercial design and technology evaluation

    NASA Technical Reports Server (NTRS)

    1980-01-01

    A technology evaluation of five coal gasifier systems (Koppers-Totzek, Texaco, Babcock and Wilcox, Lurgi and BGC/Lurgi) and procedures and criteria for evaluating competitive commercial coal gasification designs is presented. The technology evaluation is based upon the plant designs and cost estimates developed by the BDM-Mittelhauser team.

  7. Numerical simulation of PWR response to a small break LOCA (loss-of-coolant accident) with reactor coolant pumps operating

    SciTech Connect

    Adams, J.P.; Dobbe, C.A.; Bayless, P.D.

    1986-01-01

    Calculations have been made of the response of pressurized water reactors (PWRs) during a small-break, loss-of-coolant accident with the reactor coolant pumps (RCPs) operating. This study was conducted, as part of a comprehensive project, to assess the relationship between measurable RCP parameters, such as motor power or current, and fluid density, both local (at the RCP inlet) and global (average reactor coolant system). Additionally, the efficacy of using these RCP parameters, together with fluid temperature, to identify an off-nominal transient as either a LOCA, a heatup transient, or a cooldown transient and to follow recovery from the transient was assessed. The RELAP4 and RELAP5 computer codes were used with three independent sets of RCP, two-phase degradation multipliers. These multipliers were based on data obtained in two-phase flow conditions for the Semiscale, LOFT, and Creare/Combustion Engineering (CE)/Electric Power Research Institute (EPRI) pumps, respectively. Two reference PWRs were used in this study: Zion, a four-loop, 1100-MWe, Westinghouse plant operated by Commonwealth Edison Co. in Zion, Illinois and Bellefonte, a two-by-four loop, 1213 MWe, Babcock and Wilcox designed plant being built by the Tennessee Valley Authority in Scottsboro, Alabama. The results from this study showed that RCP operation resulted in an approximately homogeneous reactor coolant system and that this result was independent of reference plant, computer code, or two-phase RCP head degradation multiplier used in the calculation.

  8. The development and application of k0-standardization method of neutron activation analysis at Es-Salam research reactor

    NASA Astrophysics Data System (ADS)

    Alghem, L.; Ramdhane, M.; Khaled, S.; Akhal, T.

    2006-01-01

    In recent years the k0-NAA method has been applied and developed at the 15 MW Es-Salam research reactor, which includes: (1) the detection efficiency calibration of γ-spectrometer used in k0-NAA, (2) the determination of reactor neutron spectrum parameters such as α and f factors in the irradiation channel, and (3) the validation of the developed k0-NAA procedure by analysing SRM, namely AIEA-Soil7 and CRM, namely IGGE-GSV4. The analysis results obtained by k0-NAA with 27 elements of Soil-7 standard and 14 elements of GSV-4 standard were compared with certified values. The analysis results showed that the deviations between experimental and certified values were mostly less than 10%. The k0-NAA procedure established at Es-Salam research reactor has been regarded as a reliable standardization method of NAA and as available for practical applications.

  9. On the meaning of formative measurement and how it differs from reflective measurement: comment on Howell, Breivik, and Wilcox (2007).

    PubMed

    Bagozzi, Richard P

    2007-06-01

    D. Howell, E. Breivik, and J. B. Wilcox (2007) have presented an important and interesting analysis of formative measurement and have recommended that researchers abandon such an approach in favor of reflective measurement. The author agrees with their recommendations but disagrees with some of the bases for their conclusions. He suggests that although latent variables refer to mental states or mental events that have objective reality, to gain knowledge of the existence of these states or events requires that emphasis be placed on the nature and interpretation of the relationship between latent and manifest variables. This relationship is not a causal one but rather a kind of correspondence rule that contains theoretical, empirical, operational, and logical meanings as part of its content and structure. Implications of the above views are discussed for formative and reflective measurement.

  10. Interview with Professor Mark Wilcox.

    PubMed

    Wilcox, Mark

    2016-08-01

    Mark Wilcox speaks to Georgia Patey, Commissioning Editor: Professor Mark Wilcox is a Consultant Microbiologist and Head of Microbiology at the Leeds Teaching Hospitals (Leeds, UK), the Professor of Medical Microbiology at the University of Leeds (Leeds, UK), and is the Lead on Clostridium difficile and the Head of the UK C. difficile Reference Laboratory for Public Health England (PHE). He was the Director of Infection Prevention (4 years), Infection Control Doctor (8 years) and Clinical Director of Pathology (6 years) at the Leeds Teaching Hospitals. He is Chair of PHE's Rapid Review Panel (reviews utility of infection prevention and control products for National Health Service), Deputy Chair of the UK Department of Health's Antimicrobial Resistance and Healthcare Associated Infection Committee and a member of PHE's HCAI/AR Programme Board. He is a member of UK/European/US working groups on C. difficile infection. He has provided clinical advice as part of the FDA/EMA submissions for the approval of multiple novel antimicrobial agents. He heads a healthcare-associated infection research team at University of Leeds, comprising approximately 30 doctors, scientists and nurses; projects include multiple aspects of C. difficile infection, diagnostics, antimicrobial resistance and the clinical development of new antimicrobial agents. He has authored more than 400 publications, and is the coeditor of Antimicrobial Chemotherapy (5th/6th/7th Editions, 15 December 2007).

  11. Potentiometric surfaces of aquifers in the Cockfield Formation in southeastern Arkansas and the Wilcox Group in southern and northeastern Arkansas, 2000

    USGS Publications Warehouse

    Schrader, Tony P.; Joseph, Robert L.

    2000-01-01

    The Cockfield and lower Wilcox aquifers are sources of water for local use in southern and northeastern Arkansas, where in 1995 more than 51 million gallons per day of water was withdrawn. During January through April 2000, 54 water-level measurements were made in wells completed in the Cockfield aquifer, 13 water-level measurements were made in wells completed in the lower Wilcox aquifer in southern Arkansas, and 43 water-level measurements were made in wells completed in the lower Wilcox aquifer in northeastern Arkansas. The potentiometric surface data reveal spatial trends in both aquifers across the study areas. The regional direction of ground-water flow of the Cockfield aquifer is generally toward the east and south, away from the outcrop area, except in areas of intense ground-water withdrawals. The configuration of the potentiometric surface indicates that heavy pumpage has probably altered or reversed the natural direction of flow in these areas. A potentiometric low caused by the pumpage near Greenville, Mississippi, extends into Chicot, Desha, and Drew Counties. Water levels in five wells showed average declines between 0.5 and 0.8 foot per year. The regional direction of ground-water flow in the lower Wilcox aquifers is generally east and south, away from the outcrop, except in areas of intense ground-water withdrawals. Potentiometric depressions, where flow is toward centers of pumping, indicate that heavy pumpage has probably altered or reversed the natural direction of flow. Two potentiometric depressions are centered in the vicinity of Paragould and West Memphis, Arkansas, where ground-water withdrawals probably have altered the natural direction of flow. Long-term hydrographs of seven wells show water-level declines in the lower Wilcox aquifer in northeastern Arkansas. The average water-level decline in two wells was between 0.8 and 1.0 foot per year and in five wells was between 1.2 and 1.8 foot per year.

  12. Peter Wilcox: A new purple-skin, yellow flesh fresh market potato cultivar

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Peter Wilcox is a new, medium-maturing, purple-skin, yellow-flesh potato cultivar for fresh market. Peter Wilcox also produces light-colored chips, although it is being released primarily as a fresh market potato because of its skin and flesh colors. Tubers of Peter Wilcox are attractive, smooth, wi...

  13. Altitude of the water table in the alluvial and Wilcox aquifers in the vicinity of Richland and Tehuacana creeks and the Trinity River, Texas, December 1979

    USGS Publications Warehouse

    Garza, Sergio

    1980-01-01

    This map shows the altitude of the water table in the alluvial and Wilcox aquifers in the vicinity of Richland and Tehuacana Creeks and the Trinity River, Tex., in December 1979. The water-table contours were constructed on the basis of water-level control derived from an inventory of shallow wells in the area, topographic maps, and field locations of numerous small springs and seeps. (USGS)

  14. Water-quality assessment of the Trinity River basin, Texas : ground-water quality of the Trinity, Carrizo-Wilcox, and Gulf Coast aquifers, February-August 1994

    USGS Publications Warehouse

    Reutter, David C.; Dunn, David D.

    2000-01-01

    Ground-water samples were collected from wells in the outcrops of the Trinity, Carrizo-Wilcox, and Gulf Coast aquifers during February-August 1994 to determine the quality of ground water in the three major aquifers in the Trinity River Basin study unit, Texas. These samples were collected and analyzed for selected properties, nutrients, major inorganic constituents, trace elements, pesticides, dissolved organic carbon, total phenols, methylene blue active substances, and volatile organic compounds as part of the U.S. Geological Survey National Water-Quality Assessment Program. Quality-control practices included the collection and analysis of blank, duplicate, and spiked samples. Samples were collected from 12 shallow wells (150 feet or less) and from 12 deep wells (greater than 150 feet) in the Trinity aquifer, 11 shallow wells and 12 deep wells in the Carrizo-Wilcox aquifer, and 14 shallow wells and 10 deep wells in the Gulf Coast aquifer. The three aquifers had similar water chemistries-calcium was the dominant cation and bicarbonate the dominant anion. Statistical tests relating well depths to concentrations of nutrients and major inorganic constituents indicated correlations between well depth and concentrations of ammonia nitrogen, nitrite plus nitrate nitrogen, bicarbonate, sodium, and dissolved solids in the Carrizo-Wilcox aquifer and between well depth and concentrations of sulfate in the Gulf Coast aquifer. The tests indicated no significant correlations for the Trinity aquifer. Concentrations of dissolved solids were larger than the secondary maximum contaminant level of 500 milligrams per liter established for drinking water by the U.S. Environmental Protection Agency in 12 wells in the Trinity aquifer, 4 wells in the Carrizo-Wilcox aquifer, and 6 wells in the Gulf Coast aquifer. Iron concentrations were larger than the secondary maximum contaminant level of 300 micrograms per liter in at least 3 samples from each aquifer, and manganese concentrations

  15. Potentiometric surface of the Cockfield Aquifer in southeastern Arkansas and the Wilcox Aquifers in southern and northeastern Arkansas, October 1996-July 1997

    USGS Publications Warehouse

    Joseph, Robert L.

    1998-01-01

    The Cockfield and Wilcox aquifers are secondary sources of water for local use in southern and northeastern Arkansas, where in 1995 more than 51 million gallons per day of water was withdrawn. During October 1996 to July 1997, water levels in the Cockfield and Wilcox aquifers were measured in 104 wells in Arkansas. The potentiometric surface data reveal spatial trends in both aquifers across the study areas. The regional direction of ground-water flow of the Cockfield aquifer is generally southeastward, away from the outcrop area, except where affected by intense ground-water withdrawals. The potentiometric surface indicates that heavy pumpage has altered or reversed the natural direction of flow in some areas. Flow in these areas is toward centers of pumping within cones of depression. A cone of depression caused by the pumpage near Greenville, Mississippi, extends into Chicot, Desha, and Drew Counties. This cone of depression has altered flow patternArkansas. Long-term hydrographs of six wells, during the period 1971-1996, showed water levels declined at an average rate between 0.5 and 1.0 foot per year at these locations. The regional direction of ground-water flow in the Wilcox aquifers is generally toward the east and south, away from the outcrop except where water levels are affected by intense ground-water withdrawals. The potentiometric surface indicates that heavy pumpage has altered or reversed the natural direction of ground-water flow in some areas. Flow in these areas is toward centers of pumping within cones of depression. Two cones of depression are centered in the vicinity of Paragould and West Memphis, Arkansas, where ground-water withdrawals have altered the natural direction of flow. Long-term hydrographs of seven wells, during the period 1971- 1996, show water-level declines in the Wilcox aquifer in northeastern Arkansas generally were between 0.5 and 1.0 foot per year but were more than 1.0 foot per year in two wells. The U.S. Geological Survey

  16. 78 FR 73898 - Operator Licensing Examination Standards for Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-12-09

    ... COMMISSION Operator Licensing Examination Standards for Power Reactors AGENCY: Nuclear Regulatory Commission... Standards for Power Reactors.'' DATES: Submit comments by February 7, 2014. Comments received after this... of New Reactors; or Timothy Kolb, Office of Nuclear Reactor Regulation, U.S. Nuclear...

  17. Study of Channel Morphology and Infill Lithology in the Wilcox Group Central Louisiana Using Seismic Attribute Analysis

    NASA Astrophysics Data System (ADS)

    Chen, Feng

    The fluvial and deltaic Wilcox Group is a major target for hydrocarbon and coal exploration in northern and central Louisiana. However, the characterization and delineation of fluvial systems is a difficult task due to the variability and complexity of fluvial systems and their internal heterogeneities. Seismic geomorphology is studied by recognizing paleogeographic features in seismic stratal slices, which are seismic images of paleo-depositional surfaces. Seismic attributes, which are extracted along seismic stratal slices, can reveal information that is not readily apparent in raw seismic data. The existence and distribution of fluvial channels are recognized by the channel geomorphology in seismic attributes displayed on stratal slices. The lithologies in the channels are indicated by those seismic attributes that are directly related to the physical properties of rocks. Selected attributes utilized herein include similarity, spectral decomposition, sweetness, relative acoustic impedance, root mean square (RMS) amplitude, and curvature. Co-rendering and Red/Green/Blue (RGB) display techniques are also included to better illuminate the channel geometry and lithology distribution. Hydrocarbons may exist in the channel sand-bodies, but are not explicitly identified herein. Future drilling plans for oil and gas exploration may benefit from the identification of the channels and the lithologies that fill them.

  18. The "Plant Drosophila": E.B. Babcock, the genus "Crepis," and the evolution of a genetics research program at Berkeley, 1915-1947.

    PubMed

    Smocovitis, Vassiliki Betty

    2009-01-01

    This paper explores the research and administrative efforts of Ernest Brown Babcock, head of the Division of Genetics in the College of Agriculture at the University of California, Berkeley, the first academic unit so named in the United States. It explores the rationale for his choice of "model organism," the development--and transformation--of his ambitious genetics research program centering on the weedy plant genus named "Crepis" (commonly known as the hawkbeard), along with examining in detail the historical development of the understanding of genetic mechanisms of evolutionary change in plants leading to the period of the evolutionary synthesis. Chosen initially as the plant counterpart of Thomas Hunt Morgan's "Drosophila melanogaster," the genus "Crepis" instead came to serve as the counterpart of Theodosius Dobzhansky's "Drosophila pseudoobscura," leading the way in plant evolutionary genetics, and eventually providing the first comprehensive systematic treatise of any genus that was part of the movement known as biosystematics, or the "new" systematics. The paper also suggests a historical rethinking of the application of the terms model organism, research program, and experimental system in the history of biology.

  19. FABRICATION PROCESS AND PRODUCT QUALITY IMPROVEMENTS IN ADVANCED GAS REACTOR UCO KERNELS

    SciTech Connect

    Charles M Barnes

    2008-09-01

    A major element of the Advanced Gas Reactor (AGR) program is developing fuel fabrication processes to produce high quality uranium-containing kernels, TRISO-coated particles and fuel compacts needed for planned irradiation tests. The goals of the AGR program also include developing the fabrication technology to mass produce this fuel at low cost. Kernels for the first AGR test (“AGR-1) consisted of uranium oxycarbide (UCO) microspheres that werre produced by an internal gelation process followed by high temperature steps tot convert the UO3 + C “green” microspheres to first UO2 + C and then UO2 + UCx. The high temperature steps also densified the kernels. Babcock and Wilcox (B&W) fabricated UCO kernels for the AGR-1 irradiation experiment, which went into the Advance Test Reactor (ATR) at Idaho National Laboratory in December 2006. An evaluation of the kernel process following AGR-1 kernel production led to several recommendations to improve the fabrication process. These recommendations included testing alternative methods of dispersing carbon during broth preparation, evaluating the method of broth mixing, optimizing the broth chemistry, optimizing sintering conditions, and demonstrating fabrication of larger diameter UCO kernels needed for the second AGR irradiation test. Based on these recommendations and requirements, a test program was defined and performed. Certain portions of the test program were performed by Oak Ridge National Laboratory (ORNL), while tests at larger scale were performed by B&W. The tests at B&W have demonstrated improvements in both kernel properties and process operation. Changes in the form of carbon black used and the method of mixing the carbon prior to forming kernels led to improvements in the phase distribution in the sintered kernels, greater consistency in kernel properties, a reduction in forming run time, and simplifications to the forming process. Process parameter variation tests in both forming and sintering steps led

  20. CONVECTIVE BABCOCK-LEIGHTON DYNAMO MODELS

    SciTech Connect

    Miesch, Mark S.; Brown, Benjamin P.

    2012-02-20

    We present the first global, three-dimensional simulations of solar/stellar convection that take into account the influence of magnetic flux emergence by means of the Babcock-Leighton (BL) mechanism. We have shown that the inclusion of a BL poloidal source term in a convection simulation can promote cyclic activity in an otherwise steady dynamo. Some cycle properties are reminiscent of solar observations, such as the equatorward propagation of toroidal flux near the base of the convection zone. However, the cycle period in this young sun (rotating three times faster than the solar rate) is very short ({approx}6 months) and it is unclear whether much longer cycles may be achieved within this modeling framework, given the high efficiency of field generation and transport by the convection. Even so, the incorporation of mean-field parameterizations in three-dimensional convection simulations to account for elusive processes such as flux emergence may well prove useful in the future modeling of solar and stellar activity cycles.

  1. Hot Isostatic Press Can Optimization for Aluminum Cladding of U-10Mo Reactor Fuel Plates: FY12 Final Report and FY13 Update

    SciTech Connect

    Clarke, Kester D.; Crapps, Justin M.; Scott, Jeffrey E.; Aikin, Beverly; Vargas, Victor D.; Dvornak, Matthew J.; Duffield, Andrew N.; Weinberg, Richard Y.; Alexander, David J.; Montalvo, Joel D.; Hudson, Richard W.; Mihaila, Bogdan; Liu, Cheng; Lovato, Manuel L.; Dombrowski, David E.

    2013-08-26

    Currently, the proposed processing path for low enriched uranium – 10 wt. pct. molybdenum alloy (LEU-10Mo) monolithic fuel plates for high power research and test reactors includes hot isostatic pressing (HIP) to bond the aluminum cladding that encapsulates the fuel foil. Initial HIP experiments were performed at Idaho National Laboratory (INL) on approximately ¼ scale “mini” fuel plate samples using a HIP can design intended for these smaller experimental trials. These experiments showed that, with the addition of a co-rolled zirconium diffusion barrier on the LEU-10Mo alloy fuel foil, the HIP bonding process is a viable method for producing monolithic fuel plates. Further experimental trials at Los Alamos National Laboratory (LANL) effectively scaled-up the “mini” can design to produce full-size fuel prototypic plates. This report summarizes current efforts at LANL to produce a HIP can design that is further optimized for higher volume production runs. The production-optimized HIP can design goals were determined by LANL and Babcock & Wilcox (B&W) to include maintaining or improving the quality of the fuel plates produced with the baseline scaled-up mini can design, while minimizing material usage, improving dimensional stability, easing assembly and disassembly, eliminating machining, and significantly reducing welding. The initial small-scale experiments described in this report show that a formed-can approach can achieve the goals described above. Future work includes scaling the formed-can approach to full-size fuel plates, and current progress toward this goal is also summarized here.

  2. Trap types vs productivity of significant Wilcox gas fields in the south Texas, listric growth fault trend, and the divergent origin of its two largest producers

    SciTech Connect

    Stricklin, F.L. Jr.

    1996-09-01

    Detailed mapping and analysis of 23 Wilcox fields in the subject trend indicates that gas production is related to trap type. Of total cumulative production of 3.4 TCFG, 65% is from upthrown fault blocks implying very effective fault seals due to differential pressure and/or shale smears. NE Thompsonville and Bob West fields have produced 650 and 200 BCFG, respectively, with 400 BCFG remaining reserves in the latter. The field structures are not attributed to listric growth faulting, as is suggested by their trend location. NE Thompsonville is a 9-mile-long turtle structure that originated through depositional loading of an upper slope basin, followed by tilting, and then eventual collapse of a sediment squeeze-up mound due to gravitational instability. These events provide an excellent example of basin evolution through sediment loading accompanied by withdrawal of a salt-shale substrate; the basin flanks are defined by basin-dipping listric faulting that accommodated subsidence and merge beneath its floor. Bob West Field lies along the edge of the Laramide fold belt. The 1-1/2 x 4 mile field anticline adjoins a deep-seated fault that slices over and across a buried structural ridge of probable Cretaceous age. Uplift of the latter, immediately following deposition of 20+ stacked, shelf-bar producing sands, upwarped the fault and resulted in rollover growth of the Wilcox anticline. The fault shows no downward decrease in dip typical of listric faults. NE Thompsonville and Bob West fields both produce upthrown along crestal faults. This analysis indicates that {open_quotes}high-side{close_quotes} closures, irrespective of diverse origins, have achieved head-of-the-class stature as Wilcox gas producers.

  3. A proposed standard on medical isotope production in fission reactors

    SciTech Connect

    Schenter, R. E.; Brown, G. J.; Holden, C. S.

    2006-07-01

    Authors Robert E. Sehenter, Garry Brown and Charles S. Holden argue that a Standard for 'Medical Isotope Production' is needed. Medical isotopes are becoming major components of application for the diagnosis and treatment of all the major diseases including all forms of cancer, heart disease, arthritis, Alzheimer's, among others. Current nuclear data to perform calculations is incomplete, dated or imprecise or otherwise flawed for many isotopes that could have significant applications in medicine. Improved data files will assist computational analyses to design means and methods for improved isotope production techniques in the fission reactor systems. Initial focus of the Standard is expected to be on neutron cross section and branching data for both fast and thermal reactor systems. Evaluated and reviewed tables giving thermal capture cross sections and resonance integrals for the major target and product medical isotopes would be the expected 'first start' for the 'Standard Working Group'. (authors)

  4. A Numerical Model of Deuterium and Oxygen-18 Diffusion in the Confined Lower Wilcox Aquifer of the Lower Mississippi Valley (USA)

    NASA Astrophysics Data System (ADS)

    Currens, B. J.; Sawyer, A. H.; Fryar, A. E.; Parris, T. M.; Zhu, J.

    2015-12-01

    Deuterium and oxygen-18 are routinely used with noble gases and radioisotopes (e.g., 2H, 14C, 36Cl) to infer climate during groundwater recharge. However, diffusion of 2H and 18O between a confined aquifer and bounding aquitards could alter total isotope concentrations and the inferred temperature during recharge if groundwater flow is sufficiently slow. Hendry and Schwartz (WRR 24(10), 1988) explained anomalous 2H and 18O enrichment in the Milk River aquifer of Alberta by analytically modeling isotope diffusion between the lower bounding aquitard and the aquifer. Haile (PhD dissertation, U. Kentucky, 2011) inferred the same mechanism to explain 2H and 18O enrichment along a flowpath in the confined Lower Wilcox aquifer of the northern Gulf Coastal Plain in Missouri and Arkansas. Based on the geologic and hydraulic properties of the Lower Wilcox aquifer, a numerical model has been constructed to determine how diffusion may influence 2H and 18O concentrations in regional aquifers with residence times on the order of 104 to 105 years. The model combines solutions for a 1D forward-in-time, finite-difference groundwater flow equation with an explicit-implicit Crank-Nicholson algorithm for advection and diffusion to solve for flow velocity and isotope concentration. Initial results are consistent with the analytical solution of Hendry and Schwartz (1988), indicating diffusion as a means of isotopic enrichment along regional groundwater flowpaths.

  5. Coal gasification systems engineering and analysis. Appendix D: Cost and economic studies

    NASA Technical Reports Server (NTRS)

    1980-01-01

    The detailed cost estimate documentation for the designs prepared in this study are presented. The include: (1) Koppers-Totzek, (2) Texaco (3) Babcock and Wilcox, (4) BGC-Lurgi, and (5) Lurgi. The alternate product cost estimates include: (1) Koppers-Totzek and Texaco single product facilities (methane, methanol, gasoline, hydrogen), (2) Kopers-Totzek SNG and MBG, (3) Kopers-Totzek and Texaco SNG and MBG, and (4) Lurgi-methane and Lurgi-methane and methanol.

  6. Properties and chemical constituents in ground water from the lower Wilcox Aquifer, Mississippi Embayment Aquifer System, south-central United States

    USGS Publications Warehouse

    Pettijohn, Robert A.; Busby, John F.; Beckman, Jeffery D.

    1993-01-01

    The Gulf Coast Regional Aquifer-System Analysis is a study of regional aquifers composed of sediments of mostly Cenozoic age that underlie about 230,000 sq mi of the Gulf Coastal Plain. These regional aquifers are part of three aquifer systems: (1) the Mississippi Embayment Aquifer System, (2) the Texas Coastal Uplands Aquifer System, and (3) the Coastal Lowlands Aquifer System. The water chemistry of the Lower Wilcox Aquifer, which is part of the Mississippi Embayment Aquifer System is presented by a series of maps. These maps show the areal distribution of (1) the concentration of dissolved solids and temperature, (2) the primary water types and pH, (3) the concentration of major ions and silica, and (4) the milliequivalent ratios of selected ions. Dissolved constituents, pH, temperature, and ratios are based on the median values of all samples in each 100-sq-mi area. The concentration of dissolved solids in water from the Lower Wilcox Aquifer ranges from 18 mg/L near the outcrop in western Tennessee to 122,000 mg/L in a down-dip area in southern Mississippi. The primary water type is calcium bicarbonate in the outcrop area and sodium bicarbonate in all other areas of the aquifer within the limits of available data. The concentrations of major ions generally increase from the outcrop area to the down-dip limit of the data in the southern part of the aquifer area east of the Mississippi River. The milliequivalent ratio maps of selected ions in water from the Lower Wilcox Aquifer indicate some trends. The milliequivalent ratio of magnesium plus calcium to bicarbonate ranges from less than 0.1 to 40.4 and generally decreases from outcrop to down-dip limit of the data in the southern part of the aquifer area east of the Mississippi River. The milliequivalent ratio of bicarbonate to chloride ranges from 0.01 in southern Mississippi to 52.3 in northwestern Mississippi. This ratio increases from the outcrop toward the Mississippi River and from north to south in the

  7. Potentiometric map of the Meridian-Upper Wilcox Aquifer in Mississippi, fall 1983

    USGS Publications Warehouse

    Darden, Daphne

    1986-01-01

    The Meridian-upper Wilcox aquifer consists of the Meridian Sand member of the Tallahatta Formation and the uppermost sand beds of the Wilcox Group. Thickness of the aquifer ranges from 50 ft to 500 ft. Precipitation recharges the Meridian-upper Wilcox in the outcrop area, which extends from Benton County, MS, in the north to Clarke County, MS, in the south. The potentiometric map is based on water level measurements made in about 170 wells in the Meridian-upper Wilcox aquifer in October 1983, and on the approximate altitudes of water surfaces in some major streams. The contours show altitudes at which water levels would have stood in tightly cased unpumped wells. This map, the second in the series for the Meridian-upper Wilcox aquifer, updates a map that delineated the potentiometric surface of the aquifer in 1979. (Lantz-PTT)

  8. Non-Power Reactor Operator Licensing Examiner Standards. Revision 1

    SciTech Connect

    1995-06-01

    The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, these standards will be revised periodically to accommodate comments and reflect new information or experience.

  9. Perspective view of Wilcox Building (7 North E Street), with ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    Perspective view of Wilcox Building (7 North E Street), with Eli Cafe (7 North E Street), the Palace Saloon (11 North E Street), and Fetsche's (15 North E Street) to left of frame, view looking north on E Street - Lakeview Downtown Historic District, E, F & G Streets between Second Street North & First Street South, Lakeview, Lake County, OR

  10. Azalea's Worst Nightmare: The Strawberry Rootworm, Paria fargariae Wilcox

    Technology Transfer Automated Retrieval System (TEKTRAN)

    The strawberry rootworm (SRW), Paria fargariae Wilcox, is an emergent pest of azaleas in commercial production nurseries in the southeastern US. Larvae feed on roots but do minimal damage. Adults feed at night and make small holes in the foliage. Severe damage has been reported in many nurseries, es...

  11. Benson R. Wilcox--industry, genius, judgment.

    PubMed

    Murray, Gordon F

    2010-09-01

    The 29th President of the Society of Thoracic Surgeons, Benson R. Wilcox, MD, died at his home in Fearington Village Center, NC, on May 11, 2010. In 2 weeks, he would have been 78 years old, and the focus of a birthday celebration for his wife, Patsy Davis, his 4 children, Adelaide, Sandra, Melissa, and Reid, 11 grandchildren, and many loved ones. With his death, caused by brain cancer, the University of North Carolina lost one of its most prolific and loyal sons, and the profession of thoracic surgery, one of its wisest leaders. Two facts are certain: Ben will be remembered forever by his students, residents, and colleagues, and our sky is, indeed, Carolina blue.

  12. Development of ground water from the Carrizo sand and Wilcox group in Dimmit, Zavala, Maverick, Frio, Atacosa, Median, Bexar, Live Oak, McMullen, La Salle, and Webb Counties, Texas

    USGS Publications Warehouse

    Moulder, E.A.

    1957-01-01

    The development of ground water for irrigation from the Carrizo sand south and southwest of San Antonio, Tex., has increased rapidly during the past few years. Declining pumping water levels in irrigation wells, caused by increased withdrawals, have caused considerable concern among the residents of the area. In response, the Nueces River Conservation and Reclamation District entered into a cooperative agreement with the Texas Board of Water Engineers and the United States Geological Survey to determine the extent of development and the rate of withdrawal that has cause the decline. All wells that discharged more than 150 gallons per minute for extended periods of time in 1955 from either the Carrizo sand or sands of the Wilcox group were studied and are shown on [late 1. Estimates were made of the total withdrawals by county and are given in table 2. Similar estimates of withdrawals in some of the counties for the irrigation years 1929-30, 1938-39, 1944-45, and 1947-48 are presented for comparison in table 3. Although the Carrizo sand is the principal source of ground water pumped in the area, estimate of withdrawals of water from the Wilcox were included in this inventory because (1) the formation appears to be hydraulically connected to the Carrizo sand, (2) the quality of water generally is good in the outcrop area of the Wilcox, and (3) appreciable withdrawals are being made from the Wilcox for irrigation in a few areas. The investigation covered an area of about 7,500 square miles and included all or parts of the following counties: Dimmit, Zavala, Maverick, Frio, Atascosa, Medina, Bexar, Live Oak, McMullen, La Salle, and Webb (fig. 1).

  13. Validation of standardized computer analyses for licensing evaluation/TRITON two-dimensional and three-dimensional models for light water reactor fuel

    SciTech Connect

    Bowman, S. M.; Gill, D. F.

    2006-07-01

    The isotopic depletion capabilities of the new Standardized Computer Analyses for Licensing Evaluation control module TRITON, coupled with ORIGEN-S, were evaluated using spent fuel assays from several commercial light water reactors with both standard and mixed-oxide fuel assemblies. Calculations were performed using the functional modules NEWT and KENO-VI. NEWT is a two-dimensional, arbitrary-geometry, discrete-ordinates transport code, and KENO-VI is a three-dimensional Monte Carlo transport code capable of handling complex three-dimensional geometries. To validate the codes and data used in depletion calculations, numerical predictions were compared with experimental measurements for a total of 29 samples taken from the Calvert Cliffs, Obrigheim, and San Onofre pressurized water reactors and the Gundremmingen boiling water reactor. Similar comparisons have previously been performed at the Oak Ridge National Laboratory for the one-dimensional SAS2H control module. The SAS2H, TRITON/KENO-VI, and TRITON/NEWT results were compared for corresponding samples. All analyses showed that TRITON/KENO-VI and TRITON/NEWT produced typically similar or better results than SAS2H. The calculations performed in this validation study demonstrate that the depletion capabilities of TRITON accurately model spent fuel depletion and decay. (authors)

  14. SPACE-R Thermionic Space Nuclear Power System: Design and Technology Demonstration Program. Semiannual technical progress report for period ending March 1993

    SciTech Connect

    Not Available

    1993-05-01

    This Semiannual Technical Progress Report summarizes the technical progress and accomplishments for the Thermionic Space Nuclear Power System (TI-SNPS) Design and Technology Demonstration Program of the Prime Contractor, Space Power Incorporated (SPI), its subcontractors and supporting National Laboratories during the first half of the Government Fiscal Year (GFY) 1993. SPI`s subcontractors and supporting National Laboratories include: Babcock & Wilcox for the reactor core and externals; Space Systems/Loral for the spacecraft integration; Thermocore for the radiator heat pipes and the heat exchanger; INERTEK of CIS for the TFE, core elements and nuclear tests; Argonne National Laboratories for nuclear safety, physics and control verification; and Oak Ridge National laboratories for materials testing. Parametric trade studies are near completion. However, technical input from INERTEK has yet to be provided to determine some of the baseline design configurations. The INERTEK subcontract is expected to be initiated soon. The Point Design task has been initiated. The thermionic fuel element (TFE) is undergoing several design iterations. The reactor core vessel analysis and design has also been started.

  15. Rolling Process Modeling Report. Finite-Element Model Validation and Parametric Study on various Rolling Process parameters

    SciTech Connect

    Soulami, Ayoub; Lavender, Curt A.; Paxton, Dean M.; Burkes, Douglas

    2015-06-15

    Pacific Northwest National Laboratory (PNNL) has been investigating manufacturing processes for the uranium-10% molybdenum alloy plate-type fuel for high-performance research reactors in the United States. This work supports the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization Reactor Conversion Program. This report documents modeling results of PNNL’s efforts to perform finite-element simulations to predict roll-separating forces for various rolling mill geometries for PNNL, Babcock & Wilcox Co., Y-12 National Security Complex, Los Alamos National Laboratory, and Idaho National Laboratory. The model developed and presented in a previous report has been subjected to further validation study using new sets of experimental data generated from a rolling mill at PNNL. Simulation results of both hot rolling and cold rolling of uranium-10% molybdenum coupons have been compared with experimental results. The model was used to predict roll-separating forces at different temperatures and reductions for five rolling mills within the National Nuclear Security Administration Fuel Fabrication Capability project. This report also presents initial results of a finite-element model microstructure-based approach to study the surface roughness at the interface between zirconium and uranium-10% molybdenum.

  16. A concept of the innovative nuclear technology based on standardized fast reactors SVBR-75/100 with lead-bismuth coolant for modular nuclear power plants of different capacity and purpose

    SciTech Connect

    Zrodnikov, A.V.; Toshinsky, G.I.; Komlev, O.G.; Dragunov, Yu.G.; Stepanov, V.S.; Generalov, V.N.; Krushelnitsky, V.N.

    2007-07-01

    Today's nuclear power is in the state of an intrinsic conflict between economic and safety requirements. This fact makes difficult its sustainable development. One of the ways of finding the solution to the problem can be use of modular fast reactors SVBR-75/100 cooled by lead-bismuth coolant that has been mastered in conditions of operating reactors of Russian nuclear submarines. The inherent self-protection and passive safety properties are peculiar to that reactor due to physical features of small power fast reactors ({approx} 100 MWe), chemical inertness and high boiling point of lead-bismuth coolant, integral design of the pool type primary circuit equipment. Due to small power of the reactor, it is possible to fabricate the whole reactor at the factory and to deliver it to the NPP site in practical readiness by using any kind of transport including the railway. Substantiation of the high level of reactor safety, adaptability of the SVBR-75/100 reactor relative to the fuel type and fuel cycle, issues of non-proliferation of nuclear fissile materials, opportunities of multi-purpose usage of the standard SVBR-75/100 reactors have been viewed in the paper. (authors)

  17. A THREE-DIMENSIONAL BABCOCK-LEIGHTON SOLAR DYNAMO MODEL

    SciTech Connect

    Miesch, Mark S.; Dikpati, Mausumi

    2014-04-10

    We present a three-dimensional (3D) kinematic solar dynamo model in which poloidal field is generated by the emergence and dispersal of tilted sunspot pairs (more generally bipolar magnetic regions, or BMRs). The axisymmetric component of this model functions similarly to previous 2.5 dimensional (2.5D, axisymmetric) Babcock-Leighton (BL) dynamo models that employ a double-ring prescription for poloidal field generation but we generalize this prescription into a 3D flux emergence algorithm that places BMRs on the surface in response to the dynamo-generated toroidal field. In this way, the model can be regarded as a unification of BL dynamo models (2.5D in radius/latitude) and surface flux transport models (2.5D in latitude/longitude) into a more self-consistent framework that builds on the successes of each while capturing the full 3D structure of the evolving magnetic field. The model reproduces some basic features of the solar cycle including an 11 yr periodicity, equatorward migration of toroidal flux in the deep convection zone, and poleward propagation of poloidal flux at the surface. The poleward-propagating surface flux originates as trailing flux in BMRs, migrates poleward in multiple non-axisymmetric streams (made axisymmetric by differential rotation and turbulent diffusion), and eventually reverses the polar field, thus sustaining the dynamo. In this Letter we briefly describe the model, initial results, and future plans.

  18. A Three-dimensional Babcock-Leighton Solar Dynamo Model

    NASA Astrophysics Data System (ADS)

    Miesch, Mark S.; Dikpati, Mausumi

    2014-04-01

    We present a three-dimensional (3D) kinematic solar dynamo model in which poloidal field is generated by the emergence and dispersal of tilted sunspot pairs (more generally bipolar magnetic regions, or BMRs). The axisymmetric component of this model functions similarly to previous 2.5 dimensional (2.5D, axisymmetric) Babcock-Leighton (BL) dynamo models that employ a double-ring prescription for poloidal field generation but we generalize this prescription into a 3D flux emergence algorithm that places BMRs on the surface in response to the dynamo-generated toroidal field. In this way, the model can be regarded as a unification of BL dynamo models (2.5D in radius/latitude) and surface flux transport models (2.5D in latitude/longitude) into a more self-consistent framework that builds on the successes of each while capturing the full 3D structure of the evolving magnetic field. The model reproduces some basic features of the solar cycle including an 11 yr periodicity, equatorward migration of toroidal flux in the deep convection zone, and poleward propagation of poloidal flux at the surface. The poleward-propagating surface flux originates as trailing flux in BMRs, migrates poleward in multiple non-axisymmetric streams (made axisymmetric by differential rotation and turbulent diffusion), and eventually reverses the polar field, thus sustaining the dynamo. In this Letter we briefly describe the model, initial results, and future plans.

  19. Fate of injected CO2 in the Wilcox Group, Louisiana, Gulf Coast Basin: Chemical and isotopic tracers of microbial-brine-rock-CO2 interactions

    USGS Publications Warehouse

    Shelton, Jenna L.; McIntosh, Jennifer C.; Warwick, Peter D.; Lee Zhi Yi, Amelia

    2016-01-01

    The “2800’ sandstone” of the Olla oil field is an oil and gas-producing reservoir in a coal-bearing interval of the Paleocene–Eocene Wilcox Group in north-central Louisiana, USA. In the 1980s, this producing unit was flooded with CO2 in an enhanced oil recovery (EOR) project, leaving ∼30% of the injected CO2 in the 2800’ sandstone post-injection. This study utilizes isotopic and geochemical tracers from co-produced natural gas, oil and brine to determine the fate of the injected CO2, including the possibility of enhanced microbial conversion of CO2 to CH4 via methanogenesis. Stable carbon isotopes of CO2, CH4 and DIC, together with mol% CO2 show that 4 out of 17 wells sampled in the 2800’ sandstone are still producing injected CO2. The dominant fate of the injected CO2appears to be dissolution in formation fluids and gas-phase trapping. There is some isotopic and geochemical evidence for enhanced microbial methanogenesis in 2 samples; however, the CO2 spread unevenly throughout the reservoir, and thus cannot explain the elevated indicators for methanogenesis observed across the entire field. Vertical migration out of the target 2800’ sandstone reservoir is also apparent in 3 samples located stratigraphically above the target sand. Reservoirs comparable to the 2800’ sandstone, located along a 90-km transect, were also sampled to investigate regional trends in gas composition, brine chemistry and microbial activity. Microbial methane, likely sourced from biodegradation of organic substrates within the formation, was found in all oil fields sampled, while indicators of methanogenesis (e.g. high alkalinity, δ13C-CO2 and δ13C-DIC values) and oxidation of propane were greatest in the Olla Field, likely due to its more ideal environmental conditions (i.e. suitable range of pH, temperature, salinity, sulfate and iron concentrations).

  20. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-16

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of... GE Hitachi Nuclear Energy (GEH) for the economic simplified boiling water reactor (ESBWR)...

  1. Organic geochemistry and petrology of subsurface Paleocene-Eocene Wilcox and Claiborne Group coal beds, Zavala County, Maverick Basin, Texas, USA

    USGS Publications Warehouse

    Hackley, Paul C.; Warwick, Peter; Hook, Robert W.; Alimi, Hossein; Mastalerz, Maria; Swanson, Sharon M.

    2012-01-01

    Coal samples from a coalbed methane exploration well in northern Zavala County, Maverick Basin, Texas, were characterized through an integrated analytical program. The well was drilled in February, 2006 and shut in after coal core desorption indicated negligible gas content. Cuttings samples from two levels in the Eocene Claiborne Group were evaluated by way of petrographic techniques and Rock–Eval pyrolysis. Core samples from the Paleocene–Eocene Indio Formation (Wilcox Group) were characterized via proximate–ultimate analysis in addition to petrography and pyrolysis. Two Indio Formation coal samples were selected for detailed evaluation via gas chromatography, and Fourier transform infrared (FTIR) and 13C CPMAS NMR spectroscopy. Samples are subbituminous rank as determined from multiple thermal maturity parameters. Elevated rank (relative to similar age coal beds elsewhere in the Gulf Coast Basin) in the study area is interpreted to be a result of stratigraphic and/or structural thickening related to Laramide compression and construction of the Sierra Madre Oriental to the southwest. Vitrinite reflectance data, along with extant data, suggest the presence of an erosional unconformity or change in regional heat flow between the Cretaceous and Tertiary sections and erosion of up to >5 km over the Cretaceous. The presence of liptinite-rich coals in the Claiborne at the well site may indicate moderately persistent or recurring coal-forming paleoenvironments, interpreted as perennially submerged peat in shallow ephemeral lakes with herbaceous and/or flotant vegetation. However, significant continuity of individual Eocene coal beds in the subsurface is not suggested. Indio Formation coal samples contain abundant telovitrinite interpreted to be preserved from arborescent, above-ground woody vegetation that developed during the middle portion of mire development in forested swamps. Other petrographic criteria suggest enhanced biological, chemical and physical

  2. NEUTRONIC REACTOR CONSTRUCTION AND OPERATION

    DOEpatents

    West, J.M.; Weills, J.T.

    1960-03-15

    A method is given for operating a nuclear reactor having a negative coefficient of reactivity to compensate for the change in reactor reactivity due to the burn-up of the xenon peak following start-up of the reactor. When it is desired to start up the reactor within less than 72 hours after shutdown, the temperature of the reactor is lowered prior to start-up, and then gradually raised after start-up.

  3. REACTOR AND NOVEL METHOD

    DOEpatents

    Young, G.J.; Ohlinger, L.A.

    1958-06-24

    A nuclear reactor of the type which uses a liquid fuel and a method of controlling such a reactor are described. The reactor is comprised essentially of a tank for containing the liquid fuel such as a slurry of discrete particles of fissionnble material suspended in a heavy water moderator, and a control means in the form of a disc of neutron absorbirg material disposed below the top surface of the slurry and parallel thereto. The diameter of the disc is slightly smaller than the diameter of the tank and the disc is perforated to permit a flow of the slurry therethrough. The function of the disc is to divide the body of slurry into two separate portions, the lower portion being of a critical size to sustain a nuclear chain reaction and the upper portion between the top surface of the slurry and the top surface of the disc being of a non-critical size. The method of operation is to raise the disc in the reactor until the lower portion of the slurry has reached a critical size when it is desired to initiate the reaction, and to lower the disc in the reactor to reduce the size of the lower active portion the slurry to below criticality when it is desired to stop the reaction.

  4. Occurrence of a Pseudophragmina (Proporocyclina) Zaragosensis bank in upper Wilcox (lower Eocene), Pointe Coupee Parish, Louisiana

    SciTech Connect

    Nunn, L.L.; Lemoine, R.C.

    1987-09-01

    Rock-forming buildups of larger foraminifera, though not unknown in Louisiana, are rare, and are less common in the Gulf Coast than in Eocene strata from the Tethyan and Caribbean regions. In strata from the upper Wilcox Group (lower Eocene) in Pointe Coupee Parish, Louisiana, they have identified a thin, very rich faunal deposit, or bank made up almost exclusively of the tests of the larger foraminifera species, Pseudophragmina (Proporocyclina) zaragosensis. This is the first report of P. (P.) zaragosensis from the Louisiana Wilcox Group. The bank fauna colonized the sand sea floor on the continental shelf, flourished briefly, and then gradually declined, probably in response to increasing water depth, caused either by compaction-induced subsidence and/or a local or regional rise in relative sea level. This bank indicates that, during late Wilcox deposition, warm, shallow continental-shelf conditions prevailed in this region.

  5. Potentiometric map of the Meridian-Upper Wilcox Aquifer in Mississippi, fall 1979

    USGS Publications Warehouse

    Wasson, B.E.

    1980-01-01

    The potentiometric map of the Meridian-upper Wilcox aquifer is one of a series of maps, prepared by the U.S. Geological Survey in cooperation with the Mississippi Department of Natural Resources, Bureau of Land and Water Resources, delineating the potentiometric surfaces of the major aquifers in Mississippi. In the outcrop area of the Meridian-upper Wilcox aquifer the potentiometric surface is strongly affected by recharge from precipitation, by topography, and by drainage of the aquifer by streams. The potentiometric surface slopes downward generally to the west away from the area of outcrop and is strongly affected by large groundwater withdrawals in the Greenwood, Indianola, Marks, Grenada, and Memphis areas. Historically, water levels in or near the outcrop of the Meridian-upper Wilcox have shown little or no long-term changes, but during the past 20 years water levels have declined from 1 to 2 feet per year in much of the confined part of the aquifer. (USGS)

  6. Depth maps from seismic velocities help Wilcox exploration

    SciTech Connect

    Guzman, C.E.; Ramaswamy, M.; Wright, B.K.; Lawler, K.P.

    1996-10-28

    Depth maps generated from a combination of time maps and maximum coherency seismic (MCS) velocities have proven useful in the Tertiary Wilcox play in Louisiana and Texas. Applied aggressively in conjunction with new processing techniques, depth maps generated in this fashion can identify prospects not visible on time maps. Since depth conversion depends on velocities, the method relies on precision velocity measurements derived from an event-oriented algorithm. The velocities are carefully analyzed for misties and then contoured within geologic boundaries. Well information, if any, is also incorporated into the velocity map, and then the time map is converted to a depth map. Two case studies presented here calibrate the method. Other examples show how the technique helped locate a drill-site and predict geologic horizon depth and how it identified a structure not visible on time maps. The paper discusses Wilcox characteristics, the depth-conversion method, constant-velocity calibration, varying-velocity calibration, two applications, and coarse lithology prediction.

  7. Nuclear Reactors and Technology

    SciTech Connect

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

    1992-01-01

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

  8. Converting the Audience: A Conversation with Agnes Wilcox

    ERIC Educational Resources Information Center

    Becker, Becky

    2006-01-01

    This article presents a conversation with Agnes Wilcox, Executive Director of Prison Performing Arts in St. Louis, Missouri, about Prison Performing Arts. Although the average person might balk at the notion of interacting with prison inmates, finding it intimidating, worrisome, or self-sacrificial, for Wilcox, Prison Performing Arts is a…

  9. Examination of fast-reactor fuels and FBR analytical quality-assurance standards and methods. Progress report, October 1-December 31, 1980

    SciTech Connect

    Maraman, W.J.

    1981-03-01

    This project is directed toward the examination and comparison of the effects of neutron irradiation on Liquid Metal Fast Breeder Reactor (LMFBR) Program fuel materials. Unirradiated and irradiated materials will be examined as requested by the Reference Fuels System Branch of the Division of Reactor Research and Technology (DRRT). Capabilities have been established and are being expanded for providing conventional preirradiation and postirradiation examinations. Nondestructive tests will be conducted in a hot-cell facility specifically modified for examining irradiated prototype fuel pins at a rate commensurate with schedules established by DRRT.

  10. CFB boiler for Southern Illinois University: Planning and design

    SciTech Connect

    Silvey, M.; Roth, N.; Haake, A.

    1995-12-31

    Southern Illinois University (SIU) is in the process of installing a Babcock and Wilcox (B and W) coal fired circulating fluidized bed (CFB) boiler at its Carbondale, Illinois campus. The CFB boiler will be used for cogeneration. Funding for this project was made possible by the State of Illinois Capital Development Board. Illinois coal will be fired in this CFB boiler. This paper provides a description of the planning process and design of the CFB boiler and related equipment with specific emphasis on particulate removal and recirculation. The startup of this new installation is scheduled for the summer of 1996, with commercial operation by fall of 1996.

  11. Looking Southwest at Reactor Box Furnaces With Reactor Boxes and ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    Looking Southwest at Reactor Box Furnaces With Reactor Boxes and Repossessed Uranium in Recycle Recovery Building - Hematite Fuel Fabrication Facility, Recycle Recovery Building, 3300 State Road P, Festus, Jefferson County, MO

  12. Tdp studies and tests for C. A. Energia Electrica de Venezuela (enelven) at planta ramon laguna, units RL-17 and RL-10. Volume 2. Unit RL-10 boiler condition assessment report. Export trade information

    SciTech Connect

    Not Available

    1991-03-28

    The study, conducted by Babcock and Wilcox, was funded by the U.S. Trade and Development agency on behalf of Enelven. In order to maximize generated power output and minimize operating costs at Planta Ramon Laguna, tests were done to evaluate the condition of equipment at the plant. In order to identify any damage and determine the operating output of each unit, assessments were done of the furnaces, boilers, generators and boiler feed pumps being used in the plant. The report presents the results of these tests. This is the second of three volumes and it includes the following section: (1) Condition Assessment of Unit RL-10 Boiler.

  13. GPU v. B and W lawsuit review and its effect on TMI-1 (Docket 50-289)

    SciTech Connect

    Not Available

    1983-09-01

    This report documents a review by the Nuclear Regulatory Commission (NRC) staff of the General Public Utilities Corporation, et al. v. the Babcock and Wilcox Company, et al. (GPU v. B and W) lawsuit record to assess whether any of the staff's previous conclusions or their principal bases presented at the Three Mile Island Unit 1 (TMI-1) restart hearing, supporting restart of TMI-1, should be amended in light of the information contained in the lawsuit record. Details of the lawsuit record are provided in the appendices contained in Volume II of this report.

  14. Guidelines for preparing and reviewing applications for the licensing of non-power reactors: Standard review plan and acceptance criteria. NUREG - 1537, Part 2

    SciTech Connect

    1996-02-01

    NUREG - 1537, Part 2 gives guidance on the conduct of licensing action reviews to NRC staff who review non-power reactor licensing applications. These licensing actions include construction permits and initial operating licenses, license renewals, amendments, conversions from highly enriched uranium to low-enriched uranium, decommissioning, and license termination.

  15. Comment on a Wilcox Test Statistic for Comparing Means When Variances Are Unequal.

    ERIC Educational Resources Information Center

    Hsiung, Tung-Hsing; And Others

    1994-01-01

    The alternative proposed by Wilcox (1989) to the James second-order statistic for comparing population means when variances are heterogeneous can sometimes be invalid. The degree to which the procedure is invalid depends on differences in sample size, the expected values of the observations, and population variances. (SLD)

  16. Growing Readers: Wendy Wilcox--West Bloomfield Township Public Library, MI

    ERIC Educational Resources Information Center

    Library Journal, 2005

    2005-01-01

    In 2001 youth services librarian Wendy Wilcox begged her boss for the chance to make West Bloomfield Township Public Library (WBTPL) one of 20 demonstration sites for the Public Library Association (PLA)/Association for Library Service to Children initiative Every Child Ready To Read. While all participating libraries teach parents and caregivers…

  17. Standardizing electrocoagulation reactor design: iron electrodes for NOM removal.

    PubMed

    Dubrawski, Kristian L; Mohseni, Madjid

    2013-03-01

    A novel systematic approach for reactor design was described for iron electrocoagulation (EC) and applied to drinking water treatment. Suwannee NOM was used as a model compound; performance was quantified by UV-abs-254 and DOC removal. Significant EC design variables were identified and examined: current density (i) (2.43-26.8 mA cm(-2)), coagulant or charge loading rate (CLR) (100-1000 CL(-1) min(-1)), and flocculation methodology ("fast" and "slow"). A correlation was found between increased i and decreased current efficiency (φ), optimum NOM removal was found at i ~10 mA cm(-2). A lower CLR showed greater total DOC removal, while a higher CLR led to less reactor residence time and required either longer flocculation times or greater coagulant dose for similar NOM removal. This paper defines and describes the four general EC "classes" of operation that have implications on several important measures of success: coagulant dose, electrical consumption, process speed, volumetric footprint, and post-EC flocculation requirements. Two classes were further examined with or without pH adjustment for DOC removal, showing that a "fast" EC mode without flocculation is more appropriate for smaller applications, while a "slow" EC mode is more effective for large permanent applications, where flocculation and settling can reduce coagulant and electrical consumption. The effect of pH adjustment showed greater impact with the "fast" dosing mode than with the "slow" mode, adjustment to pH 6 with the "fast" mode gave 13.8% and 29.1% greater DOC and UV-abs-254 removal, respectively, compared to the baseline without pH adjustment.

  18. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

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

  19. A critical assembly designed to measure neutronic benchmarks in support of the Space Nuclear Thermal Propulsion program

    NASA Astrophysics Data System (ADS)

    Parma, E. J.; Ball, R. M.; Hoovler, G. S.; Selcow, E. C.; Cerbone, R. J.

    1992-10-01

    A reactor designed to perform criticality experiments in support of the Space Nuclear Thermal Propulsion program is currently in operation at the Sandia National Laboratories' reactor facility. The reactor is a small, water-moderated system that uses highly enriched uranium particle fuel in a 19-element configuration. Its purpose is to obtain neutronic measurements under a variety of experimental conditions that are subsequently used to benchmark reactor-design computer codes. Brookhaven National Laboratory, Babcock & Wilcox, and Sandia National Laboratories participated in determining the reactor's performance requirements, design, follow on experimentation, and in obtaining the licensing approvals. Brookhaven National Laboratory is primarily responsible for the analytical support, Babcock & Wilcox the hardware design, and Sandia National Laboratories the operational safety. All of the team members participate in determining the experimentation requirements, performance, and data reduction. Initial criticality was achieved in October 1989. An over-all description of the reactor is presented along with key design features and safety-related aspects.

  20. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    NASA Technical Reports Server (NTRS)

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

  1. 76 FR 31381 - Office Of New Reactors; Proposed Revision 4 to Standard Review Plan; Section 8.1 on Electric...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-05-31

    ... COMMISSION Office Of New Reactors; Proposed Revision 4 to Standard Review Plan; Section 8.1 on Electric Power... Analysis Reports for Nuclear Power Plants,'' on a proposed Revision 4 to Standard Review Plan (SRP), Section 8.1 on ``Electric Power--Introduction,'' (Agencywide Documents Access and Management System...

  2. NEUTRONIC REACTOR CHARGING AND DISCHARGING

    DOEpatents

    Zinn, W.H.

    1959-07-14

    A method and arrangement is presented for removing a fuel element from a neutronic reactor tube through which a liquid coolant is being circulaled. The fuel element is moved into a section of the tube beyond the reactor proper, and then the coolant in the tube between the fuel element and the reactor proper is frozen, so that the fuel element may be removed from the tube without loss of the coolant therein. The method is particularly useful in the case of a liquid metal- cooled reactor.

  3. Fast quench reactor and method

    DOEpatents

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

    1998-05-12

    A fast quench reactor includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This ``freezes`` the desired end product(s) in the heated equilibrium reaction stage. 7 figs.

  4. A critical assembly designed to measure neutronic benchmarks in support of the space nuclear thermal propulsion program

    NASA Astrophysics Data System (ADS)

    Parma, Edward J.; Ball, Russell M.; Hoovler, Gary S.; Selcow, Elizabeth C.; Cerbone, Ralph J.

    1993-01-01

    A reactor designed to perform criticality experiments in support of the Space Nuclear Thermal Propulsion program is currently in operation at the Sandia National Laboratories' reactor facility. The reactor is a small, water-moderated system that uses highly enriched uranium particle fuel in a 19-element configuration. Its purpose is to obtain neutronic measurements under a variety of experimental conditions that are subsequently used to benchmark rector-design computer codes. Brookhaven National Laboratory, Babcock & Wilcox, and Sandia National Laboratories participated in determining the reactor's performance requirements, design, follow-on experimentation, and in obtaining the licensing approvals. Brookhaven National Laboratory is primarily responsible for the analytical support, Babcock & Wilcox the hardware design, and Sandia National Laboratories the operational safety. All of the team members participate in determining the experimentation requirements, performance, and data reduction. Initial criticality was achieved in October 1989. An overall description of the reactor is presented along with key design features and safety-related aspects.

  5. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

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

  6. Thermochemical reactor systems and methods

    SciTech Connect

    Lipinski, Wojciech; Davidson, Jane Holloway; Chase, Thomas Richard

    2016-11-29

    Thermochemical reactor systems that may be used to produce a fuel, and methods of using the thermochemical reactor systems, utilizing a reactive cylindrical element, an optional energy transfer cylindrical element, an inlet gas management system, and an outlet gas management system.

  7. Fast reactors and nuclear nonproliferation

    SciTech Connect

    Avrorin, E.N.; Rachkov, V.I.; Chebeskov, A.N.

    2013-07-01

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

  8. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

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

  9. REACTOR PHYSICS MODELING OF SPENT NUCLEAR RESEARCH REACTOR FUEL FOR SNM ATTRIBUTION AND NUCLEAR FORENSICS

    SciTech Connect

    Sternat, M.; Beals, D.; Webb, R.; Nichols, T.

    2010-06-09

    Nuclear research reactors are the least safeguarded type of reactor; in some cases this may be attributed to low risk and in most cases it is due to difficulty from dynamic operation. Research reactors vary greatly in size, fuel type, enrichment, power and burnup providing a significant challenge to any standardized safeguard system. If a whole fuel assembly was interdicted, based on geometry and other traditional forensics work, one could identify the material's origin fairly accurately. If the material has been dispersed or reprocessed, in-depth reactor physics models may be used to help with the identification. Should there be a need to attribute research reactor fuel material, the Savannah River National Laboratory would perform radiochemical analysis of samples of the material as well as other non-destructive measurements. In depth reactor physics modeling would then be performed to compare to these measured results in an attempt to associate the measured results with various reactor parameters. Several reactor physics codes are being used and considered for this purpose, including: MONTEBURNS/ORIGEN/MCNP5, CINDER/MCNPX and WIMS. In attempt to identify reactor characteristics, such as time since shutdown, burnup, or power, various isotopes are used. Complexities arise when the inherent assumptions embedded in different reactor physics codes handle the isotopes differently and may quantify them to different levels of accuracy. A technical approach to modeling spent research reactor fuel begins at the assembly level upon acquiring detailed information of the reactor to be modeled. A single assembly is run using periodic boundary conditions to simulate an infinite lattice which may be repeatedly burned to produce input fuel isotopic vectors of various burnups for a core level model. A core level model will then be constructed using the assembly level results as inputs for the specific fuel shuffling pattern in an attempt to establish an equilibrium cycle. The

  10. Regional appraisal of the Wilcox Group in Texas for subsurface storage of fluid wastes: Part 1: Geology

    USGS Publications Warehouse

    Jones, Paul Hastings; Stevens, P.R.; Wasselman, J.B.; Wallace, R.H.

    1976-01-01

    Sandy beds of the Wilcox Group in Texas are underlain and overlain by clays of the Midway and Claiborne Groups. The Wilcox is divided by a persistent shale wedge. Contrasting delta systems described as high-constructive and high-destructive exist in the divisions. Merged high constructive deltas characterize the lower, while the upper Wilcox was deposited as a high destructive system. Maximum sand development occurs downdip parallel to regional depositional strike in the delta facies. Further downdip, shales replace sands, and sediments are (abnormally high pressured) geopressured. The structure o5 the updip port[on is that of a stable shelf. Downdip, greatly thickened beds are cut by large faults. These faults and the resultant deposits are recognized as being the result of a previously unnamed phenomenon--here termed hydrothermal tectonism. This process is keyed to the thermal diagenesis of the clay mineral montmorillonite and resultant transfer of overburden pressure from the rock framework to the contained fluid.

  11. Reactor and method of operation

    DOEpatents

    Wheeler, John A.

    1976-08-10

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

  12. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect

    R. Wigeland; J. Cahalan

    2009-09-01

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

  13. Overview and status of first 25 MW(e) IR-CFB boiler in India

    SciTech Connect

    Kavidass, S.; Bakshi, V.K.; Diwakar, K.K.

    1997-12-31

    The Babcock and Wilcox (B and W) internal recirculation CFB (IR-CFB) boiler is unique in design. Worldwide, B and W offers IR-CFB boilers up to 150 MW(e) both reheat and non-reheat, and is pursuing units up to 300 MW(e). This paper discusses an overview and status of the construction, commissioning, initial start-up operation and milestones of the ongoing 25 MW(e) IR-CFB boiler project at Kanoria Chemicals and Industries Ltd., Renukoot, India. This IR-CFB boiler is designed, supplied and installed by Thermax Babcock and Wilcox Ltd. (TBW), a joint venture company of the B and W and Thermax in India. The boiler parameters are, steam flow of 29.2 kg/s (23,420 lbs/hr), 6.4 MPa (925 psig), and 485 C (905 F) with feedwater temperature of 180 C (356 F). The boiler will utilize high-ash content (> 45%), subbituminous coal with a heating value of 3,500 KCal/kg (6,300 Btu/lb). This paper also discusses the various aspects of the boiler design, performance, auxiliary equipment, advantages and initial start-up operating performance.

  14. Nuclear Reactors and Technology; (USA)

    SciTech Connect

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

    1991-01-01

    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.

  15. 76 FR 23630 - Office of New Reactors; Proposed Revision 2 to Standard Review Plan, Section 1.0 on Introduction...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-04-27

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Office of New Reactors; Proposed Revision 2 to Standard Review Plan, Section 1.0 on Introduction...), Section 1.0, ``Introduction and Interfaces'' (Agencywide Documents Access and Management System...

  16. A new Mapinguari Papavero & Wilcox (Diptera, Mydidae, Mydinae) from Minas Gerais State, Brazil.

    PubMed

    Calhau, Julia; Lamas, Carlos José Einicker; Nihei, Silvio Shigueo

    2016-10-31

    Mapinguari Papavero & Wilcox, 1974 (Diptera, Mydidae, Mydinae) is a very rare monotypic genus, with the type-species, M. politus (Wiedemann, 1828), occurring exclusively in Amazonia. With the description of Mapinguari uai sp. nov. from a remnant of the Atlantic Forest in southeastern Brazil, the distribution of the genus is greatly expanded. In addition, an updated diagnosis for the genus and its type-species is provided.

  17. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Pressurized Water Reactor Standard Core Loading Benchmark Problem

    NASA Astrophysics Data System (ADS)

    Arzu Alpan, F.; Kulesza, Joel A.

    2016-02-01

    This paper compares contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a pressurized water reactor calculational benchmark problem with a standard out-in core loading. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission and used the Oak Ridge National Laboratory two-dimensional discrete ordinates code DORT and the BUGLE-93 cross-section library for the calculations. In this paper, a Westinghouse three-dimensional discrete ordinates code with parallel processing, the RAPTOR-M3G code was used. A variety of cross section libraries were used with RAPTOR-M3G including the BUGLE-93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory, and the broad-group ALPAN-VII.0 cross-section library developed at Westinghouse. In comparing the calculation-to-calculation reaction rates using the BUGLE-93 cross-section library at the thermal shield, pressure vessel, and cavity capsules, for eleven dosimetry reaction rates, a maximum relative difference of 5% was observed, with the exception of 65Cu(n,2n) in the pressure vessel capsule that had a 90% relative difference with respect to the reference results. It is thought that the 65Cu(n,2n) reaction rate reported in the reference for the pressure vessel capsule is not correct. In considering the libraries developed after BUGLE-93, a maximum relative difference of 12% was observed in reaction rates, with respect to the reference results, for 237Np(n,f) in the cavity capsule using BUGLE-B7.

  18. New CEM systems measure up to the job

    SciTech Connect

    Bright, R.K.

    1996-11-01

    The Clean Air Act Amendments of 1990 require utilities to face the problem of providing power while ensuring their plants comply with clean air standards. This paper describes a Clean Environment Development Facility (CEM) constructed by Babcock and Wilcox research. The facility is dedicated to finding cleaner and more efficient means of producing power and has incorporated advanced continuous emissions monitoring systems.

  19. NTRE extended life feasibility assessment

    NASA Technical Reports Server (NTRS)

    1993-01-01

    Results of a feasibility analysis of a long life, reusable nuclear thermal rocket engine are presented in text and graph form. Two engine/reactor concepts are addressed: the Particle Bed Reactor (PBR) design and the Commonwealth of Independent States (CIS) concept. Engine design, integration, reliability, and safety are addressed by various members of the NTRE team from Aerojet Propulsion Division, Energopool (Russia), and Babcock & Wilcox.

  20. Fast quench reactor and method

    DOEpatents

    Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

    2002-09-24

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

  1. Fast quench reactor and method

    DOEpatents

    Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

    1998-01-01

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

  2. Fast quench reactor and method

    DOEpatents

    Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

    2002-01-01

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

  3. The origin and distribution of HAPs elements in relation to maceral composition of the A1 lignite bed (Paleocene, Calvert Bluff Formation, Wilcox Group), Calvert mine area, east-central Texas

    USGS Publications Warehouse

    Crowley, S.S.; Warwick, P.D.; Ruppert, L.F.; Pontolillo, J.

    1997-01-01

    The origin and distribution of twelve potentially Hazardous Air Pollutants (HAPs; As, Be, Cd, Cr, Co, Hg, Mn, Ni, Pb Sb, Se, and U) identified in the 1990 Clean Air Act Amendments were examined in relation to the maceral composition of the A1 bed (Paleocene, Calvert Bluff Formation, Wilcox Group) of the Calvert mine in east-central Texas. The 3.2 m-thick A1 bed was divided into nine incremental channel samples (7 lignite samples and 2 shaley coal samples) on the basis of megascopic characteristics. Results indicate that As, Cd, Cr, Ni, Pb, Sb, and U are strongly correlated with ash yield and are enriched in the shaley coal samples. We infer that these elements are associated with inorganic constituents in the coal bed and may be derived from a penecontemporaneous stream channel located several kilometers southeast of the mining block. Of the HAPs elements studied, Mn and Hg are the most poorly correlated to ash yield. We infer an organic association for Mn; Hg may be associated with pyrite. The rest of the trace elements (Be, Co, and Se) are weakly correlated with ash yield. Further analytical work is necessary to determine the mode of occurrence for these elements. Overall, concentrations of the HAPs elements are generally similar to or less than those reported in previous studies of lignites of the Wilcox Group, east-central region, Texas. Petrographic analysis indicates the following ranges in composition for the seven lignite samples: liptinites (5-8%), huminites (88-95%), and inertinites (trace amounts to 7%). Samples from the middle portion of the A1 bed contain abundant crypto-eugelinite compared to the rest of the samples; this relationship suggests that the degradation of plant material was an important process during the development of the peat mire. With the exception of Hg and Mn, relatively low levels of the HAPs elements studied are found in the samples containing abundant crypto-eugelinite. We infer that the peat-forming environment for this portion

  4. The IRIS network site at the Wilcox Solar Observatory

    NASA Technical Reports Server (NTRS)

    Hoeksema, J. T.; Scherrer, P. H.

    1991-01-01

    The site for the International Research on the Interior of the Sun (IRIS) instrument housed at the Wilcox Solar Observatory at Stanford University (near San Francisco, USA) is described together with the instrument operation procedure. The IRIS instrument, which measures global oscillations of the sun, operates continuously every clear day since it was installed in August 1987.

  5. Environmental Assessment: Geothermal Energy Geopressure Subprogram. Gulf Coast Well Drilling and Testing Activity (Frio, Wilcox, and Tuscaloosa Formations, Texas and Louisiana)

    SciTech Connect

    1981-09-01

    The Department of Energy (DOE) has initiated a program to evaluate the feasibility of developing the geothermal-geopressured energy resources of the Louisiana-Texas Gulf Coast. As part of this effort, DOE is contracting for the drilling of design wells to define the nature and extent of the geopressure resource. At each of several sites, one deep well (4000-6400 m) will be drilled and flow tested. One or more shallow wells will also be drilled to dispose of geopressured brines. Each site will require about 2 ha (5 acres) of land. Construction and initial flow testing will take approximately one year. If initial flow testing is successful, a continuous one-year duration flow test will take place at a rate of up to 6400 m{sup 3} (40,000 bbl) per day. Extensive tests will be conducted on the physical and chemical composition of the fluids, on their temperature and flow rate, on fluid disposal techniques, and on the reliability and performance of equipment. Each project will require a maximum of three years to complete drilling, testing, and site restoration.

  6. Turbulent magnetic pumping in a Babcock-Leighton solar dynamo model

    NASA Astrophysics Data System (ADS)

    Guerrero, G.; de Gouveia Dal Pino, E. M.

    2008-07-01

    Context: The turbulent pumping effect corresponds to the transport of magnetic flux due to the presence of density and turbulence gradients in convectively unstable layers. In the induction equation it appears as an advective term and for this reason it is expected to be important in the solar and stellar dynamo processes. Aims: We explore the effects of turbulent pumping in a flux-dominated Babcock-Leighton solar dynamo model with a solar-like rotation law. Methods: As a first step, only vertical pumping has been considered through the inclusion of a radial diamagnetic term in the induction equation. In the second step, a latitudinal pumping term was included and then, a near-surface shear was included. Results: The results reveal the importance of the pumping mechanism in solving current limitations in mean field dynamo modeling, such as the storage of the magnetic flux and the latitudinal distribution of the sunspots. If a meridional flow is assumed to be present only in the upper part of the convective zone, it is the full turbulent pumping that regulates both the period of the solar cycle and the latitudinal distribution of the sunspot activity. In models that consider shear near the surface, a second shell of toroidal field is generated above r=0.95~R⊙ at all latitudes. If the full pumping is also included, the polar toroidal fields are efficiently advected inwards, and the toroidal magnetic activity survives only at the observed latitudes near the equator. With regard to the parity of the magnetic field, only models that combine turbulent pumping with near-surface shear always converge to the dipolar parity. Conclusions: This result suggests that, under the Babcock-Leighton approach, the equartorward motion of the observed magnetic activity is governed by the latitudinal pumping of the toroidal magnetic field rather than by a large scale coherent meridional flow. Our results support the idea that the parity problem is related to the quadrupolar imprint of

  7. A novel sorbent for transport reactors and fluidized bed reactors

    SciTech Connect

    Copeland, R.; Cesario, M.; Gershanovich, Y.; Sibold, J.; Windecker, B.

    1998-12-31

    Coal Fired Gasifier Combined Cycles (GCC) have both high efficiency and very low emissions. GCCs critically need a method of removing the H{sub 2}S produced from the sulfur in the coal from the hot gases. There has been extensive research on hot gas cleanup systems, focused on the use of a zinc oxide based sorbent (e.g., zinc titanate). TDA Research, Inc. (TDA) is developing a novel sorbent with improved attrition resistance for transport reactors and fluidized bed reactors. The authors are testing sorbents at conditions simulating the operating conditions of the Pinon Pine clean coal technology plant. TDA sulfided several different formulations at 538 C and found several that have high sulfur capacity when tested in a fluidized bed reactor. TDA initiated sorbent regeneration at 538 C. The sorbents retained chemical activity with multiple cycles. Additional tests will be conducted to evaluate the best sorbent formulation.

  8. Plane-polar Fresnel and far-field computations using the Fresnel-Wilcox and Jacobi-Bessel expansions. [for large aperture antennas

    NASA Technical Reports Server (NTRS)

    Rahmat-Samii, Y.; Galindo-Israel, V.

    1981-01-01

    It is pointed out that the computation of the Fresnel fields for large aperture antennas is significant for many applications. The present investigation is concerned with an approach for the effective utilization of the coefficients of the Jacobi-Bessel series for the far-field to obtain an analytically continuous representation of the antenna field which is valid from the Fresnel region into the far field. Attention is given to exact formulations and closed form solutions, Fresnel and Fresnel small angle approximations, aspects of field expansion, the accuracy of the Fresnel and Fresnel small angle approximations, and the Jacobi-Bessel expansion applied to the Fresnel small angle approximation.

  9. Results and Analysis of the Infrastructure Request for Information (DE-SOL-0008318)

    SciTech Connect

    Heidrich, Brenden John

    2015-07-01

    The Department of Energy (DOE) Office of Nuclear Energy (NE) released a request for information (RFI) (DE-SOL-0008318) for “University, National Laboratory, Industry and International Input on Potential Office of Nuclear Energy Infrastructure Investments” on April 13, 2015. DOE-NE solicited information on five specific types of capabilities as well as any others suggested by the community. The RFI proposal period closed on June 19, 2015. From the 26 responses, 34 individual proposals were extracted. Eighteen were associated with a DOE national laboratory, including Argonne National Laboratory (ANL), Brookhaven National Laboratory (BNL), Idaho National Laboratory (INL), Los Alamos National Laboratory (LANL), Pacific Northwest National Laboratory (PNNL) and Sandia National Laboratory (SNL). Oak Ridge National Laboratory (ORNL) was referenced in a proposal as a proposed capability location, although the proposal did not originate with ORNL. Five US universities submitted proposals (Massachusetts Institute of Technology, Pennsylvania State University, Rensselaer Polytechnic Institute, University of Houston and the University of Michigan). Three industrial/commercial institutions submitted proposals (AREVA NP, Babcock and Wilcox (B&W) and the Electric Power Research Institute (EPRI)). Eight major themes emerged from the submissions as areas needing additional capability or support for existing capabilities. Two submissions supported multiple areas. The major themes are: Advanced Manufacturing (AM), High Performance Computing (HPC), Ion Irradiation with X-Ray Diagnostics (IIX), Ion Irradiation with TEM Visualization (IIT), Radiochemistry Laboratories (RCL), Test Reactors, Neutron Sources and Critical Facilities (RX) , Sample Preparation and Post-Irradiation Examination (PIE) and Thermal-Hydraulics Test Facilities (THF).

  10. Integrated reformer and shift reactor

    DOEpatents

    Bentley, Jeffrey M.; Clawson, Lawrence G.; Mitchell, William L.; Dorson, Matthew H.

    2006-06-27

    A hydrocarbon fuel reformer for producing diatomic hydrogen gas is disclosed. The reformer includes a first reaction vessel, a shift reactor vessel annularly disposed about the first reaction vessel, including a first shift reactor zone, and a first helical tube disposed within the first shift reactor zone having an inlet end communicating with a water supply source. The water supply source is preferably adapted to supply liquid-phase water to the first helical tube at flow conditions sufficient to ensure discharge of liquid-phase and steam-phase water from an outlet end of the first helical tube. The reformer may further include a first catalyst bed disposed in the first shift reactor zone, having a low-temperature shift catalyst in contact with the first helical tube. The catalyst bed includes a plurality of coil sections disposed in coaxial relation to other coil sections and to the central longitudinal axis of the reformer, each coil section extending between the first and second ends, and each coil section being in direct fluid communication with at least one other coil section.

  11. Research to understand the embrittlement behavior of Yankee/BR3 surveillance plate and other outlier RPV steels

    SciTech Connect

    Fabry, A.; Velde, J. van de; Puzzolante, J.L.; Ransbeeck, T. van; Verstrepen, A.; Carter, R.G.; Petrova, T.

    1996-12-31

    The reactor pressure vessels at the Yankee Rowe and Belgian BR3 nuclear plants were constructed by Babcock and Wilcox in 1958. The plates of an open-hearth fabrication were welded using a submerged-arc process with Linde 80 flux as the filler. The original surveillance programs at the two plants were limited to representative A302B specimens; they feature similar chemistries as the ASTM reference plate, but coarser microstructure. The present testing program includes sixteen Charpy-V and four tensile specimens of the surveillance plate, irradiated at BR3 at a dose rate of {approx} 7 E10 cm{sup {minus}2}.s-1 (> 1 MeV) over a period of 25 years; the investigation also addresses annealing and notch orientation effects. The new experimental results are compared to previously published data for the same and/or related melts. The Yankee/BR3 surveillance plate displays an anomalously large 41J C{sub v}-shift as compared to the ASTM reference plate and to Regulatory predictions. Some of the Linde 80 welds investigated by the BR3 program are also found to behave as outliers. The data are evaluated in the light of state-of-the-art damage modeling and fracture micromechanics concepts, which are currently being incorporated into a new, consolidated strategy for improved RPV surveillance. The approach makes extensive use of the information contained in the load-deflection response of the instrumented C{sub v} test. The implications of such analysis in terms of RPV steel embrittlement trend curve development are discussed. 90 refs.

  12. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

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

  13. A comparison of three self-tuning control algorithms developed for the Bristol-Babcock controller

    SciTech Connect

    Tapp, P.A.

    1992-04-01

    A brief overview of adaptive control methods relating to the design of self-tuning proportional-integral-derivative (PID) controllers is given. The methods discussed include gain scheduling, self-tuning, auto-tuning, and model-reference adaptive control systems. Several process identification and parameter adjustment methods are discussed. Characteristics of the two most common types of self-tuning controllers implemented by industry (i.e., pattern recognition and process identification) are summarized. The substance of the work is a comparison of three self-tuning proportional-plus-integral (STPI) control algorithms developed to work in conjunction with the Bristol-Babcock PID control module. The STPI control algorithms are based on closed-loop cycling theory, pattern recognition theory, and model-based theory. A brief theory of operation of these three STPI control algorithms is given. Details of the process simulations developed to test the STPI algorithms are given, including an integrating process, a first-order system, a second-order system, a system with initial inverse response, and a system with variable time constant and delay. The STPI algorithms` performance with regard to both setpoint changes and load disturbances is evaluated, and their robustness is compared. The dynamic effects of process deadtime and noise are also considered. Finally, the limitations of each of the STPI algorithms is discussed, some conclusions are drawn from the performance comparisons, and a few recommendations are made. 6 refs.

  14. A comparison of three self-tuning control algorithms developed for the Bristol-Babcock controller

    SciTech Connect

    Tapp, P.A.

    1992-04-01

    A brief overview of adaptive control methods relating to the design of self-tuning proportional-integral-derivative (PID) controllers is given. The methods discussed include gain scheduling, self-tuning, auto-tuning, and model-reference adaptive control systems. Several process identification and parameter adjustment methods are discussed. Characteristics of the two most common types of self-tuning controllers implemented by industry (i.e., pattern recognition and process identification) are summarized. The substance of the work is a comparison of three self-tuning proportional-plus-integral (STPI) control algorithms developed to work in conjunction with the Bristol-Babcock PID control module. The STPI control algorithms are based on closed-loop cycling theory, pattern recognition theory, and model-based theory. A brief theory of operation of these three STPI control algorithms is given. Details of the process simulations developed to test the STPI algorithms are given, including an integrating process, a first-order system, a second-order system, a system with initial inverse response, and a system with variable time constant and delay. The STPI algorithms' performance with regard to both setpoint changes and load disturbances is evaluated, and their robustness is compared. The dynamic effects of process deadtime and noise are also considered. Finally, the limitations of each of the STPI algorithms is discussed, some conclusions are drawn from the performance comparisons, and a few recommendations are made. 6 refs.

  15. Wilcox sandstone reservoirs in the deep subsurface along the Texas Gulf Coast: their potential for production of geopressured geothermal energy. Report of Investigations No. 117

    SciTech Connect

    Debout, D.G.; Weise, B.R.; Gregory, A.R.; Edwards, M.B.

    1982-01-01

    Regional studies of the lower Eocene Wilcox Group in Texas were conducted to assess the potential for producing heat energy and solution methane from geopressured fluids in the deep-subsurface growth-faulted zone. However, in addition to assembling the necessary data for the geopressured geothermal project, this study has provided regional information of significance to exploration for other resources such as lignite, uranium, oil, and gas. Because the focus of this study was on the geopressured section, emphasis was placed on correlating and mapping those sandstones and shales occurring deeper than about 10,000 ft. The Wilcox and Midway Groups comprise the oldest thick sandstone/shale sequence of the Tertiary of the Gulf Coast. The Wilcox crops out in a band 10 to 20 mi wide located 100 to 200 mi inland from the present-day coastline. The Wilcox sandstones and shales in the outcrop and updip shallow subsurface were deposited primarily in fluvial environments; downdip in the deep subsurface, on the other hand, the Wilcox sediments were deposited in large deltaic systems, some of which were reworked into barrier-bar and strandplain systems. Growth faults developed within the deltaic systems, where they prograded basinward beyond the older, stable Lower Cretaceous shelf margin onto the less stable basinal muds. Continued displacement along these faults during burial resulted in: (1) entrapment of pore fluids within isolated sandstone and shale sequences, and (2) buildup of pore pressure greater than hydrostatic pressure and development of geopressure.

  16. An example of Ensemble Kalman Filter data assimilation in a Babcock-Leighton solar dynamo model

    NASA Astrophysics Data System (ADS)

    Dikpati, Mausumi; Anderson, Jeffrey L.

    2016-05-01

    Atmospheric and oceanic prediction models have been greatly advanced over the past 40 years by using modern data assimilation techniques. Application of similar techniques in solar models started about 7 years ago. However, acceptance of such techniques by the solar community has been slow to develop. In order to make accurate predictions of solar activity as well as reconstruction of certain model parameters that cannot be directly measured, it will be essential to implement sophisticated data assimilation techniques as used by atmospheric and oceanic models. We will present here an example of parameter reconstruction, namely the time variation in meridional flow-speed, done by assimilating data into a Babcock-Leighton solar dynamo model in the framework of NCAR's Data Assimilation Research Testbed (NCAR-DART). By performing many 'Observing System Simulation Experiments' (OSSEs) we find that an optimally good reconstruction in time series of meridional circulation can be obtained by using 16 ensemble members and assimilating one magnetic observation with less than 40 percent observational error. However, the RMS error in reconstruction reduces with increase in ensemble size, increase in number of observations and decrease in observational error. We also find that assimilation of magnetic field observations taken from low-to-mid latitudes at the surface compared to any other locations produces the best reconstruction. We will close by showing that assimilation cycle of 15 days is optimal; generally a longer assimilation cycle deteriorates the results, but the Dynamo DART system needs a minimum time to develop the dynamics.

  17. Reactor monitoring and safeguards using antineutrino detectors

    SciTech Connect

    Bowden, N S

    2008-09-07

    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 offer 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 efforts to develop this monitoring technique are underway across the globe.

  18. The IAEA international conference on fast reactors and related fuel cycles: highlights and main outcomes

    SciTech Connect

    Monti, S.; Toti, A.

    2013-07-01

    The 'International Conference on Fast Reactors and Related Fuel Cycles', which is regularly held every four years, represents the main international event dealing with fast reactors technology and related fuel cycles options. Main topics of the conference were new fast reactor concepts, design and simulation capabilities, safety of fast reactors, fast reactor fuels and innovative fuel cycles, analysis of past experience, fast reactor knowledge management. Particular emphasis was put on safety aspects, considering the current need of developing and harmonizing safety standards for fast reactors at the international level, taking also into account the lessons learned from the accident occurred at the Fukushima- Daiichi nuclear power plant in March 2011. Main advances in the several key areas of technological development were presented through 208 oral presentations during 41 technical sessions which shows the importance taken by fast reactors in the future of nuclear energy.

  19. Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

    SciTech Connect

    D. Kokkinos

    2005-04-28

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.

  20. Chromatographic and Related Reactors.

    DTIC Science & Technology

    1988-01-07

    Mechanism of the Base - Catalyzed Esterification of Tetrachloroterephthaloyl Chloride with HPLC and Carbon-13 FT-NMR", by Stanley H. Langer, Alexander Chu...In connection with this, new information on the kinetics and course of the base catalyzed esterification reaction of tetrachloroterephthaloyl...the course of reaction which could be clarified with NMR. With a knowledge of the base catalyzed solvolysis rate in a methanolic mobile phase, it was

  1. BOILING SLURRY REACTOR AND METHOD FO CONTROL

    DOEpatents

    Petrick, M.; Marchaterre, J.F.

    1963-05-01

    The control of a boiling slurry nuclear reactor is described. The reactor consists of a vertical tube having an enlarged portion, a steam drum at the top of the vertical tube, and at least one downcomer connecting the steam drum and the bottom of the vertical tube, the reactor being filled with a slurry of fissionabie material in water of such concentration that the enlarged portion of the vertical tube contains a critical mass. The slurry boils in the vertical tube and circulates upwardly therein and downwardly in the downcomer. To control the reactor by controlling the circulation of the slurry, a gas is introduced into the downcomer. (AEC)

  2. Italian hybrid and fission reactors scenario analysis

    SciTech Connect

    Ciotti, M.; Manzano, J.; Sepielli, M.

    2012-06-19

    Italy is a country where a long tradition of studies both in the fission and fusion field is consolidated; nevertheless a strong public opinion concerned with the destination of the Spent Nuclear Fuel hinders the development of nuclear power. The possibility to a severe reduction of the NSF mass generated from a fleet of nuclear reactors employing an hypothetical fusionfission hybrid reactor has been investigated in the Italian framework. The possibility to produce nuclear fuel for the fission nuclear reactors with the hybrid reactor was analyzed too.

  3. Hydrogasification reactor and method of operating same

    SciTech Connect

    Hobbs, Raymond; Karner, Donald; Sun, Xiaolei; Boyle, John; Noguchi, Fuyuki

    2013-09-10

    The present invention provides a system and method for evaluating effects of process parameters on hydrogasification processes. The system includes a hydrogasification reactor, a pressurized feed system, a hopper system, a hydrogen gas source, and a carrier gas source. Pressurized carbonaceous material, such as coal, is fed to the reactor using the carrier gas and reacted with hydrogen to produce natural gas.

  4. Reactor technology assessment and selection utilizing systems engineering approach

    SciTech Connect

    Zolkaffly, Muhammed Zulfakar; Han, Ki-In

    2014-02-12

    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.

  5. Reactor technology assessment and selection utilizing systems engineering approach

    NASA Astrophysics Data System (ADS)

    Zolkaffly, Muhammed Zulfakar; Han, Ki-In

    2014-02-01

    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.

  6. 75 FR 36126 - Office of New Reactors; Proposed Revision to Standard Review Plan Section 13.6.1, Revision 1 on...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-06-24

    ... Physical Security--Combined License and Operating Reactors AGENCY: Nuclear Regulatory Commission (NRC... Revision 1 to Standard Review Plan (SRP), Section 13.6.1 on ``Physical Security--Combined License and...). The Office of Nuclear Security and Incident Response is revising SRP Section 13.6.1, which updates...

  7. Scanning tunneling microscope assembly, reactor, and system

    SciTech Connect

    Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

    2014-11-18

    An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

  8. Coal geology of the Paleocene-Eocene Calvert Bluff Formation (Wilcox Group) and the Eocene Manning Formation (Jackson Group) in east-central Texas; field trip guidebook for the Society for Organic Petrology, Twelfth Annual Meeting, The Woodlands, Texas, August 30, 1995

    USGS Publications Warehouse

    Warwick, Peter D.; Crowley, Sharon S.

    1995-01-01

    The Jackson and Wilcox Groups of eastern Texas (fig. 1) are the major lignite producing intervals in the Gulf Region. Within these groups, the major lignite-producing formations are the Paleocene-Eocene Calvert Bluff Formation (Wilcox) and the Eocene Manning Formation (Jackson). According to the Keystone Coal Industry Manual (Maclean Hunter Publishing Company, 1994), the Gulf Coast basin produces about 57 million short tons of lignite annually. The state of Texas ranks number 6 in coal production in the United States. Most of the lignite is used for electric power generation in mine-mouth power plant facilities. In recent years, particular interest has been given to lignite quality and the distribution and concentration of about a dozen trace elements that have been identified as potential hazardous air pollutants (HAPs) by the 1990 Clean Air Act Amendments. As pointed out by Oman and Finkelman (1994), Gulf Coast lignite deposits have elevated concentrations of many of the HAPs elements (Be, Cd, Co, Cr, Hg, Mn, Se, U) on a as-received gm/mmBtu basis when compared to other United States coal deposits used for fuel in thermo-electric power plants. Although regulations have not yet been established for acceptable emissions of the HAPs elements during coal burning, considerable research effort has been given to the characterization of these elements in coal feed stocks. The general purpose of the present field trip and of the accompanying collection of papers is to investigate how various aspects of east Texas lignite geology might collectively influence the quality of the lignite fuel. We hope that this collection of papers will help future researchers understand the complex, multifaceted interrelations of coal geology, petrology, palynology and coal quality, and that this introduction to the geology of the lignite deposits of east Texas might serve as a stimulus for new ideas to be applied to other coal basins in the U.S. and abroad.

  9. Gallium and Reactor Neutrino Anomalies

    NASA Astrophysics Data System (ADS)

    Acero, M. A.; Giunti, C.; Laveder, M.

    2009-03-01

    The observed deficit in the Gallium radioactive source experiments may be interpreted as a possible indication of active-sterile ν mixing. In the effective framework of two-neutrino mixing we obtain sin2ϑ≳0.03 and Δm≳0.1 eV. The compatibility of this result with the data of the Bugey reactor ν disappearance experiments is studied. It is found that the Bugey data present a hint of neutrino oscillations with 0.02≲sin2ϑ≲0.08 and Δm≈1.8 eV, which is compatible with the Gallium allowed region of the mixing parameters. This hint persists in the combined analysis of Gallium, Bugey, and Chooz data.

  10. 75 FR 42791 - Office of New Reactors; Proposed Revision 1 to Standard Review Plan; Section 13.5.1.1 on...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-07-22

    ... Review of Safety Analysis Reports for Nuclear Power Plants,'' on a proposed Revision 1 to Standard Review... and Management System (ADAMS) Accession No. ML101340264). The Office of New Reactors (NRO) is revising... review of design certification and combined license applications for NRO. The NRC staff intends...

  11. Advanced PPA Reactor and Process Development

    NASA Technical Reports Server (NTRS)

    Wheeler, Raymond; Aske, James; Abney, Morgan B.; Miller, Lee A.; Greenwood, Zachary

    2012-01-01

    Design and development of a second generation Plasma Pyrolysis Assembly (PPA) reactor is currently underway as part of NASA s Atmosphere Revitalization Resource Recovery effort. By recovering up to 75% of the hydrogen currently lost as methane in the Sabatier reactor effluent, the PPA helps to minimize life support resupply costs for extended duration missions. To date, second generation PPA development has demonstrated significant technology advancements over the first generation device by doubling the methane processing rate while, at the same time, more than halving the required power. One development area of particular interest to NASA system engineers is fouling of the PPA reactor with carbonaceous products. As a mitigation plan, NASA MSFC has explored the feasibility of using an oxidative plasma based upon metabolic CO2 to regenerate the reactor window and gas inlet ports. The results and implications of this testing are addressed along with the advanced PPA reactor development work.

  12. Instrumentation, Monitoring and NDE for New Fast Reactors

    SciTech Connect

    Bond, Leonard J.; Doctor, Steven; Bunch, Kyle; Good, Morris; Waltar, Alan E.

    2007-07-01

    The Global Nuclear Energy Partnership (GNEP) will require the development of actinide transmutation, which can most effectively be accomplished in a fast-spectrum reactor. To achieve higher standards of safety and reliability, operate with longer intervals between outages, and achieve high operating capacity factors, new instrumentation and on-line monitoring capabilities will be required-- during both fabrication and operation. This paper reports parts of a knowledge capture and technology state-of-the-art assessment for fast-reactor instrumentation and controls, monitoring and diagnostics. (authors)

  13. Compact Reactor

    SciTech Connect

    Williams, Pharis E.

    2007-01-30

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date.

  14. METHOD AND APPARATUS FOR CONTROLLING DIRECT-CYCLE NEUTRONIC REACTORS

    DOEpatents

    Reed, G.A.

    1961-01-10

    A control arrangement is offered for a boiling-water reactor. Boric acid is maintained in the water in the reactor and the amount in the reactor is controlled by continuously removing a portion of the water from the reactor, concentrating the boric acid by evaporating the water therefrom, returning a controlled amount of the acid to the reactor, and simultaneously controlling the water level by varying the rate of spent steam return to the reactor.

  15. Reactor antineutrinos and nuclear physics

    NASA Astrophysics Data System (ADS)

    Balantekin, A. B.

    2016-11-01

    Short-baseline reactor neutrino experiments successfully measured the neutrino parameters they set out to measure, but they also identified a shape distortion in the 5-7 MeV range as well as a reduction from the predicted value of the flux. Nuclear physics input into the calculations of reactor antineutrino spectra needs to be better refined if this anomaly is to be interpreted as due to sterile neutrino states.

  16. Coal gasification systems engineering and analysis. Volume 1: Executive summary

    NASA Technical Reports Server (NTRS)

    1980-01-01

    Feasibility analyses and systems engineering studies for a 20,000 tons per day medium Btu (MBG) coal gasification plant to be built by TVA in Northern Alabama were conducted. Major objectives were as follows: (1) provide design and cost data to support the selection of a gasifier technology and other major plant design parameters, (2) provide design and cost data to support alternate product evaluation, (3) prepare a technology development plan to address areas of high technical risk, and (4) develop schedules, PERT charts, and a work breakdown structure to aid in preliminary project planning. Volume one contains a summary of gasification system characterizations. Five gasification technologies were selected for evaluation: Koppers-Totzek, Texaco, Lurgi Dry Ash, Slagging Lurgi, and Babcock and Wilcox. A summary of the trade studies and cost sensitivity analysis is included.

  17. Ceramic oxygen transport membrane array reactor and reforming method

    DOEpatents

    Kelly, Sean M.; Christie, Gervase Maxwell; Robinson, Charles; Wilson, Jamie R.; Gonzalez, Javier E.; Doraswami, Uttam R.

    2016-11-08

    The invention relates to a commercially viable modular ceramic oxygen transport membrane reforming reactor configured using repeating assemblies of oxygen transport membrane tubes and catalytic reforming reactors.

  18. Coupled reactor kinetics and heat transfer model for heat pipe cooled reactors

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.; Houts, Michael

    2001-02-01

    Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). This paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities. .

  19. GaAs epitaxy using alternate pulses in a standard LP-MOVPE reactor

    NASA Astrophysics Data System (ADS)

    Wisser, J.; Czuprin, P.; Grundmann, D.; Balk, P.; Waschbüsch, M.; Lückerath, R.; Richter, W.

    1991-01-01

    Pulsed flow epitaxy (PFE) has been used to explore conditions under which atomic layer epitaxy (ALE), i.e. growth of one monolayer per cycle, may occur in a wafer scale LP-MOVPE reactor using trimethylgallium (TMG) and arsine as precursors. For certain combinations of experimental parameters the rates were indeed independent of the partial pressure of TMG and arsine, but a strong dependence on the deposition temperature and the TMG pulse length was found. However, self-limiting growth at a rate of one monolayer per cycle, i.e. ALE, was not found in this type of reactor. Also for UV (λ=193 nm) stimulated PFE at low temperature where otherwise no thermally-activated deposition occurs, self-limited growth involving one monolayer per cycle was not found.

  20. Nuclear reactors for research and radioisotope production in Argentina

    SciTech Connect

    Duran, H.H.

    1981-01-01

    In Argentina, the construction, operation, and use of research and radioisotope production reactors is and has been an important method of personnel preparation for the nuclear power program. Moreover, it is a very suitable means for technology transfer to countries developing their own nuclear programs. At present, the following research reactors are in operation in Argentina: Argentine Reactor 0 (RA-0); Argentine Reactor 1 (RA-1); Argentine Reactor 2 (RA-2); Argentine Reactor 3 (RA-3); Argentine Reactor 4 (RA-4). The Argentine Reactor 6 (RA-6), under construction, should reach criticality in 1981.

  1. (UA1 reactor fuels safety and performance)

    SciTech Connect

    Taleyarkhan, R.P.

    1990-07-13

    The traveler visited several reactor and hot cell experimental facilities connected with JAERI at the Oarai and Tokai establishments. Uranium silicide fission product release experimental data and related acquisition systems were discussed. A presentation was made by the traveler on analysis and modeling of fission product release from UAl reactor fuels. Data obtained by JAERI thus far were offered to the traveler for Oak Ridge National Laboratory (ORNL) review and analysis. This data confirmed key aspects of ORNL theoretical model predictions and will be useful for Advanced Neutron Source (ANS) design. The Oarai establishment expressed their interest and willingness to pursue ORNL/JAERI cooperative efforts in understanding volatile fission product release behavior from silicide fuels. The traveler also presented a perspective overview on ORNL severe accident analysis technology and identified areas for cooperation in JAERI's forthcoming transient testing program. JAERI staff presented plans for evaluating silicide fuel performance under transient reactivity insertion accident conditions in the Nuclear Safety Research Reactor (NSRR) facility. A surprise announcement was made concerning JAERI's most recent initiative relating to the construction of a safety demonstration reactor (SDR) at the Tokai site. The purpose of this reactor facility would be to demonstrate operational safety of both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs) in support of Japan's nuclear power industry.

  2. USA B and W`s IR-CFB coal-fired boiler operating experiences

    SciTech Connect

    Kavidass, S.; Maryamchik, M.; Kanoria, M.; Price, C.S.

    1998-12-31

    This paper updates operating experience of two Babcock and Wilcox (B and W) coal-fired, internal recirculation circulating fluidized-bed (IR-CFB) boilers. The first boiler is located at Southern Illinois University (SIU) in Carbondale, Illinois and is designed for 35 MWt output for cogeneration application, utilizing high sulfur, low ash Illinois coal. The second boiler is located at Kanoria Chemicals and Industries Ltd. (KCIL) in Renukoot, India and is designed for 81 MWt output for captive power requirements, firing high ash, low sulfur coal. This boiler was supplied by Thermax B and W (TBW) Ltd., a joint venture company of B and W and Thermax in India. The CFB technology is selected for these two units based on the fuel and environmental considerations. This paper discusses the various aspects of the two IR-CFB boilers` design features, performance, and operating experience including emissions.

  3. AGR-5/6/7 LEUCO Kernel Fabrication Readiness Review

    SciTech Connect

    Marshall, Douglas W.; Bailey, Kirk W.

    2015-02-01

    In preparation for forming low-enriched uranium carbide/oxide (LEUCO) fuel kernels for the Advanced Gas Reactor (AGR) fuel development and qualification program, Idaho National Laboratory conducted an operational readiness review of the Babcock & Wilcox Nuclear Operations Group – Lynchburg (B&W NOG-L) procedures, processes, and equipment from January 14 – January 16, 2015. The readiness review focused on requirements taken from the American Society Mechanical Engineers (ASME) Nuclear Quality Assurance Standard (NQA-1-2008, 1a-2009), a recent occurrence at the B&W NOG-L facility related to preparation of acid-deficient uranyl nitrate solution (ADUN), and a relook at concerns noted in a previous review. Topic areas open for the review were communicated to B&W NOG-L in advance of the on-site visit to facilitate the collection of objective evidences attesting to the state of readiness.

  4. Tdp studies and tests for C. A. Energia Electrica de Venezuela (enelven) at planta ramon laguna, units RL-17 and RL-10. Volume 1. Executive summary, RL-17 test report, and gas conversion proposals. Export trade information

    SciTech Connect

    Not Available

    1991-03-28

    The study, conducted by Babcock and Wilcox, was funded by the U.S. Trade and Development agency on behalf of Enelven. In order to maximize generated power output and minimize operating costs at Planta Ramon Laguna, tests were done to evaluate the condition of equipment at the plant. In order to identify any damage and determine the operating output of each unit, assessments were done of the furnaces, boilers, generators and boiler feed pumps being used in the plant. The report presents the results of these tests. This is the first of three volumes and it is divided into the following sections: (1) Executive Summary; (2) Hydrogen Damage Assessment; (3) RL-17 Gas Conversion Proposal; (4) RL-10 and RL-11 Gas Conversion Proposals.

  5. Tdp studies and tests for C. A. Energia Electrica de Venezuela (enelven) at planta ramon laguna, units RL-17 and RL-10. Volume 3. Unit RL-10 turbine generator condition assessment report and units RL-10 and RL-11 boiler feed pump conditon assessment report. Export trade information

    SciTech Connect

    Not Available

    1991-03-28

    The study, conducted by Babcock and Wilcox, was funded by the U.S. Trade and Development agency on behalf of Enelven. In order to maximize generated power output and minimize operating costs at Planta Ramon Laguna, tests were done to evaluate the condition of equipment at the plant. In order to identify any damage and determine the operating output of each unit, assessments were done of the furnaces, boilers, generators and boiler feed pumps being used in the plant. The report presents the results of these tests. This is the last of three volumes and it is divided into the following sections: (1) Condition Assessment of Unit RL-10 Turbine-Generator; (2) Condition Assessment of Unit RL-10 and RL-11 Boiler Feed Pumps.

  6. TVA commercial demonstration plant project. Volume 2. Basis of study assessments and project selection. Final report

    SciTech Connect

    Not Available

    1980-11-01

    The Tennessee Valley Authority (TVA) is considering the design, construction, and operation of a commercial scale coal gasification facility to produce a clean, medium Btu fuel gas (MBG). The project includes all process and support systems required to convert approximately 20,000 tons per day of Kentucky No. 9 bituminous coal, as fed to the gasifiers, into MBG equivalent to about 300 billion Btu per day. The first phase of the proposed project involves conceptual design, environmental and siting studies and economic analyses of commercial plants emphasizing the following gasification technologies: Babcock and Wilcox entrained flow gasifier, Lurgi dry ash gasifier, BGC/Lurgi slagging gasifier, Texaco entrained flow gasifier, and Koppers Totzek entrained flow gasifier. Foster Wheeler's study and assessments/process selection is summarized in this volume.

  7. Intermediate leak protection/automatic shutdown for B and W helical coil steam generator

    SciTech Connect

    Not Available

    1981-01-01

    The report summarizes a follow-on study to the multi-tiered Intermediate Leak/Automatic Shutdown System report. It makes the automatic shutdown system specific to the Babcock and Wilcox (B and W) helical coil steam generator and to the Large Development LMFBR Plant. Threshold leak criteria specific to this steam generator design are developed, and performance predictions are presented for a multi-tier intermediate leak, automatic shutdown system applied to this unit. Preliminary performance predictions for application to the helical coil steam generator were given in the referenced report; for the most part, these predictions have been confirmed. The importance of including a cover gas hydrogen meter in this unit is demonstrated by calculation of a response time one-fifth that of an in-sodium meter at hot standby and refueling conditions.

  8. Reactor Simulator Integration and Testing

    NASA Technical Reports Server (NTRS)

    Schoenfield, M. P.; Webster, K. L.; Pearson, J. B.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator (RxSim) test loop was designed and built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing were to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V because the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This Technical Memorandum summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained, which was lower than the predicted 750 K but 156 K higher than the cold temperature, indicating the design provided some heat regeneration. The annular linear induction pump tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  9. Environment of deposition of downdip Lower Wilcox sandstones, Provident City field, Lavaca County, Texas

    SciTech Connect

    Vest, S.W.

    1990-09-01

    The Lower Wilcox section at Provident City field produces dry gas from thin-bedded, silty sandstones, at depths of 12,500 to 14,100 ft (3,810 to 4,298 m). Cores show that sandstone cosets range 0.1 to 2.7 ft (0.03 to 0.82 m) and average 0.5 8 ft (0. 18 m) in thickness. Sedimentary structures within the cosets range upward from a massive unit (A) to a planar-laminated unit (B) to a ripple-laminated unit (C). The cosets have an average composition of lithic arkose and show textural grading indicative of deposition from turbidity flows. The sandstones lie within the Wilcox fault zone, downdip of the Colorado and Guadalupe deltas of the Rockdale Delta System. Regional stratigraphy and structural trends indicate that the sandstones were deposited in a deep marine environment. A growth fault, having approximately 1000 ft (3048 m) of throw at a depth of 12,300 ft (3750 m), bounds the field to the northwest and largely controls the distribution of lithofacies. Stacked, AB-type, turbidite cosets indicate channel facies. The M Sandstone was deposited as a constructional channel, with abrupt lateral grading to overbank facies, where turbidites of the BC- and C-type are dominant. The S Sandstone was deposited as a series of thin, constructional channels, mostly with turbidites of the AB- and ABC-type that are generally stacked, causing superimposed, dip-trending lobes on an otherwise strike-trending sandstone.

  10. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-01-20

    ... COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.'' DATES... developed using this Catalog along with the Operator Licensing Examination Standards for Power...

  11. Prism sodium-cooled reactor design and performance

    SciTech Connect

    Kwant, W.; Magee, P.M.; Patel, M.R. )

    1989-01-01

    The Power Reactor Inherently Safe Module (PRISM) program is being conducted at General Electric (GE) under U.S. Department of Energy sponsorship to develop a conceptual design for an advanced sodium-cooled liquid-metal reactor plant. The PRISM design emphasizes inherent safety, modular construction, and factory fabrication. A PRISM power plant includes a number of reactor modules, which will be fabricated in a factory and shipped by whatever combination of barge, rail, and road transport that is most economical for a particular site. The target commercial PRISM plant utilizes nine reactor modules arranged in three identical 465-MW(electric) power blocks for an overall plant net electrical rating of 1395 MW(electric). Each power block has three identical reactor modules, each with its own steam generator, that jointly supply saturated steam to a single turbine generator. The PRISM's features of fewer and simpler safety systems, seismic isolation, passive decay heat removal, inherent reactivity control, and generous margins from structural and fuel damage limits during potential accident situations will result in significant gains in public safety and protection of the owner's investment. The use of standardized modular construction and extensive factory fabrication is resulting in a plant design that is economically competitive against projected coal plants and other nuclear design approaches.

  12. NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

    DOEpatents

    Moore, W.T.

    1958-09-01

    This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

  13. TRAC PF1/MOD1 calculations and data comparisons for mist feed and bleed and steam generator tube rupture experiments

    SciTech Connect

    Siebe, D.A.; Boyack, B.E.; Steiner, J.L.

    1988-01-01

    Los Alamos National Laboratory is a participant in the Integral System Test (IST) program initiated in June 1983 for the purpose of providing integral system test data on specific issues/phenomena relevant to post-small-break loss-of-coolant accidents, loss of feedwater and other transients in Babcock and Wilcox (BandW) plant designs. The Multi-Loop Integral System Test (MIST) facility is the largest single component in the IST program. MIST is a 2 /times/ 4 (two hot legs and steam generators (SGs), four cold legs and reactor coolant pumps) representation of lowered-loop reactor system of the BandW design. It is a full-height, full-pressure facility with 1/817 power and volume scaling. Two other integral experimental facilities are included in the IST program: test loops at the University of Maryland, College Park, and at SRI International (SRI-2). The objective of the IST tests is to generate high-quality experimental data to be used for assessing thermal-hydraulic safety computer codes. Efforts are under way at Los Alamos to assess TRAC-PF1/MOD1 against data from each of the IST facilities. Calculations and data comparisons for TRAC-PF1/MOD1 assessment are presented for two transients run in the MIST facility. These are MIST Test 330302, a feed and bleed test with delayed high-pressure injection; and Test 3404AA, an SG tube-rupture test with the affected SG isolated. Only MIST assessment results are presented in this paper. The TRAC-PF1/MOD1 calculations completed to date for MIST tests are in reasonable agreement with the data from these tests. Reasonable agreement is defined as meaning that major trends are predicted correctly, although TRAC values are frequently outside the range of data uncertainty. We believe that correct conclusions will be reached if the code is used in similar applications despite minor code/model deficiencies. 7 refs., 5 figs., 2 tabs.

  14. Developments and Tendencies in Fission Reactor Concepts

    NASA Astrophysics Data System (ADS)

    Adamov, E. O.; Fuji-Ie, Y.

    This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC

  15. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors

    NASA Astrophysics Data System (ADS)

    Recktenwald, Geoff; Deinert, Mark

    2010-03-01

    Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks.

  16. FBR and RBR particle bed space reactors

    SciTech Connect

    Powell, J.R.; Botts, T.E.

    1983-01-01

    Compact, high-performance nuclear reactor designs based on High-Temperature Gas Reactors (HTGRs) particulate fuel are investigated. The large surface area available with the small-diameter (approx. 500 microns) particulate fuel allows very high power densities (MW's/liter), small temperature differences between fuel and coolant (approx. 10/sup 0/K), high coolant-outlet temperatures (1500 to 3000/sup 0/K, depending on design), and fast reactor startup (approx. 2 to 3 seconds). Two reactor concepts are developed - the Fixed Bed Reactor (FBR), where the fuel particles are packed into a thin annular bed between two porous cylindrical drums, and the Rotating Bed Reactor (RBR), where the fuel particles are held inside a cold rotating (typically approx. 500 rpm) porous cylindrical drum. The FBR can operate steady-state in the closed-cycle He-cooled mode or in the open-cycle H/sub 2/-cooled mode. The RBR will operate only in the open-cycle H/sub 2/-cooled mode.

  17. Process regime variability across growth faults in the Paleogene Lower Wilcox Guadalupe Delta, South Texas Gulf Coast

    NASA Astrophysics Data System (ADS)

    Olariu, Mariana I.; Ambrose, William A.

    2016-07-01

    The Wilcox Group in Texas is a 3000 m thick unit of clastic sediments deposited along the Gulf of Mexico coast during early Paleogene. This study integrates core facies analysis with subsurface well-log correlation to document the sedimentology and stratigraphy of the Lower Wilcox Guadalupe Delta. Core descriptions indicate a transition from wave- and tidally-influenced to wave-dominated deposition. Upward-coarsening facies successions contain current ripples, organic matter, low trace fossil abundance and low diversity, which suggest deposition in a fluvial prodelta to delta front environment. Heterolithic stratification with lenticular, wavy and flaser bedding indicate tidal influence. Pervasively bioturbated sandy mudstones and muddy sandstones with Cruziana ichnofacies and structureless sandstones with Ophiomorpha record deposition in wave-influenced deltas. Tidal channels truncate delta front deposits and display gradational upward-fining facies successions with basal lags and sandy tabular cross-beds passing into heterolithic tidal flats and biologically homogenized mudstones. Growth faults within the lower Wilcox control expanded thickness of sedimentary units (up to 4 times) on the downdip sides of faults. Increased local accommodation due to fault subsidence favors a stronger wave regime on the outer shelf due to unrestricted fetch and water depth. As the shoreline advances during deltaic progradation, successively more sediment is deposited in the downthrown depocenters and reworked along shore by wave processes, resulting in a thick sedimentary unit characterized by repeated stacking of shoreface sequences. Thick and laterally continuous clean sandstone successions in the downthrown compartments represent attractive hydrocarbon reservoirs. As a consequence of the wave dominance and increased accommodation, thick (tens of meters) sandstone-bodies with increased homogeneity and vertical permeability within the stacked shoreface successions are created.

  18. Improvement of transport-corrected scattering stability and performance using a Jacobi inscatter algorithm for 2D-MOC

    DOE PAGES

    Stimpson, Shane; Collins, Benjamin; Kochunas, Brendan

    2017-03-10

    The MPACT code, being developed collaboratively by the University of Michigan and Oak Ridge National Laboratory, is the primary deterministic neutron transport solver being deployed within the Virtual Environment for Reactor Applications (VERA) as part of the Consortium for Advanced Simulation of Light Water Reactors (CASL). In many applications of the MPACT code, transport-corrected scattering has proven to be an obstacle in terms of stability, and considerable effort has been made to try to resolve the convergence issues that arise from it. Most of the convergence problems seem related to the transport-corrected cross sections, particularly when used in the 2Dmore » method of characteristics (MOC) solver, which is the focus of this work. Here in this paper, the stability and performance of the 2-D MOC solver in MPACT is evaluated for two iteration schemes: Gauss-Seidel and Jacobi. With the Gauss-Seidel approach, as the MOC solver loops over groups, it uses the flux solution from the previous group to construct the inscatter source for the next group. Alternatively, the Jacobi approach uses only the fluxes from the previous outer iteration to determine the inscatter source for each group. Consequently for the Jacobi iteration, the loop over groups can be moved from the outermost loop$-$as is the case with the Gauss-Seidel sweeper$-$to the innermost loop, allowing for a substantial increase in efficiency by minimizing the overhead of retrieving segment, region, and surface index information from the ray tracing data. Several test problems are assessed: (1) Babcock & Wilcox 1810 Core I, (2) Dimple S01A-Sq, (3) VERA Progression Problem 5a, and (4) VERA Problem 2a. The Jacobi iteration exhibits better stability than Gauss-Seidel, allowing for converged solutions to be obtained over a much wider range of iteration control parameters. Additionally, the MOC solve time with the Jacobi approach is roughly 2.0-2.5× faster per sweep. While the performance and stability of

  19. Optimally moderated nuclear fission reactor and fuel source therefor

    DOEpatents

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

    2008-07-22

    An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

  20. Minutes of Workshop on Composite Materials Test Methods Held at the Institute for Defense Analyses on February 27, 1985.

    DTIC Science & Technology

    1985-04-01

    Appendix 4). NDE was used to character- ize the panels; void content of 2% maximum was specified for the test panels. The publication of this...HEXCEL (415) 823-4200 CHANNON, STAN IDA (703) 845-2256 (714) 683-7357 CLEVINGER, Gary Babcock & Wilcox (804) 522-5286 DAVIS, JOHN G., JR NASA-LANGLEY...suppliers, contractors and ASNT. Yes Yes Yes Only from a Standardization/Documentation standpoint. No Yes Not to define "the" method since NDE is

  1. Cross-flow electrochemical reactor cells, cross-flow reactors, and use of cross-flow reactors for oxidation reactions

    DOEpatents

    Balachandran, Uthamalingam; Poeppel, Roger B.; Kleefisch, Mark S.; Kobylinski, Thaddeus P.; Udovich, Carl A.

    1994-01-01

    This invention discloses cross-flow electrochemical reactor cells containing oxygen permeable materials which have both electron conductivity and oxygen ion conductivity, cross-flow reactors, and electrochemical processes using cross-flow reactor cells having oxygen permeable monolithic cores to control and facilitate transport of oxygen from an oxygen-containing gas stream to oxidation reactions of organic compounds in another gas stream. These cross-flow electrochemical reactors comprise a hollow ceramic blade positioned across a gas stream flow or a stack of crossed hollow ceramic blades containing a channel or channels for flow of gas streams. Each channel has at least one channel wall disposed between a channel and a portion of an outer surface of the ceramic blade, or a common wall with adjacent blades in a stack comprising a gas-impervious mixed metal oxide material of a perovskite structure having electron conductivity and oxygen ion conductivity. The invention includes reactors comprising first and second zones seprated by gas-impervious mixed metal oxide material material having electron conductivity and oxygen ion conductivity. Prefered gas-impervious materials comprise at least one mixed metal oxide having a perovskite structure or perovskite-like structure. The invention includes, also, oxidation processes controlled by using these electrochemical reactors, and these reactions do not require an external source of electrical potential or any external electric circuit for oxidation to proceed.

  2. MEANS FOR SHIELDING AND COOLING REACTORS

    DOEpatents

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

    1959-02-10

    Reactors of the water-cooled type and a means for shielding such a rcactor to protect operating personnel from harmful radiation are discussed. In this reactor coolant tubes which contain the fissionable material extend vertically through a mass of moderator. Liquid coolant enters through the bottom of the coolant tubes and passes upwardly over the fissionable material. A shield tank is disposed over the top of the reactor and communicates through its bottom with the upper end of the coolant tubes. A hydrocarbon shielding fluid floats on the coolant within the shield tank. With this arrangements the upper face of the reactor can be opened to the atmosphere through the two superimposed liquid layers. A principal feature of the invention is that in the event radioactive fission products enter thc coolant stream. imposed layer of hydrocarbon reduces the intense radioactivity introduced into the layer over the reactors and permits removal of the offending fuel material by personnel shielded by the uncontaminated hydrocarbon layer.

  3. Instrumentation, Monitoring and NDE for New Fast Reactors

    SciTech Connect

    Bond, Leonard J.; Doctor, Steven R.; Bunch, Kyle J.; Good, Morris S.; Waltar, Alan E.

    2007-07-28

    The Global Nuclear Energy Partnership (GNEP) has been proposed as a viable system in which to close the fuel cycle in a manner consistent with markedly expanding the global role of nuclear power while significantly reducing proliferation risks. A key part of this system relies on the development of actinide transmutation, which can only be effectively accomplished in a fast-spectrum reactor. The fundamental physics for fast reactors is well established. However, to achieve higher standards of safety and reliability, operate with longer intervals between outages, and achieve high operating capacity factors, new instrumentation and on-line monitoring capabilities will be required--during both fabrication and operation. Since the Fast Flux Test Facility (FFTF) and Experimental Breeder Reactor – II (EBR-II) reactors were operational in the USA, there have been major advances in instrumentation, not the least being the move to digital systems. Some specific capabilities have been developed outside the USA, but new or at least re-established capabilities will be required. In many cases the only available information is in reports and papers. New and improved sensors and instrumentation will be required. Advanced instrumentation has been developed for high-temperature/high-flux conditions in some cases, but most of the original researchers and manufacturers are retired or no longer in business.

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

    SciTech Connect

    Wittenbrock, N. G.

    1982-01-01

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

  5. Reactor performances and microbial communities of biogas reactors: effects of inoculum sources.

    PubMed

    Han, Sheng; Liu, Yafeng; Zhang, Shicheng; Luo, Gang

    2016-01-01

    Anaerobic digestion is a very complex process that is mediated by various microorganisms, and the understanding of the microbial community assembly and its corresponding function is critical in order to better control the anaerobic process. The present study investigated the effect of different inocula on the microbial community assembly in biogas reactors treating cellulose with various inocula, and three parallel biogas reactors with the same inoculum were also operated in order to reveal the reproducibility of both microbial communities and functions of the biogas reactors. The results showed that the biogas production, volatile fatty acid (VFA) concentrations, and pH were different for the biogas reactors with different inocula, and different steady-state microbial community patterns were also obtained in different biogas reactors as reflected by Bray-Curtis similarity matrices and taxonomic classification. It indicated that inoculum played an important role in shaping the microbial communities of biogas reactor in the present study, and the microbial community assembly in biogas reactor did not follow the niche-based ecology theory. Furthermore, it was found that the microbial communities and reactor performances of parallel biogas reactors with the same inoculum were different, which could be explained by the neutral-based ecology theory and stochastic factors should played important roles in the microbial community assembly in the biogas reactors. The Bray-Curtis similarity matrices analysis suggested that inoculum affected more on the microbial community assembly compared to stochastic factors, since the samples with different inocula had lower similarity (10-20 %) compared to the samples from the parallel biogas reactors (30 %).

  6. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    SciTech Connect

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  7. Design of a weapons-grade plutonium assembly for optimal burnup in a standard pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Alonso-Vargas, Gustavo

    We created a new MOX fuel assembly design that can be used in standard Westinghouse pressurized water reactors (PWR) to maximize the plutonium throughput while introducing the lowest perturbation possible to the control and safety systems of the reactor. Our assembly design, which is called MIX-33, appears to be a good option for the disposition of weapons-grade plutonium (WG-Pu), increasing the plutonium disposition rate by 8% compared to a previous Westinghouse design. It is based in two novel ideas: the use of both uranium and plutonium fuel pins in the same assembly, and the increase of the moderation ratio of the assembly. We replaced 8 fuel pins by water holes to increase the moderation ratio. We can transition smoothly from a full LEU core to a full MIX-33 core meeting the operational and safety regulations of a standard PWR. Given a MOX supply interruption scenario we can transition smoothly to full LEU meeting the safety regulations and using standard LEU assemblies with uniform enriched pin-wise distribution. If the MOX supply is interrupted for only one cycle, we are able to transition back to full MIX-33 core. However, in this case we probably need to de-rate the power by a few percent for a few weeks at the beginning of the cycle (BOC) to accommodate high peaking. For comparison we created another assembly design without extra water holes, which we called "MIX-25". It behaves in all the conditions analyzed in a similar way to the MIX-33 but it does present minor control problems. These can be solved by making small modifications to the control and safety systems, namely by enriching the boron-10 content of some boron absorbers. Thus, the addition of water holes replacing fuel pins helps to improve the MIX-33 performance and eliminate the difficulties seen in the MIX-25 design. We also performed a benchmarking analysis to test the code CASMO-3 to analyze WG-Pu assemblies, using the code MCNP-4A to compare. We found good agreement between CASMO-3 and

  8. Geoneutrinos and reactor antineutrinos at SNO+

    NASA Astrophysics Data System (ADS)

    Baldoncini, M.; Strati, V.; Wipperfurth, S. A.; Fiorentini, G.; Mantovani, F.; McDonough, W. F.; Ricci, B.

    2016-05-01

    In the heart of the Creighton Mine near Sudbury (Canada), the SNO+ detector is foreseen to observe almost in equal proportion electron antineutrinos produced by U and Th in the Earth and by nuclear reactors. SNO+ will be the first long baseline experiment to measure a reactor signal dominated by CANDU cores (~55% of the total reactor signal), which generally burn natural uranium. Approximately 18% of the total geoneutrino signal is generated by the U and Th present in the rocks of the Huronian Supergroup-Sudbury Basin: the 60% uncertainty on the signal produced by this lithologic unit plays a crucial role on the discrimination power on the mantle signal as well as on the geoneutrino spectral shape reconstruction, which can in principle provide a direct measurement of the Th/U ratio in the Earth.

  9. NEUTRONIC REACTOR SHIELD AND SPACER CONSTRUCTION

    DOEpatents

    Wigner, E.P.; Ohlinger, L.A.

    1958-11-18

    Reactors of the heterogeneous, graphite moderated, fluid cooled type and shielding and spacing plugs for the coolant channels thereof are reported. In this design, the coolant passages extend horizontally through the moderator structure, accommodating the fuel elements in abutting end-to-end relationship, and have access openings through the outer shield at one face of the reactor to facilitate loading of the fuel elements. In the outer ends of the channels which extend through the shields are provided spacers and shielding plugs designed to offer minimal reslstance to coolant fluid flow while preventing emanation of harmful radiation through the access openings when closed between loadings.

  10. Fast Reactors

    NASA Astrophysics Data System (ADS)

    Esposito, S.; Pisanti, O.

    The following sections are included: * Elementary Considerations * The Integral Equation to the Neutron Distribution * The Critical Size for a Fast Reactor * Supercritical Reactors * Problems and Exercises

  11. Distributed computing and nuclear reactor analysis

    SciTech Connect

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

    1994-03-01

    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.

  12. Techniques for in-service inspection of heat-transfer tubes in steam generators

    SciTech Connect

    McClung, R.W.; Day, R.A.; Neely, H.H.; Powers, T.

    1981-01-01

    A multifaceted development program is in progress in the United States to study techniques for in-service inspection (ISI) of heat transfer tubes in breeder reactor steam generators. Several steam generator designs are involved. Although there are some similarities in the approaches, many of the details of techniques and capabilities are specific to the steam generator design. This paper describes the ultrasonic, eddy-current and penetrating radiation techniques being studied for the various steam generators, including the Large Leak Test Rig, the Clinch River Breeder Reactor design, and alternate steam generators being developed by Westinghouse and Babcock and Wilcox.

  13. Standards and Certification. Symposium.

    ERIC Educational Resources Information Center

    2002

    This document contains three papers from a symposium on standards and certification in human resource development (HRD). "Implementing Management Standards in the UK" (Jonathan Winterton, Ruth Winterton) reports on a study that explored the implementation of management standards in 16 organizations and identified 36 key themes and…

  14. Light water reactor program

    SciTech Connect

    Franks, S.M.

    1994-12-31

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

  15. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    SciTech Connect

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  16. Hydrogen and water reactor safety: proceedings

    SciTech Connect

    Not Available

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

  17. Safety of evolutionary and innovative nuclear reactors: IAEA activities and world efforts

    SciTech Connect

    Saito, T.; Gasparini, M.

    2004-07-01

    'Defence in Depth' approach constitutes the basis of the IAEA safety standards for nuclear power plants. Lessons learned from the current generation of reactors suggest that, for the next generation of reactor designs, the Defence in Depth philosophy should be retained, and that its implementation should be guided by the probabilistic insights. Recent developments in the area of general safety requirements based on Defence in Depth approach are examined and summarized. Global efforts to harmonize safety requirements for evolutionary nuclear power plants have involved many countries and organizations such as IAEA, US EPRI and European Utility EUR Organization. In recent years, developments of innovative nuclear power plants are also being discussed. The IAEA is currently developing a safety approach specifically for innovative nuclear reactors. This approach will eventually lead to a proposal of safety requirements for innovative reactors. Such activities related to safety requirements of evolutionary and innovative reactors are introduced. Various evolutionary and innovative reactor designs are reported in the world. The safety design features of evolutionary large LWRs, innovative LWRs, Modular High Temperature Gas Reactors and Small Liquid Metal Cooled LMRs are also introduced. Enhanced safety features proposed in such reactors are discussed and summarized according to the levels of Defence in Depth. For future nuclear plants, international cooperation and harmonization, especially in the area of safety, appear to be inevitable. Based on the past experience with many member states, the IAEA believes itself to be the uniquely positioned international organization to play this key role. (authors)

  18. Fuels for research and test reactors, status review: July 1982

    SciTech Connect

    Stahl, D.

    1982-12-01

    A thorough review is provided on nuclear fuels for steady-state thermal research and test reactors. The review was conducted to provide a documented data base in support of recent advances in research and test reactor fuel development, manufacture, and demonstration in response to current US policy on availability of enriched uranium. The review covers current fabrication practice, fabrication development efforts, irradiation performance, and properties affecting fuel utilization, including thermal conductivity, specific heat, density, thermal expansion, corrosion, phase stability, mechanical properties, and fission-product release. The emphasis is on US activities, but major work in Europe and elsewhere is included. The standard fuel types discussed are the U-Al alloy, UZrH/sub x/, and UO/sub 2/ rod fuels. Among new fuels, those given major emphasis include H/sub 3/Si-Al dispersion and UO/sub 2/ caramel plate fuels.

  19. Current Abstracts Nuclear Reactors and Technology

    SciTech Connect

    Bales, J.D.; Hicks, S.C.

    1993-01-01

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

  20. Summary of advanced LMR (Liquid Metal Reactor) evaluations: PRISM (Power Reactor Inherently Safe Module) and SAFR (Sodium Advanced Fast Reactor)

    SciTech Connect

    Van Tuyle, G.J.; Slovik, G.C.; Chan, B.C.; Kennett, R.J.; Cheng, H.S.; Kroeger, P.G. )

    1989-10-01

    In support of the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) has performed independent analyses of two advanced Liquid Metal Reactor (LMR) concepts. The designs, sponsored by the US Department of Energy (DOE), the Power Reactor Inherently Safe Module (PRISM) (Berglund, 1987) and the Sodium Advanced Fast Reactor (SAFR) (Baumeister, 1987), were developed primarily by General Electric (GE) and Rockwell International (RI), respectively. Technical support was provided to DOE, RI, and GE, by the Argonne National Laboratory (ANL), particularly with respect to the characteristics of the metal fuels. There are several examples in both PRISM and SAFR where inherent or passive systems provide for a safe response to off-normal conditions. This is in contrast to the engineered safety systems utilized on current US Light Water Reactor (LWR) designs. One important design inherency in the LMRs is the inherent shutdown'', which refers to the tendency of the reactor to transition to a much lower power level whenever temperatures rise significantly. This type of behavior was demonstrated in a series of unscrammed tests at EBR-II (NED, 1986). The second key design feature is the passive air cooling of the vessel to remove decay heat. These systems, designated RVACS in PRISM and RACS in SAFR, always operate and are believed to be able to prevent core damage in the event that no other means of heat removal is available. 27 refs., 78 figs., 3 tabs.

  1. Performance and reliability monitoring of advanced reactors

    SciTech Connect

    Robinson, D. G.

    2006-07-01

    Advanced nuclear reactor designs must be designed for long core life and high fuel burn-up. This will pose new challenges for monitoring the state of the highly integrated and largely inaccessible components within the primary reactor vessel. As reactors age, characterizing and predicting the internal operational performance (e.g. vibration/flow characteristics) will be critical for determining maintenance strategies and early identification of safety and reliability issues. This paper discusses a current research program into a new approach for monitoring that provides the capability to combine information from in-situ monitoring sensors within a reactor with maintenance and repair histories to characterize the current probability of a failure event. The proposed approach suggests the application of Bayesian concepts similar to that used in medical monitoring. System-based medical monitoring utilizes hierarchical Bayesian methods such as Markov Chain Monte Carlo (MCMC) and particle filtering methods to deliver estimates of the state of health of a process based on data taken from the process in real-time. Preliminary results from the application of the methodology to the monitoring of corrosion in electronics is presented. (authors)

  2. Research reactor de-fueling and fuel shipment

    SciTech Connect

    Ice, R.D.; Jawdeh, E.; Strydom, J.

    1998-08-01

    Planning for the Georgia Institute of Technology Research Reactor operations during the 1996 Summer Olympic Games began in early 1995. Before any details could be outlined, several preliminary administrative decisions had to be agreed upon by state, city, and university officials. The two major administrative decisions involving the reactor were (1) the security level and requirements and (2) the fuel status of the reactor. The Georgia Tech Research Reactor (GTRR) was a heavy-water moderated and cooled reactor, fueled with high-enriched uranium. The reactor was first licensed in 1964 with an engineered lifetime of thirty years. The reactor was intended for use in research applications and as a teaching facility for nuclear engineering students and reactor operators. Approximately one year prior to the olympics, the Georgia Tech administration decided that the GTRR fuel would be removed. In addition, a heightened, beyond regulatory requirements, security system was to be implemented. This report describes the scheduling, operations, and procedures.

  3. Reactor Physics Methods and Analysis Capabilities in SCALE

    SciTech Connect

    DeHart, Mark D; Bowman, Stephen M

    2011-01-01

    The TRITON sequence of the SCALE code system provides a powerful, robust, and rigorous approach for performing reactor physics analysis. This paper presents a detailed description of TRITON in terms of its key components used in reactor calculations. The ability to accurately predict the nuclide composition of depleted reactor fuel is important in a wide variety of applications. These applications include, but are not limited to, the design, licensing, and operation of commercial/research reactors and spent-fuel transport/storage systems. New complex design projects such as next-generation power reactors and space reactors require new high-fidelity physics methods, such as those available in SCALE/TRITON, that accurately represent the physics associated with both evolutionary and revolutionary reactor concepts as they depart from traditional and well-understood light water reactor designs.

  4. Reactor Physics Methods and Analysis Capabilities in SCALE

    SciTech Connect

    Mark D. DeHart; Stephen M. Bowman

    2011-05-01

    The TRITON sequence of the SCALE code system provides a powerful, robust, and rigorous approach for performing reactor physics analysis. This paper presents a detailed description of TRITON in terms of its key components used in reactor calculations. The ability to accurately predict the nuclide composition of depleted reactor fuel is important in a wide variety of applications. These applications include, but are not limited to, the design, licensing, and operation of commercial/research reactors and spent-fuel transport/storage systems. New complex design projects such as next-generation power reactors and space reactors require new high-fidelity physics methods, such as those available in SCALE/TRITON, that accurately represent the physics associated with both evolutionary and revolutionary reactor concepts as they depart from traditional and well-understood light water reactor designs.

  5. Reactor dosimetry and RPV life management

    SciTech Connect

    Belousov, S.; Ilieva, K.; Mitev, M.

    2011-07-01

    Reactor dosimetry (RD) is a tool that provides data for neutron fluence accumulated over the reactor pressure vessel (RPV) during the reactor operation. This information, however, is not sufficient for RPV lifetime assessment. The life management of RPV is a multidisciplinary task. To assess whether the RPV steel properties at the current stage (for actual accumulated neutron fluence) of reactor operation are still 'safe enough,' the dependence of material properties on the fluence must be known; this is a task for material science (MS). Moreover, the mechanical loading over the RPV during normal operation and accidence have to be known, as well, for evaluation, if the RPV material integrity in this loading condition and existing cracks is provided. The crack loading path in terms of stress intensity factor is carried out by structural analyses (SA). Pressure and temperature distribution over RPV used in these analyses are obtained from a thermal hydraulic (TH) calculation. The conjunction of RD and other disciplines in RPV integrity assessment is analyzed in accordance with the FFP (fitness for purpose) approach. It could help to improve the efficiency in multi-disciplinary tasks solutions. (authors)

  6. AIR DUCTS STAND NEXT TO (AND OUTSIDE OF) REACTOR CABINET ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    AIR DUCTS STAND NEXT TO (AND OUTSIDE OF) REACTOR CABINET AT THE SOUTHWEST AND NORTHEAST CORNERS OF THE REACTOR'S THERMAL SHIELD. THEY WILL BE ENVELOPED IN BIOLOGICAL CONCRETE SHIELD. IN THE SUB-BASEMENT, THE TWO DUCTS WILL JOIN TOGETHER AND EXIT THE BUILDING TO THE FAN HOUSE. CAMERA FACING NORTH. INL NEGATIVE NO. 1625. Unknown Photographer, 3/6/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  7. Reactor Coolant Pump seal issues and their applicability to new reactor designs

    SciTech Connect

    Ruger, C.J.; Higgins, J.C.

    1993-11-01

    Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970`s and early 1980`s raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants.

  8. REACTOR CORE SURROUNDED BY BERYLLIUM MODERATOR. CAMERA LOOKS DOWN AND ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    REACTOR CORE SURROUNDED BY BERYLLIUM MODERATOR. CAMERA LOOKS DOWN AND TOWARD NORTH INTO LOWER GRID CASTING. HOLES OF VARIOUS SIZES ACCOMMODATE COOLANT WATER AND EXPERIMENTAL POSITIONS. INL NEGATIVE NO. 4197. Unknown Photographer, 2/11/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  9. R- AND P- REACTOR BUILDING IN-SITU DECOMISSIONING VISUALIZATION

    SciTech Connect

    Bobbitt, J.; Vrettos, N.; Howard, M.

    2010-06-15

    During the early 1950s, five production reactor facilities were built at the Savannah River Site. These facilities were built to produce materials to support the building of the nation's nuclear weapons stockpile in response to the Cold War. R-Reactor and P-Reactor were the first two facilities completed in 1953 and 1954.

  10. Offsite dose calculation manual guidance: Standard radiological effluent controls for boiling water reactors

    SciTech Connect

    Meinke, W.W.; Essig, T.H.

    1991-04-01

    This report contains guidance which may be voluntarily used by licensees who choose to implement the provision of Generic Letter 89-- 01, which allows Radiological Effluent Technical Specifications (RETS) to be removed from the main body of the Technical Specifications and placed in the Offsite Dose Calculation Manual (ODCM). Guidance is provided for Standard Effluent Controls definitions, Controls for effluent monitoring instrumentation, Controls for effluent releases, Controls for radiological environmental monitoring, and the basis for Controls. Guidance on the formulation of RETS has been available in draft form for a number of years; the current effort simply recasts those RETS into Standard Radiological Effluent Controls for application to the ODCM. 11 tabs.

  11. NEUTRONIC REACTOR

    DOEpatents

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

    1958-09-16

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

  12. JANUS reactor d and d project.

    SciTech Connect

    Fellhauer, C. R.

    1998-02-16

    Argonne National Laboratory (ANL-E) has recently completed the decontamination and decommissioning (D and D) of the JANUS Reactor Facility located in Building 202. The 200 KW reactor operated from August 1963 to March 1992. The facility was used to study the effects of both high and low doses of fission neutrons in animals. There were two exposure rooms on opposite sides of the reactor and the reactor was therefore named after the two-faced Roman god. The High Dose Room was capable of specimen exposure at a dose rate of 3,600 rads per hour. During calendar year 1996 a detailed characterization of the facility was performed by ANL-E Health Physics personnel. ANL-E Analytical Services performed the required sample analysis. An Auditable Safety Analysis and an Environmental Assessment were completed. D and D plans, procedures and procurement documents were prepared and approved. A D and D subcontractor was selected and a firm, fixed price contract awarded for the field work and final survey effort. The D and D subcontractor was mobilized to ANL-E in January 1997. Electrical isolation of all reactor equipment and control panels was accomplished and the equipment removed. A total of 207,230 pounds (94,082 Kg) of lead shielding was removed, surveyed and sampled, and free-released for recycle. All primary and secondary piping was removed, size reduced and packaged for disposal or recycled as appropriate. The reactor vessel was removed, sized reduced and packaged as radioactive waste in April. The activated graphite block reflector was removed next, followed by the bioshield concrete and steel. All of this material was packaged as low level waste. Total low level radioactive waste generation was 4002.1 cubic feet (113.3 cubic meters). Mixed waste generation was 538 cubic feet (15.2 cubic meters). The Final Release Survey was completed in September. The project field work was completed in 38 weeks without any lost-time accidents, personnel contaminations or unplanned

  13. Integral reactor system and method for fuel cells

    DOEpatents

    Fernandes, Neil Edward; Brown, Michael S; Cheekatamarla, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F

    2013-11-19

    A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

  14. Integral reactor system and method for fuel cells

    DOEpatents

    Fernandes, Neil Edward; Brown, Michael S.; Cheekatamaria, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F.

    2017-03-07

    A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert higher hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

  15. Rapid and Enhanced Proteolytic Digestion using Electric-Field-oriented Enzyme Reactor

    PubMed Central

    Zhou, Yu; Yi, Tie; Park, Sung-Soo; Chadwick, Wayne; Shen, Rong-Fong; Wu, Wells W.; Martin, Bronwen; Maudsley, Stuart

    2011-01-01

    We have created a novel enzyme reactor using electric field-mediated orientation and immobilization of proteolytic enzymes (trypsin/chymotrypsin) on biocompatible PVDF membranes in a continuous flow-through chamber. Using less than 5 minutes, this reactor in various enzyme combinations can produce enhanced rapid digestion for standardized prototypic proteins, hydrophilic proteins and hydrophobic transmembrane proteins when compared to in-solution techniques. With improved digestive efficiency, our reactor improved the overall functional analysis of lipid raft proteomes by identifying more closely functionally linked proteins and elucidated a richer set of biological processes and pathways linked to the proteins than traditional in-solution methods. PMID:21338726

  16. Standards and Administration.

    ERIC Educational Resources Information Center

    Gross, S. P.

    1978-01-01

    Presents a literature review of water quality standards and administration, covering publications of 1976-77. Consideration is given to municipal facilities, National Pollutant Discharge Elimination Systems, regional and international water quality management, and effluent standards. A list of 99 references is also presented. (HM)

  17. Thermal and neutron-physical features of the nuclear reactor for a power pulsation plant for space applications

    NASA Astrophysics Data System (ADS)

    Gordeev, É. G.; Kaminskii, A. S.; Konyukhov, G. V.; Pavshuk, V. A.; Turbina, T. A.

    2012-05-01

    We have explored the possibility of creating small-size reactors with a high power output with the provision of thermal stability and nuclear safety under standard operating conditions and in emergency situations. The neutron-physical features of such a reactor have been considered and variants of its designs preserving the main principles and approaches of nuclear rocket engine technology are presented.

  18. Nuclear reactor flow control method and apparatus

    DOEpatents

    Church, John P.

    1993-01-01

    Method and apparatus for improving coolant flow in a nuclear reactor during accident as well as nominal conditions. The reactor has a plurality of fuel elements in sleeves and a plenum above the fuel and through which the sleeves penetrate. Holes are provided in the sleeve so that coolant from the plenum can enter the sleeve and cool the fuel. The number and size of the holes are varied from sleeve to sleeve with the number and size of holes being greater for sleeves toward the center of the core and less for sleeves toward the periphery of the core. Preferably the holes are all the same diameter and arranged in rows and columns, the rows starting from the bottom of every sleeve and fewer rows in peripheral sleeves and more rows in the central sleeves.

  19. Nuclear reactor flow control method and apparatus

    DOEpatents

    Church, J.P.

    1993-03-30

    Method and apparatus for improving coolant flow in a nuclear reactor during accident as well as nominal conditions. The reactor has a plurality of fuel elements in sleeves and a plenum above the fuel and through which the sleeves penetrate. Holes are provided in the sleeve so that coolant from the plenum can enter the sleeve and cool the fuel. The number and size of the holes are varied from sleeve to sleeve with the number and size of holes being greater for sleeves toward the center of the core and less for sleeves toward the periphery of the core. Preferably the holes are all the same diameter and arranged in rows and columns, the rows starting from the bottom of every sleeve and fewer rows in peripheral sleeves and more rows in the central sleeves.

  20. The role of inoculum and reactor configuration for microbial community composition and dynamics in mainstream partial nitritation anammox reactors.

    PubMed

    Agrawal, Shelesh; Karst, Søren M; Gilbert, Eva M; Horn, Harald; Nielsen, Per H; Lackner, Susanne

    2017-03-10

    Implementation of partial nitritation anammox (PNA) in the mainstream (municipal wastewater treatment) is still under investigation. Microbial community structure and reactor type can influence the performance of PNA reactor; yet, little is known about the role of the community composition of the inoculum and the reactor configuration under mainstream conditions. Therefore, this study investigated the community structure of inocula of different origin and their consecutive community dynamics in four different lab-scale PNA reactors with 16S rRNA gene amplicon sequencing. These reactors were operated for almost 1 year and subjected to realistic seasonal temperature fluctuations as in moderate climate regions, that is, from 20°C in summer to 10°C in winter. The sequencing analysis revealed that the bacterial community in the reactors comprised: (1) a nitrifying community (consisting of anaerobic ammonium-oxidizing bacteria (AnAOB), ammonia-oxidizing bacteria (AOB), and nitrite-oxidizing bacteria (NOB)); (2) different heterotrophic denitrifying bacteria and other putative heterotrophic bacteria (HB). The nitrifying community was the same in all four reactors at the genus level, although the biomasses were of different origin. Community dynamics revealed a stable community in the moving bed biofilm reactors (MBBR) in contrast to the sequencing batch reactors (SBR) at the genus level. Moreover, the reactor design seemed to influence the community dynamics, and reactor operation significantly influenced the overall community composition. The MBBR seems to be the reactor type of choice for mainstream wastewater treatment.

  1. State Standards and Evolution

    ERIC Educational Resources Information Center

    Moore, Randy

    2004-01-01

    Throughout the United States various individuals and groups have tried to subvert science education by removing or weakening the treatment of evolution in state science-education standards. Most states' science-education standards support the teaching of evolution, but many in the general public and some policymakers want science classrooms to…

  2. A Proposed Paradigm for Solar Cycle Dynamics Mediated via Turbulent Pumping of Magnetic Flux in Babcock-Leighton-type Solar Dynamos

    NASA Astrophysics Data System (ADS)

    Hazra, Soumitra; Nandy, Dibyendu

    2016-11-01

    At present, the Babcock-Leighton flux transport solar dynamo models appear to be the most promising models for explaining diverse observational aspects of the sunspot cycle. The success of these flux transport dynamo models is largely dependent upon a single-cell meridional circulation with a deep equatorward component at the base of the Sun’s convection zone. However, recent observations suggest that the meridional flow may in fact be very shallow (confined to the top 10% of the Sun) and more complex than previously thought. Taken together, these observations raise serious concerns on the validity of the flux transport paradigm. By accounting for the turbulent pumping of magnetic flux, as evidenced in magnetohydrodynamic simulations of solar convection, we demonstrate that flux transport dynamo models can generate solar-like magnetic cycles even if the meridional flow is shallow. Solar-like periodic reversals are recovered even when meridional circulation is altogether absent. However, in this case, the solar surface magnetic field dynamics does not extend all the way to the polar regions. Very importantly, our results demonstrate that the Parker-Yoshimura sign rule for dynamo wave propagation can be circumvented in Babcock-Leighton dynamo models by the latitudinal component of turbulent pumping, which can generate equatorward propagating sunspot belts in the absence of a deep, equatorward meridional flow. We also show that variations in turbulent pumping coefficients can modulate the solar cycle amplitude and periodicity. Our results suggest the viability of an alternate magnetic flux transport paradigm—mediated via turbulent pumping—for sustaining solar-stellar dynamo action.

  3. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    SciTech Connect

    Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  4. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    SciTech Connect

    Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG&G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  5. The search for sterile neutrinos at reactors and underground laboratories

    NASA Astrophysics Data System (ADS)

    Langford, Thomas

    2017-01-01

    From the initial discovery of neutrinos to the observation of neutrino oscillations, unexpected results have lead to deeper understanding of physics. However, as experiments and theoretical predictions have improved, new anomalies have surfaced that could point to beyond the Standard Model physics. Leading hypotheses invoke a new form of matter, sterile neutrinos, as a possible resolution of these outstanding questions. New experimental efforts are underway to probe short-baseline neutrino oscillations with reactors and radioactive sources. This talk will highlight developments in current and next generation experiments and present possible outcomes for the next few years.

  6. Plasma generators, reactor systems and related methods

    SciTech Connect

    Kong, Peter C.; Pink, Robert J.; Lee, James E.

    2007-06-19

    A plasma generator, reactor and associated systems and methods are provided in accordance with the present invention. A plasma reactor may include multiple sections or modules which are removably coupled together to form a chamber. Associated with each section is an electrode set including three electrodes with each electrode being coupled to a single phase of a three-phase alternating current (AC) power supply. The electrodes are disposed about a longitudinal centerline of the chamber and are arranged to provide and extended arc and generate an extended body of plasma. The electrodes are displaceable relative to the longitudinal centerline of the chamber. A control system may be utilized so as to automatically displace the electrodes and define an electrode gap responsive to measure voltage or current levels of the associated power supply.

  7. Space reactor assessment and validation study

    NASA Technical Reports Server (NTRS)

    Gedeon, Stephen; Morey, Dennis

    1987-01-01

    The present difficulties experienced by the United States in launching payloads into space has suggested a number of problems which are associated with the handling of hazardous materials in spacecraft. The question has arisen as to the safety of launching highly radioactive material such as plutonium-238, related to the possibility of its dispersion into the atmosphere during a launch vehicle explosion. An alternative is the use of a small nuclear reactor which is not started until it is in space and contains little or no radioactivity at launch. A first order assessment of six small reactor concepts with power levels up to 100 MWe was performed. Both the nuclear feasibility of these concepts to operate at their rated power levels between 7 and 10 years and the capability of these concepts to remain subcritical both before and during launch and also in the case of water immersion during a potential launch failure or abort were investigated.

  8. Computational mathematics and physics of fusion reactors

    PubMed Central

    Garabedian, Paul R.

    2003-01-01

    Theory has contributed significantly to recent advances in magnetic fusion research. New configurations have been found for a stellarator experiment by computational methods. Solutions of a free-boundary problem are applied to study the performance of the plasma and look for islands in the magnetic surfaces. Mathematical analysis and numerical calculations have been used to study equilibrium, stability, and transport of optimized fusion reactors. PMID:14614129

  9. Reactor building assembly and method of operation

    SciTech Connect

    Fennern, L.E.; Caraway, H.A.; Hsu, Li C.

    1993-06-01

    A reactor building assembly is described comprising: a reactor pressure vessel containing a reactor core for generating heat in the form of steam; a containment vessel enclosing said pressure vessel; a first enclosure surrounding said containment vessel and spaced laterally therefrom to define a first chamber there between, and having a top and a bottom; a second enclosure surrounding said first enclosure and spaced laterally therefrom to define a second chamber there between, and having a top and a bottom; a building inlet for receiving into said second chamber fresh air from outside said second enclosure; a building outlet for discharging stale air from said first chamber; a transfer duct disposed through said first enclosure selectively joining in flow communication said first and second chambers; said building inlet being disposed at said second enclosure top, said building outlet being disposed at said first enclosure top, and said transfer duct being disposed adjacent said first enclosure bottom for allowing said fresh air to flow downwardly by gravity through said second chamber and through said transfer duct into said first chamber for cooling said first chamber, said stale air flowing upwardly by natural buoyancy for discharger from said first chamber through said building outlet; an exhaust stack disposed above said building outlet and in flow communication therewith for channeling upwardly said stale air from said first chamber for discharge into the surrounding environs; and a passive first driving means for increasing flow of said stale air from said building outlet comprising: an isolation pool containing isolation water; an isolation condenser disposed in said isolation pool, and joined in flow communication with said reactor pressure vessel for receiving primary steam therefrom, said primary steam being cooled in said isolation condenser for heating said isolation water to generate secondary steam.

  10. Design and evaluation of experimental ceramic automobile thermal reactors

    NASA Technical Reports Server (NTRS)

    Stone, P. L.; Blankenship, C. P.

    1974-01-01

    The paper summarizes the results obtained in an exploratory evaluation of ceramics for automobile thermal reactors. Candidate ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance, lasting 1100 hours in engine dynamometer tests and for more than 38,600 kilimeters (24,000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

  11. Design and evaluation of experimental ceramic automobile thermal reactors

    NASA Technical Reports Server (NTRS)

    Stone, P. L.; Blankenship, C. P.

    1974-01-01

    The results obtained in an exploratory evaluation of ceramics for automobile thermal reactors are summarized. Candidate ceramic materials were evaluated in several reactor designs by using both engine-dynamometer and vehicle road tests. Silicon carbide contained in a corrugated-metal support structure exhibited the best performance, lasting 1100 hr in engine-dynamometer tests and more than 38,600 km (24000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as those containing silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

  12. Plant maintenance and advanced reactors, 2007

    SciTech Connect

    Agnihotri, Newal

    2007-09-15

    The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: A new day for energy in America; Committed to success more than ever, by Andy White, GE--Hitachi Nuclear Energy; Competitive technology for decades, by Steve Tritch, Westinghouse Electric Company; Pioneers of positive community relationship, by Exelon Nuclear; A robust design for 60-years, by Ray Ganthner, Areva; Aiming at no evacuation plants, by Kumiaki Moriya, Hitachi-GE Nuclear Energy, Ltd.; and, Desalination and hydrogen economy, by Dr. I. Khamis, International Atomic Energy Agency. Industry innovation articles in this issue are: Reactor vessel closure head project, by Jeff LeClair, Prairie Island Nuclear Generating Plant; and Submersible remote-operated vehicle, by Michael S. Rose, Entergy's Fitzpatrick Nuclear Station.

  13. Nuclear reactor insulation and preheat system

    DOEpatents

    Wampole, Nevin C.

    1978-01-01

    An insulation and preheat system for preselected components of a fluid cooled nuclear reactor. A gas tight barrier or compartment of thermal insulation surrounds the selected components and includes devices to heat the internal atmosphere of the compartment. An external surface of the compartment or enclosure is cooled, such as by a circulating fluid. The heating devices provide for preheating of the components, as well as maintenance of a temperature sufficient to ensure that the reactor coolant fluid will not solidify during shutdown. The external cooling limits the heat transferred to other plant structures, such as supporting concrete and steel. The barrier is spaced far enough from the surrounded components so as to allow access for remote or manual inspection, maintenance, and repair.

  14. NEUTRONIC REACTOR OPERATIONAL METHOD AND CORE SYSTEM

    DOEpatents

    Winters, C.E.; Graham, C.B.; Culver, J.S.; Wilson, R.H.

    1960-07-19

    Homogeneous neutronic reactor systems are described wherein an aqueous fuel solution is continuously circulated through a spherical core tank. The pumped fuel solution-is injected tangentially into the hollow spherical interior, thereby maintaining vigorous rotation of the solution within the tank in the form of a vortex; gaseous radiolytic decomposition products concentrate within the axial vortex cavity. The evolved gas is continuously discharged through a gas- outlet port registering with an extremity of the vortex cavity. and the solution stream is discharged through an annular liquid outlet port concentrically encircling the gas outlet by virtue of which the vortex and its cavity are maintained precisely axially aligned with the gas outlet. A primary heat exchanger extracts useful heat from the hot effluent fuel solution before its recirculation into the core tank. Hollow cylinders and other alternative core- tank configurations defining geometric volumes of revolution about a principal axis are also covered. AEC's Homogeneous Reactor Experiment No. 1 is a preferred embodiment.

  15. 3-flavor oscillations with current and future reactor experiments

    NASA Astrophysics Data System (ADS)

    Dwyer, Dan

    2017-01-01

    Nuclear reactors have been a crucial tool for our understanding of neutrinos. The disappearance of electron antineutrinos emitted by nuclear reactors has firmly established that neutrino flavor oscillates, and that neutrinos consequently have mass. The current generation of precision measurements rely on some of the world's most intense reactor facilities to demonstrate that the electron antineutrino mixes with the third antineutrino mass eigenstate (v3-). Accurate measurements of antineutrino energies robustly determine the tiny difference between the masses-squared of the v3- state and the two more closely-spaced v1- and v2- states. These results have given us a much clearer picture of neutrino mass and mixing, yet at the same time open major questions about how to account for these small but non-zero masses in or beyond the Standard Model. These observations have also opened the door for a new generation of experiments which aim to measure the ordering of neutrino masses and search for potential violation of CP symmetry by neutrinos. I will provide a brief overview of this exciting field. Work supported under DOE OHEP DE-AC02-05CH11231.

  16. NUCLEAR REACTOR AND THERMIONIC FUEL ELEMENT THEREFOR

    DOEpatents

    Rasor, N.S.; Hirsch, R.L.

    1963-12-01

    The patent relates to the direct conversion of fission heat to electricity by use of thermionic plasma diodes having fissionable material cathodes, said diodes arranged to form a critical mass in a nuclear reactor. The patent describes a fuel element comprising a plurality of diodes each having a fissionable material cathode, an anode around said cathode, and an ionizable gas therebetween. Provision is made for flowing the gas and current serially through the diodes. (AEC)

  17. LBB application in the US operating and advanced reactors

    SciTech Connect

    Wichman, K.; Tsao, J.; Mayfield, M.

    1997-04-01

    The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRC and the International Piping Integrity Research Group is also briefly summarized.

  18. Steam Generator of the International Reactor Innovative and Secure

    SciTech Connect

    Cinotti, L.; Bruzzone, M.; Meda, N.; Corsini, G.; Lombardi, C.V.; Ricotti, M.; Conway, L.E.

    2002-07-01

    IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the main reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long-life core and enhanced safety to address the requirements defined by the US DOE for Generation IV reactors. The design of the steam generators, which are internally contained within the reactor vessel, is a major design effort in the development of the integral IRIS concept. The ongoing design activity about the steam generator is the subject of this paper. (authors)

  19. Offsite dose calculation manual guidance: Standard radiological effluent controls for pressurized water reactors

    SciTech Connect

    Meinke, W.W.; Essig, T.H.

    1991-04-01

    This report contains guidance which may be voluntarily used by licensees who choose to implement the provision of Generic Letter 89-01, which allows Radiological Effect Technical Specifications (RETS) to be removed from the main body of the Technical Specifications and placed in the Offsite Dose Calculation Manual (ODCM). Guidance is provided for Standard Effluent Controls definitions, Controls for effluent monitoring instrumentation, Controls for effluent releases, Controls for radiological environmental monitoring, and the basis for Controls. Guidance on the formulation of RETS has been available in draft from (NUREG-0471 and -0473) for a number of years; the current effort simply recasts those RETS into Standard Radiological Effluent Controls for application to the ODCM. Also included for completeness are: (1) radiological environmental monitoring program guidance previously which had been available as a Branch Technical Position (Rev. 1, November 1979); (2) existing ODCM guidance; and (3) a reproduction of generic Letter 89-01.

  20. Animal Guts as Ideal Reactors: An Open-Ended Project for a Course in Kinetics and Reactor Design.

    ERIC Educational Resources Information Center

    Carlson, Eric D.; Gast, Alice P.

    1998-01-01

    Presents an open-ended project tailored for a senior kinetics and reactor design course in which basic reactor design equations are used to model the digestive systems of several animals. Describes the assignment as well as the results. (DDR)

  1. Fast Reactor Alternative Studies: Effects of Transuranic Groupings on Metal and Oxide Sodium Fast Reactor Designs

    SciTech Connect

    R. Ferrer; M. Asgari; S. Bays; B. Forget

    2007-09-01

    A 1000 MWth commercial-scale Sodium Fast Reactor (SFR) design with a conversion ratio (CR) of 0.50 was selected in this study to perform perturbations on the external feed coming from Light Water Reactor Spent Nuclear Fuel (LWR SNF) and separation groupings in the reprocessing scheme. A secondary SFR design with a higher conversion ratio (CR=0.75) was also analyzed as a possible alternative, although no perturbations were applied to this model.

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

    SciTech Connect

    B. Boer; A. M. Ougouag

    2010-09-01

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

  3. Gamma dose rate estimation and operation management suggestions for decommissioning the reactor pressure vessel of HTR-PM

    SciTech Connect

    Sheng Fang; Hong Li; Jianzhu Cao; Wenqian Li; Feng Xie; Jiejuan Tong

    2013-07-01

    China is now designing and constructing a high temperature gas cooled reactor-pebble bed module (HTR-PM). In order to investigate the future decommissioning approach and evaluate possible radiation dose, gamma dose rate near the reactor pressure vessel was calculated for different cooling durations using QAD-CGA program. The source term of this calculation was provided by KORIGEN program. Based on the calculated results, the spatial distribution and temporal changes of gamma dose rate near reactor pressure vessel was systematically analyzed. A suggestion on planning decommissioning operation of reactor pressure vessel of HTRPM was given based on calculated dose rate and the Chinese Standard GB18871-2002. (authors)

  4. 75 FR 70042 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-11-16

    ... All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order... above, shall be submitted to the NRC to the attention of the Director, Office of Nuclear Reactor... properly marked and handled in accordance with 10 CFR 73.21. The Director, Office of Nuclear...

  5. 75 FR 68009 - Office of New Reactors; Notice of Availability of the Final Staff Guidance Standard Review Plan...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-11-04

    ... Access and Management System (ADAMS) Accession No. ML102230082). The NRC staff issues revisions to SRP... Security--Combined License and Operating Reactors,'' ADAMS Accession No. ML100350158. There were no... the proposed notification as stated above. ADDRESSES: The NRC maintains ADAMS, which provides text...

  6. Standardization Today and Tomorrow

    DTIC Science & Technology

    1998-05-01

    Order for the Whole World) Eberstein, H.-H.: Sicherheitstechnisches Recht und Normen ; in: Peters. 0. H. u. A. Meyna: Handbuch der Si- cherheitstechnik...on the Protection of Labor for the Development and Design Sector) Landis, J. W.: Freie Welt - freier Handel - freie Normen ; in: DIN-Mitteilungen 57...Norm auf die Vielfalt effect of company standard on variety KonformitAt zu nationalen Normen conformity to national standards Konformit6t zu

  7. The phases of small networks of chemical reactors and neurons

    PubMed

    Schinor; Schneider

    2000-07-15

    We present an experimental study of the phase relationships observed in small reactor networks consisting of two and three continuous flow stirred tank reactors. In the three-reactor network one chemical oscillator is coupled to two other reactors in parallel in analogy to a small neural net. Each reactor contains an identical reaction mixture of the excitable Belousov-Zhabotinsky reaction which is characterized by its bifurcation diagram, where the electrical current is the bifurcation parameter. Coupling between the reactors is electrical via Pt-working electrodes and it can be either repulsive (inhibitory) or attractive (excitatory). An external electrical stimulus is applied to all three reactors in the form of an asymmetric electrical current pulse which sweeps across the bifurcation diagram. As a consequence, all three reactors oscillate with characteristic oscillation patterns or remain silent in analogy to the firing of neurons. The observed phase behavior depends on the type of coupling in a complex way. This situation is analogous to the in vivo measurements on single neurons (local neurons and projection neurons) performed by G. Laurent and co-workers on the olfactory system of the locust. We propose a simple neural network similar to the reactor network using the Hodgkin-Huxley model to simulate the action potentials of the coupled single neurons. Analogies between the reactor network and the neural network are discussed.

  8. BDDR, a new CEA technological and operating reactor database

    SciTech Connect

    Soldevilla, M.; Salmons, S.; Espinosa, B.

    2013-07-01

    The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a unique repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)

  9. The role of integral experiments and nuclear cross section evaluations in space nuclear reactor design

    NASA Astrophysics Data System (ADS)

    Moses, David L.; McKnight, Richard D.

    The importance of the nuclear and neutronic properties of candidate space reactor materials to the design process has been acknowledged as has been the use of benchmark reactor physics experiments to verify and qualify analytical tools used in design, safety, and performance evaluation. Since June 1966, the Cross Section Evaluation Working Group (CSEWG) has acted as an interagency forum for the assessment and evaluation of nuclear reaction data used in the nuclear design process. CSEWG data testing has involved the specification and calculation of benchmark experiments which are used widely for commercial reactor design and safety analysis. These benchmark experiments preceded the issuance of the industry standards for acceptance, but the benchmarks exceed the minimum acceptance criteria for such data. Thus, a starting place has been provided in assuring the accuracy and uncertainty of nuclear data important to space reactor applications.

  10. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    NASA Astrophysics Data System (ADS)

    Kulesza, Joel A.; Arzu Alpan, F.

    2016-02-01

    This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  11. Geogenic organic contaminants in the low-rank coal-bearing Carrizo-Wilcox aquifer of East Texas, USA

    USGS Publications Warehouse

    Chakraborty, Jayeeta; Varonka, Matthew S.; Orem, William H.; Finkelman, Robert B.; Manton, William

    2017-01-01

    The organic composition of groundwater along the Carrizo-Wilcox aquifer in East Texas (USA), sampled from rural wells in May and September 2015, was examined as part of a larger study of the potential health and environmental effects of organic compounds derived from low-rank coals. The quality of water from the low-rank coal-bearing Carrizo-Wilcox aquifer is a potential environmental concern and no detailed studies of the organic compounds in this aquifer have been published. Organic compounds identified in the water samples included: aliphatics and their fatty acid derivatives, phenols, biphenyls, N-, O-, and S-containing heterocyclic compounds, polycyclic aromatic hydrocarbons (PAHs), aromatic amines, and phthalates. Many of the identified organic compounds (aliphatics, phenols, heterocyclic compounds, PAHs) are geogenic and originated from groundwater leaching of young and unmetamorphosed low-rank coals. Estimated concentrations of individual compounds ranged from about 3.9 to 0.01 μg/L. In many rural areas in East Texas, coal strata provide aquifers for drinking water wells. Organic compounds observed in groundwater are likely to be present in drinking water supplied from wells that penetrate the coal. Some of the organic compounds identified in the water samples are potentially toxic to humans, but at the estimated levels in these samples, the compounds are unlikely to cause acute health problems. The human health effects of low-level chronic exposure to coal-derived organic compounds in drinking water in East Texas are currently unknown, and continuing studies will evaluate possible toxicity.

  12. Geogenic organic contaminants in the low-rank coal-bearing Carrizo-Wilcox aquifer of East Texas, USA

    NASA Astrophysics Data System (ADS)

    Chakraborty, Jayeeta; Varonka, Matthew; Orem, William; Finkelman, Robert B.; Manton, William

    2017-01-01

    The organic composition of groundwater along the Carrizo-Wilcox aquifer in East Texas (USA), sampled from rural wells in May and September 2015, was examined as part of a larger study of the potential health and environmental effects of organic compounds derived from low-rank coals. The quality of water from the low-rank coal-bearing Carrizo-Wilcox aquifer is a potential environmental concern and no detailed studies of the organic compounds in this aquifer have been published. Organic compounds identified in the water samples included: aliphatics and their fatty acid derivatives, phenols, biphenyls, N-, O-, and S-containing heterocyclic compounds, polycyclic aromatic hydrocarbons (PAHs), aromatic amines, and phthalates. Many of the identified organic compounds (aliphatics, phenols, heterocyclic compounds, PAHs) are geogenic and originated from groundwater leaching of young and unmetamorphosed low-rank coals. Estimated concentrations of individual compounds ranged from about 3.9 to 0.01 μg/L. In many rural areas in East Texas, coal strata provide aquifers for drinking water wells. Organic compounds observed in groundwater are likely to be present in drinking water supplied from wells that penetrate the coal. Some of the organic compounds identified in the water samples are potentially toxic to humans, but at the estimated levels in these samples, the compounds are unlikely to cause acute health problems. The human health effects of low-level chronic exposure to coal-derived organic compounds in drinking water in East Texas are currently unknown, and continuing studies will evaluate possible toxicity.

  13. Design and performance of subgrade biogeochemical reactors.

    PubMed

    Gamlin, Jeff; Downey, Doug; Shearer, Brad; Favara, Paul

    2017-02-18

    Subgrade biogeochemical reactors (SBGRs), also commonly referred to as in situ bioreactors, are a unique technology for treatment of contaminant source areas and groundwater plume hot spots. SBGRs have most commonly been configured for enhanced reductive dechlorination (ERD) applications for chlorinated solvent treatment. However, they have also been designed for other contaminant classes using alternative treatment media. The SBGR technology typically consists of removal of contaminated soil via excavation or large-diameter augers, and backfill of the soil void with gravel and treatment amendments tailored to the target contaminant(s). In most cases SBGRs include installation of infiltration piping and a low-flow pumping system (typically solar-powered) to recirculate contaminated groundwater through the SBGR for treatment. SBGRs have been constructed in multiple configurations, including designs capable of meeting limited access restrictions at heavily industrialized sites, and at sites with restrictions on surface disturbance due to sensitive species or habitat issues. Typical performance results for ERD applications include 85 to 90 percent total molar reduction of chlorinated volatile organic compounds (CVOCs) near the SBGR and rapid clean-up of adjacent dissolved contaminant source areas. Based on a review of the literature and CH2M's field-scale results from over a dozen SBGRs with a least one year of performance data, important site-specific design considerations include: 1) hydraulic residence time should be long enough for sufficient treatment but not too long to create depressed pH and stagnant conditions (e.g., typically between 10 and 60 days), 2) reactor material should balance appropriate organic mulch as optimal bacterial growth media along with other organic additives that provide bioavailable organic carbon, 3) a variety of native bacteria are important to the treatment process, and 4) biologically mediated generation of iron sulfides along with

  14. 10 CFR 170.21 - Schedule of fees for production and utilization facilities, review of standard referenced design...

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    .... Amendment, Renewal, Other Approvals Full cost. C. Test Facility/Research Reactor/Critical Facility..., approvals of facility standard reference designs, re-qualification and replacement examinations for reactor... of fees Fees 1,2 A. Nuclear Power Reactors Application for Construction Permit Full cost. Early...

  15. 10 CFR 170.21 - Schedule of fees for production and utilization facilities, review of standard referenced design...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    .... Amendment, Renewal, Other Approvals Full cost. C. Test Facility/Research Reactor/Critical Facility..., approvals of facility standard reference designs, re-qualification and replacement examinations for reactor... of fees Fees 1,2 A. Nuclear Power Reactors Application for Construction Permit Full cost. Early...

  16. Using thermal balance model to determine optimal reactor volume and insulation material needed in a laboratory-scale composting reactor.

    PubMed

    Wang, Yongjiang; Pang, Li; Liu, Xinyu; Wang, Yuansheng; Zhou, Kexun; Luo, Fei

    2016-04-01

    A comprehensive model of thermal balance and degradation kinetics was developed to determine the optimal reactor volume and insulation material. Biological heat production and five channels of heat loss were considered in the thermal balance model for a representative reactor. Degradation kinetics was developed to make the model applicable to different types of substrates. Simulation of the model showed that the internal energy accumulation of compost was the significant heat loss channel, following by heat loss through reactor wall, and latent heat of water evaporation. Lower proportion of heat loss occurred through the reactor wall when the reactor volume was larger. Insulating materials with low densities and low conductive coefficients were more desirable for building small reactor systems. Model developed could be used to determine the optimal reactor volume and insulation material needed before the fabrication of a lab-scale composting system.

  17. Dynamic reactor modeling with applications to SPR and ZEDNA

    SciTech Connect

    Suo-Anttila, Ahti Jorma

    2011-12-01

    A dynamic reactor model has been developed for pulse-type reactor applications. The model predicts reactor power, axial and radial fuel expansion, prompt and delayed neutron population, and prompt and delayed gamma population. All model predictions are made as a function of time. The model includes the reactivity effect of fuel expansion on a dynamic timescale as a feedback mechanism for reactor power. All inputs to the model are calculated from first principles, either directly by solving systems of equations, or indirectly from Monte Carlo N-Particle Transport Code (MCNP) derived results. The model does not include any empirical parameters that can be adjusted to match experimental data. Comparisons of model predictions to actual Sandia Pulse Reactor SPR-III pulses show very good agreement for a full range of pulse magnitudes. The model is also applied to Z-pinch externally driven neutron assembly (ZEDNA) type reactor designs to model both normal and off-normal ZEDNA operations.

  18. NEUTRONIC REACTOR COUNTER METHOD AND SYSTEM

    DOEpatents

    Graham, C.B.; Spiewak, I.

    1960-05-31

    An improved method is given for controlling the rate of fission in circulating-fuel neutronic reactors in which the fuel is a homogeneous liquid containing fissionable material and a neutron moderator. A change in the rate of flssion is effected by preferentially retaining apart from the circulating fuel a variable amount of either fissionable material or moderator, thereby varying the concentration of fissionable material in the fuel. In the case of an aqueous fuel solution a portion of the water may be continuously vaporized from the circulating solution and the amount of condensate, or condensate plus make-up water, returned to the solution is varied to control the fission rate.

  19. Metal fires and their implications for advanced reactors.

    SciTech Connect

    Nowlen, Steven Patrick; Figueroa, Victor G.; Olivier, Tara Jean; Hewson, John C.; Blanchat, Thomas K.

    2010-10-01

    This report details the primary results of the Laboratory Directed Research and Development project (LDRD 08-0857) Metal Fires and Their Implications for Advance Reactors. Advanced reactors may employ liquid metal coolants, typically sodium, because of their many desirable qualities. This project addressed some of the significant challenges associated with the use of liquid metal coolants, primary among these being the extremely rapid oxidation (combustion) that occurs at the high operating temperatures in reactors. The project has identified a number of areas for which gaps existed in knowledge pertinent to reactor safety analyses. Experimental and analysis capabilities were developed in these areas to varying degrees. In conjunction with team participation in a DOE gap analysis panel, focus was on the oxidation of spilled sodium on thermally massive surfaces. These are spills onto surfaces that substantially cool the sodium during the oxidation process, and they are relevant because standard risk mitigation procedures seek to move spill environments into this regime through rapid draining of spilled sodium. While the spilled sodium is not quenched, the burning mode is different in that there is a transition to a smoldering mode that has not been comprehensively described previously. Prior work has described spilled sodium as a pool fire, but there is a crucial, experimentally-observed transition to a smoldering mode of oxidation. A series of experimental measurements have comprehensively described the thermal evolution of this type of sodium fire for the first time. A new physics-based model has been developed that also predicts the thermal evolution of this type of sodium fire for the first time. The model introduces smoldering oxidation through porous oxide layers to go beyond traditional pool fire analyses that have been carried out previously in order to predict experimentally observed trends. Combined, these developments add significantly to the safety

  20. Modeling considerations for the analysis of LMFBR steam generator tube clamps

    SciTech Connect

    Lay, D.M.; Piper, R.M.

    1982-01-01

    In the design of the Babcock and Wilcox Helical Coil Liquid Metal Fast Breeder Reactor (LMFBR) Steam Generator, the tube bundle is connected to the feedwater and steam plenums via ''inlet/outlet tubes''. Of prime importance in the design of these tubes is the tube-to-tube and tube-to-shell clamps which are provided to prevent detrimental vibration. This paper presents a method of modeling the tube-to-tube clamps to accurately predict tube-to-clamp interaction in the finite element analysis. It also demonstrates the validity of specific modeling assumptions in determining stresses in the clamp assembly.

  1. Preparation and application of immobilized enzymatic reactors for consecutive digestion with two enzymes.

    PubMed

    Wang, Bingbing; Shangguan, Lulu; Wang, Shulei; Zhang, Lingyi; Zhang, Weibing; Liu, Fan

    2016-12-16

    The bottom up strategy has drawn much attention due to the high accuracy, reliability, and reproducibility in protein identification in which proteins are digested into peptides. However, conventional solution-based digestion and enzymatic reactor with one protease immobilized cannot satisfy high throughput proteolysis of complex samples. Application of consecutive hydrolysis by enzymatic reactor can be a new strategy for high throughput proteolysis of complex samples by adjusting immobilization amount of the enzymes, enzyme ratio, as well as hydrolysis order of two enzymes. In this work, we propose immobilized enzymatic reactor for consecutive digestion with two enzymes by combining two enzyme reactors with trypsin and chymotrypsin immobilized, respectively. Each reactor was prepared individually by immobilizing only one protease (trypsin or chymotrypsin) to hybrid monolith with SBA-15 particles embedded. Proteolysis conditions including hydrolysis order and trypsin to chymotrypsin ratio etc. were studied using standard proteins. Best digestion performance was obtained when the proteins were digested by trypsin first with trypsin to chymotrypsin ratio of 1:1. When applying them to digestion of rat liver proteins, total 1651 proteins and 11011 peptides were identified by combining four enzymolysis strategies with two enzymes including proteolytic digestion in two consecutive enzymatic reactors, synergy enzymolysis with two enzymes in one immobilized enzymatic reactor and consecutive hydrolysis with two enzymes in-solution digestion respectively, in which consecutive enzymolysis in enzymatic reactors gave the best results with 1091 proteins and 5071 peptides identified. The reactors showed good digestion capability for proteins with different hydrophobicity and molecular weights, and will play an important role in high efficient and high throughput proteomics research.

  2. Preservice and Professional Standards.

    ERIC Educational Resources Information Center

    Edelfelt, Roy

    This paper, prepared for the September 5, 1968, National Education Association (NEA) Staff Conference, presents the NEA position, program, and strategy with regard to preservice and inservice teacher education and professional standards. Introductory remarks include a list of seven priorities which form the "framework of context of the NEA…

  3. Standards and Criteria

    ERIC Educational Resources Information Center

    Glass, Gene V.

    2011-01-01

    This monograph has grown out of a series of discussions and a six-month period of reading and reflecting on the literature which were initiated by Fritz Mosher's suggestions to the National Assessment of Educational Progress (NAEP) to examine the "standards" question. Conversations with Mosher himself and the staff of NAEP have been most…

  4. Flow Simulation and Optimization of Plasma Reactors for Coal Gasification

    NASA Astrophysics Data System (ADS)

    Ji, Chunjun; Zhang, Yingzi; Ma, Tengcai

    2003-10-01

    This paper reports a 3-d numerical simulation system to analyze the complicated flow in plasma reactors for coal gasification, which involve complex chemical reaction, two-phase flow and plasma effect. On the basis of analytic results, the distribution of the density, temperature and components' concentration are obtained and a different plasma reactor configuration is proposed to optimize the flow parameters. The numerical simulation results show an improved conversion ratio of the coal gasification. Different kinds of chemical reaction models are used to simulate the complex flow inside the reactor. It can be concluded that the numerical simulation system can be very useful for the design and optimization of the plasma reactor.

  5. Recent reactor testing and experience with gamma thermometers

    SciTech Connect

    Waring, J.P.; Smith, R.D.

    1983-02-01

    Recent experience with gamma thermometers for light water reactors has primarily been in the Framatome reactors operated by Electricite de France. Other recent testing has taken place at Oak Ridge National Laboratory and the Otto Hahn ship reactor. Earlier experience with gamma thermometers was in heavy water reactors at Savannah River and Halden. This paper presents recent data from the light water reactor (LWR) programs. The principles of design and operation of the Radcal gamma thermometer were presented in ''Gamma Thermometer Developments for Light Water Reactors'', Leyse and Smith/sup 1/. Observations from LWRs confirm the earlier experience from heavy water reactors that the gamma thermometer units give signals which are proportional to the power of surrounding fuel rods and virtually independent of exposure, surrounding poison and other conditions which affect signals of neutron sensitive devices. After 200 sensor-years in EdF reactors, there has been no change in the sensitivity of the devices. Nonetheless, the Radcal units can be recalibrated in-reactor by the introduction of electrical heating via a heater cable imbedded in the device. Algorithms and signal processing software have been developed to interpret and display the gamma thermometer signals. The results of this processing are illustrated here.

  6. Sensitivity Analysis of Reprocessing Cooling Times on Light Water Reactor and Sodium Fast Reactor Fuel Cycles

    SciTech Connect

    R. M. Ferrer; S. Bays; M. Pope

    2008-04-01

    The purpose of this study is to quantify the effects of variations of the Light Water Reactor (LWR) Spent Nuclear Fuel (SNF) and fast reactor reprocessing cooling time on a Sodium Fast Reactor (SFR) assuming a single-tier fuel cycle scenario. The results from this study show the effects of different cooling times on the SFR’s transuranic (TRU) conversion ratio (CR) and transuranic fuel enrichment. Also, the decay heat, gamma heat and neutron emission of the SFR’s fresh fuel charge were evaluated. A 1000 MWth commercial-scale SFR design was selected as the baseline in this study. Both metal and oxide CR=0.50 SFR designs are investigated.

  7. Reactor coolant pump monitoring and diagnostic system

    SciTech Connect

    Singer, R.M.; Gross, K.C.; Walsh, M. ); Humenik, K.E. )

    1990-01-01

    In order to reliably and safely operate a nuclear power plant, it is necessary to continuously monitor the performance of numerous subsystems to confirm that the plant state is within its prescribed limits. An important function of a properly designed monitoring system is the detection of incipient faults in all subsystems (with the avoidance of false alarms) coupled with an information system that provides the operators with fault diagnosis, prognosis of fault progression and recommended (either automatic or prescriptive) corrective action. In this paper, such a system is described that has been applied to reactor coolant pumps. This system includes a sensitive pattern-recognition technique based upon the sequential probability ratio test (SPRT) that detects incipient faults from validated signals, an expert system embodying knowledge bases on pump and sensor performance, extensive hypertext files containing operating and emergency procedures as well as pump and sensor information and a graphical interface providing the operator with easily perceived information on the location and character of the fault as well as recommended corrective action. This system is in the prototype stage and is currently being validated utilizing data from a liquid-metal cooled fast reactor (EBR-II). 3 refs., 4 figs.

  8. 10 CFR 50.43 - Additional standards and provisions affecting class 103 licenses and certifications for...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... propose nuclear reactor designs which differ significantly from light-water reactor designs that were... 10 Energy 1 2013-01-01 2013-01-01 false Additional standards and provisions affecting class 103 licenses and certifications for commercial power. 50.43 Section 50.43 Energy NUCLEAR REGULATORY...

  9. Exhaust system with emissions storage device and plasma reactor

    DOEpatents

    Hoard, John W.

    1998-01-01

    An exhaust system for a combustion system, comprising a storage device for collecting NO.sub.x, hydrocarbon, or particulate emissions, or mixture of these emissions, and a plasma reactor for destroying the collected emissions is described. After the emission is collected in by the storage device for a period of time, the emission is then destroyed in a non-thermal plasma generated by the plasma reactor. With respect to the direction of flow of the exhaust stream, the storage device must be located before the terminus of the plasma reactor, and it may be located wholly before, overlap with, or be contained within the plasma reactor.

  10. TEST REACTOR AREA PLOT PLAN CA. 1968. MTR AND ETR ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    TEST REACTOR AREA PLOT PLAN CA. 1968. MTR AND ETR AREAS SOUTH OF PERCH AVENUE. "COLD" SERVICES NORTH OF PERCH. ADVANCED TEST REACTOR IN NEW SECTION WEST OF COLD SERVICES SECTION. NEW PERIMETER FENCE ENCLOSES BETA RAY SPECTROMETER, TRA-669, AN ATR SUPPORT FACILITY, AND ATR STACK. UTM LOCATORS HAVE BEEN DELETED. IDAHO NUCLEAR CORPORATION, FROM A BLAW-KNOX DRAWING, 3/1968. INL INDEX NO. 530-0100-00-400-011646, REV. 0. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  11. Slurry reactor design studies

    SciTech Connect

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. ); Akgerman, A. ); Smith, J.M. )

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  12. R- AND P- REACTOR VESSEL IN-SITU DECOMISSIONING VISUALIZATION

    SciTech Connect

    Vrettos, N.; Bobbitt, J.; Howard, M.

    2010-06-07

    The R- & P- Reactor facilities were constructed in the early 1950's in response to Cold War efforts. The mission of the facilities was to produce materials for use in the nation's nuclear weapons stockpile. R-Reactor was removed from service in 1964 when President Johnson announced a slowdown of he nuclear arms race. PReactor continued operation until 1988 until the facility was taken off-line to modernize the facility with new safeguards. Efforts to restart the reactor ended in 1990 at the end of the Cold War. Both facilities have sat idle since their closure and have been identified as the first two reactors for closure at SRS.

  13. RSMASS-D models: An improved method for estimating reactor and shield mass for space reactor applications

    SciTech Connect

    Marshall, A.C.

    1997-10-01

    Three relatively simple mathematical models have been developed to estimate minimum reactor and radiation shield masses for liquid-metal-cooled reactors (LMRs), in-core thermionic fuel element (TFE) reactors, and out-of-core thermionic reactors (OTRs). The approach was based on much of the methodology developed for the Reactor/Shield Mass (RSMASS) model. Like the original RSMASS models, the new RSMASS-derivative (RSMASS-D) models use a combination of simple equations derived from reactor physics and other fundamental considerations, along with tabulations of data from more detailed neutron and gamma transport theory computations. All three models vary basic design parameters within a range specified by the user to achieve a parameter choice that yields a minimum mass for the power level and operational time of interest. The impact of critical mass, fuel damage, and thermal limitations are accounted for to determine the required fuel mass. The effect of thermionic limitations are also taken into account for the thermionic reactor models. All major reactor component masses are estimated, as well as instrumentation and control (I&C), boom, and safety system masses. A new shield model was developed and incorporated into all three reactor concept models. The new shield model is more accurate and simpler to use than the approach used in the original RSMASS model. The estimated reactor and shield masses agree with the mass predictions from separate detailed calculations within 15 percent for all three models.

  14. BOILING REACTORS

    DOEpatents

    Untermyer, S.

    1962-04-10

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

  15. Evaluation of rotating-cylinder and piston-cylinder reactors for ground-based emulsion polymerization

    NASA Technical Reports Server (NTRS)

    Vanderhoff, J. W.; El-Aasser, M. S.

    1987-01-01

    The objectives of this program are to apply ground-based emulsion polymerization reactor technology to improve the production of: monodisperse latex particles for calibration standards, chromatographic separation column packing, and medical research; and commercial latexes such as those used for coatings, foams, and adhesives.

  16. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1961-09-01

    A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

  17. Ultrasonic level and temperature sensor for power reactor applications

    SciTech Connect

    Dress, W.B.: Miller, G.N.

    1983-01-01

    An ultrasonic waveguide employing torsional and extensional acoustic waves has been developed for use as a level and temperature sensor in pressurized and boiling water nuclear power reactors. Features of the device include continuous measurement of level, density, and temperature producing a real-time profile of these parameters along a chosen path through the reactor vessel.

  18. Safety and licensing for small and medium power reactors

    SciTech Connect

    Trauger, D.B.

    1987-01-01

    Proposed new concepts for small and medium power reactors differ substantially from traditional Light Water Reactors (LWRs). Although designers have a large base of experience in safety and licensing, much of it is not relevant to new concepts. It can be a disadvantage if regulators apply LWR rules directly. A fresh start is appropriate. The extensive interactions between industry, regulators, and the public complicates but may enhance safety. It is basic to recognize the features that distinguish nuclear energy safety from that for other industries. These features include: nuclear reactivity, fission product radiation, and radioactive decay heat. Small and medium power reactors offer potential advantages over LWRs, particularly for reactivity and decay heat.

  19. Sodium fast reactor safety and licensing research plan. Volume II.

    SciTech Connect

    Ludewig, H.; Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.; Lambert, J.; Hayes, S.; Sackett, J.; Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  20. Cars applications in chemical reactors, combustion and heat transfer

    NASA Astrophysics Data System (ADS)

    Greenhalgh, D. A.; Porter, F. M.

    1986-08-01

    This paper illustrates the use of the CARS technique in the fields of Chemical Reactor engineering, combustion and Heat Transfer. Examples of recent results from a catalytic chemical reactor, an operating production petrol engine and an oil spray furnace are given. The experimentally determined accuracy of CARS nitrogen thermometry for both mean and single pulse measurements is presented.

  1. Oak Ridge Research reactor shutdown maintenance and surveillance

    SciTech Connect

    Coleman, G.H.; Laughlin, D.L.

    1991-05-01

    The Department of Energy ordered the Oak Ridge Research Reactor to be placed in permanent shutdown on July 14, 1987. The paper outlines routine maintenance activities and surveillance tests performed April through September, 1990, on the reactor instrumentation and controls, process system, and the gaseous waste filter system. Preparations are being made to transfer the facility to the Remedial Action Program. 6 tabs. (MHB)

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

    SciTech Connect

    Petrov, B. Y.; Kuijper, J. C.; Oppe, J.; De Haas, J. B. M.

    2012-07-01

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

  3. Current and future trends for biofilm reactors for fermentation processes.

    PubMed

    Ercan, Duygu; Demirci, Ali

    2015-03-01

    Biofilms in the environment can both cause detrimental and beneficial effects. However, their use in bioreactors provides many advantages including lesser tendencies to develop membrane fouling and lower required capital costs, their higher biomass density and operation stability, contribution to resistance of microorganisms, etc. Biofilm formation occurs naturally by the attachment of microbial cells to the support without use of any chemicals agent in biofilm reactors. Biofilm reactors have been studied and commercially used for waste water treatment and bench and pilot-scale production of value-added products in the past decades. It is important to understand the fundamentals of biofilm formation, physical and chemical properties of a biofilm matrix to run the biofilm reactor at optimum conditions. This review includes the principles of biofilm formation; properties of a biofilm matrix and their roles in the biofilm formation; factors that improve the biofilm formation, such as support materials; advantages and disadvantages of biofilm reactors; and industrial applications of biofilm reactors.

  4. On use of ZPR research reactors and associated instrumentation and measurement methods for reactor physics studies

    SciTech Connect

    Chauvin, J.P.; Blaise, P.; Lyoussi, A.

    2015-07-01

    The French atomic and alternative energies -CEA- is strongly involved in research and development programs concerning the use of nuclear energy as a clean and reliable source of energy and consequently is working on the present and future generation of reactors on various topics such as ageing plant management, optimization of the plutonium stockpile, waste management and innovative systems exploration. Core physics studies are an essential part of this comprehensive R and D effort. In particular, the Zero Power Reactor (ZPR) of CEA: EOLE, MINERVE and MASURCA play an important role in the validation of neutron (as well photon) physics calculation tools (codes and nuclear data). The experimental programs defined in the CEA's ZPR facilities aim at improving the calculation routes by reducing the uncertainties of the experimental databases. They also provide accurate data on innovative systems in terms of new materials (moderating and decoupling materials) and new concepts (ADS, ABWR, new MTR (e.g. JHR), GENIV) involving new fuels, absorbers and coolant materials. Conducting such interesting experimental R and D programs is based on determining and measuring main parameters of phenomena of interest to qualify calculation tools and nuclear data 'libraries'. Determining these parameters relies on the use of numerous and different experimental techniques using specific and appropriate instrumentation and detection tools. Main ZPR experimental programs at CEA, their objectives and challenges will be presented and discussed. Future development and perspectives regarding ZPR reactors and associated programs will be also presented. (authors)

  5. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    SciTech Connect

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  6. Frequency Standards and Metrology

    NASA Astrophysics Data System (ADS)

    Maleki, Lute

    2009-04-01

    Preface / Lute Maleki -- Symposium history / Jacques Vanier -- Symposium photos -- pt. I. Fundamental physics. Variation of fundamental constants from the big bang to atomic clocks: theory and observations (Invited) / V. V. Flambaum and J. C. Berengut. Alpha-dot or not: comparison of two single atom optical clocks (Invited) / T. Rosenband ... [et al.]. Variation of the fine-structure constant and laser cooling of atomic dysprosium (Invited) / N. A. Leefer ... [et al.]. Measurement of short range forces using cold atoms (Invited) / F. Pereira Dos Santos ... [et al.]. Atom interferometry experiments in fundamental physics (Invited) / S. W. Chiow ... [et al.]. Space science applications of frequency standards and metrology (Invited) / M. Tinto -- pt. II. Frequency & metrology. Quantum metrology with lattice-confined ultracold Sr atoms (Invited) / A. D. Ludlow ... [et al.]. LNE-SYRTE clock ensemble: new [symbol]Rb hyperfine frequency measurement - spectroscopy of [symbol]Hg optical clock transition (Invited) / M. Petersen ... [et al.]. Precise measurements of S-wave scattering phase shifts with a juggling atomic clock (Invited) / S. Gensemer ... [et al.]. Absolute frequency measurement of the [symbol] clock transition (Invited) / M. Chwalla ... [et al.]. The semiclassical stochastic-field/atom interaction problem (Invited) / J. Camparo. Phase and frequency noise metrology (Invited) / E. Rubiola ... [et al.]. Optical spectroscopy of atomic hydrogen for an improved determination of the Rydberg constant / J. L. Flowers ... [et al.] -- pt. III. Clock applications in space. Recent progress on the ACES mission (Invited) / L. Cacciapuoti and C. Salomon. The SAGAS mission (Invited) / P. Wolf. Small mercury microwave ion clock for navigation and radioScience (Invited) / J. D. Prestage ... [et al.]. Astro-comb: revolutionizing precision spectroscopy in astrophysics (Invited) / C. E. Kramer ... [et al.]. High frequency very long baseline interferometry: frequency standards and

  7. The development and application of an improved reactor analysis model for fast reactors

    NASA Astrophysics Data System (ADS)

    Hou, Jia

    Accuracy in neutron cross sections calculation and consistency in reactor physics are fundamental requirements in advanced nuclear reactor design and analysis. The work presented in this dissertation focuses on the development and advanced application of a reactor analysis model with updated cross section libraries that is suitable for online cross section generation for fast reactors. Research has been performed in two areas of interest in reactor physics. The first target of the research is to develop effcient modeling capacity of the 1- D lattice code MICROX-2 for its neutron spectrum calculation based on Collision Probability Method (CPM). Expanded master cross section libraries have been generated based on updated nuclear data and optimized fine-group energy structure to accommodate both thermal and fast reactor spectra as well as to comply with the need for advanced fuel cycle analysis. After verifying the new libraries, the solution methods have been reviewed and updated, including the update of interpolation scheme for resonance self-shielding factors and improvement of spatial self-shielding models for various fuel assembly geometries. The assessment of the updated lattice calculation models has shown that the prediction accuracy of lattice properties represented by the eigenvalue and reaction rate ratios is improved, especially for fast neutron spectrum lattices of which the importance of neutrons in the unresolved energy range is high. The second target of the research is to improve the accuracy of few-group nuclear cross section generation for the reactor core calculation. A 2-D pin-by-pin lattice model has been developed based on embedded CPM within the framework of the Nodal Expansion Method (NEM), which is capable of modeling the heterogeneity of the fuel assembly. Then, an online cross section generation methodology along with discontinuity factors has been developed based on Iterative Diffusion- Diffusion Methodology (IDDM), which can minimize the

  8. Licensed reactor nuclear safety criteria applicable to DOE reactors

    SciTech Connect

    Not Available

    1991-04-01

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

  9. DNA-Based Enzyme Reactors and Systems

    PubMed Central

    Linko, Veikko; Nummelin, Sami; Aarnos, Laura; Tapio, Kosti; Toppari, J. Jussi; Kostiainen, Mauri A.

    2016-01-01

    During recent years, the possibility to create custom biocompatible nanoshapes using DNA as a building material has rapidly emerged. Further, these rationally designed DNA structures could be exploited in positioning pivotal molecules, such as enzymes, with nanometer-level precision. This feature could be used in the fabrication of artificial biochemical machinery that is able to mimic the complex reactions found in living cells. Currently, DNA-enzyme hybrids can be used to control (multi-enzyme) cascade reactions and to regulate the enzyme functions and the reaction pathways. Moreover, sophisticated DNA structures can be utilized in encapsulating active enzymes and delivering the molecular cargo into cells. In this review, we focus on the latest enzyme systems based on novel DNA nanostructures: enzyme reactors, regulatory devices and carriers that can find uses in various biotechnological and nanomedical applications. PMID:28335267

  10. Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

    SciTech Connect

    Cooke, Conrad; Spann, Holger

    2013-07-01

    Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle Plates to

  11. Research reactors

    SciTech Connect

    Tonneson, L.C.; Fox, G.J.

    1996-04-01

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world`s research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted.

  12. 77 FR 61446 - Proposed Revision Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-09

    ... COMMISSION Proposed Revision Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors..., ``Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors.'' DATES: Submit comments by... accident assessments for new reactors submitted to support design certification (DC) and combined...

  13. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    SciTech Connect

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  14. Reactor and method for hydrocracking carbonaceous material

    DOEpatents

    Duncan, Dennis A.; Beeson, Justin L.; Oberle, R. Donald; Dirksen, Henry A.

    1980-01-01

    Solid, carbonaceous material is cracked in the presence of hydrogen or other reducing gas to provide aliphatic and aromatic hydrocarbons of lower molecular weight for gaseous and liquid fuels. The carbonaceous material, such as coal, is entrained as finely divided particles in a flow of reducing gas and preheated to near the decomposition temperature of the high molecular weight polymers. Within the reactor, small quantities of oxygen containing gas are injected at a plurality of discrete points to burn corresponding amounts of the hydrogen or other fuel and elevate the mixture to high temperatures sufficient to decompose the high molecular weight, carbonaceous solids. Turbulent mixing at each injection point rapidly quenches the material to a more moderate bulk temperature. Additional quenching after the final injection point can be performed by direct contact with quench gas or oil. The reactions are carried out in the presence of a hydrogen-containing reducing gas at moderate to high pressure which stabilizes the products.

  15. Water and Regolith Shielding for Surface Reactor Missions

    SciTech Connect

    Poston, David I.; Sadasivan, Pratap; Dixon, David D.; Ade, Brian J.; Leichliter, Katrina J.

    2006-01-20

    This paper investigates potential shielding options for surface power fission reactors. The majority of work is focused on a lunar shield that uses a combination of water in stainless-steel cans and lunar regolith. The major advantage of a water-based shield is that development, testing, and deployment should be relatively inexpensive. This shielding approach is used for three surface reactor concepts: (1) a moderated spectrum, NaK cooled, Hastalloy/UZrH reactor, (2) a fast-spectrum, NaK-cooled, SS/UO2 reactor, and (3) a fast-spectrum, K-heat-pipe-cooled, SS/UO2 reactor. For this study, each of these reactors is coupled to a 25-kWt Stirling power system, designed for 5 year life. The shields are designed to limit the dose both to the Stirling alternators and potential astronauts on the surface. The general configuration used is to bury the reactor, but several other options exist as well. Dose calculations are presented as a function of distance from reactor, depth of buried hole, water boron concentration (if any), and regolith repacked density.

  16. Deposition reactors for solar grade silicon: A comparative thermal analysis of a Siemens reactor and a fluidized bed reactor

    NASA Astrophysics Data System (ADS)

    Ramos, A.; Filtvedt, W. O.; Lindholm, D.; Ramachandran, P. A.; Rodríguez, A.; del Cañizo, C.

    2015-12-01

    Polysilicon production costs contribute approximately to 25-33% of the overall cost of the solar panels and a similar fraction of the total energy invested in their fabrication. Understanding the energy losses and the behaviour of process temperature is an essential requirement as one moves forward to design and build large scale polysilicon manufacturing plants. In this paper we present thermal models for two processes for poly production, viz., the Siemens process using trichlorosilane (TCS) as precursor and the fluid bed process using silane (monosilane, MS). We validate the models with some experimental measurements on prototype laboratory reactors relating the temperature profiles to product quality. A model sensitivity analysis is also performed, and the effects of some key parameters such as reactor wall emissivity and gas distributor temperature, on temperature distribution and product quality are examined. The information presented in this paper is useful for further understanding of the strengths and weaknesses of both deposition technologies, and will help in optimal temperature profiling of these systems aiming at lowering production costs without compromising the solar cell quality.

  17. NEUTRONIC REACTOR AND FUEL ELEMENT THEREFOR

    DOEpatents

    Szilard, L.; Young, G.J.

    1958-03-01

    This patent relates to a reactor design of the type which employs solid fuel elements disposed in channels within the moderator through which channels and around the fuel elements is conveyed a coolant fiuid. The coolant channels are comprised of aluminum tubes extending through a solid moderator such as graphite and the fuel elements are comprised of an elongated solid body of natural uranium jacketed in an aluminum jacket with the ends thereof closed by aluminum caps of substantially greater thickness than the jacket was and in good thermal contact with the fuel material to facilitate the conduction of heat from the central portion of said ends to the coolant surrounding the fuel element to prevent overheating of said central portion.

  18. Large-scale spherical fixed bed reactors: Modeling and optimization

    SciTech Connect

    Hartig, F.; Keil, F.J. )

    1993-03-01

    Iterative dynamic programming (IDP) according to Luus was used for the optimization of the methanol production in a cascade of spherical reactors. The system of three spherical reactors was compared to an externally cooled tubular reactor and a quench reactor. The reactors were modeled by the pseudohomogeneous and heterogeneous approach. The effectiveness factors of the heterogeneous model were calculated by the dusty gas model. The IDP method was compared with sequential quadratic programming (SQP) and the Box complex method. The optimized distributions of catalyst volume with the pseudohomogeneous and heterogeneous model lead to different results. The IDP method finds the global optimum with high probability. A combination of IDP and SQP provides a reliable optimization procedure that needs minimum computing time.

  19. TRA603. PRECAST CONCRETE PANELS FOR SIDING. STANDARD SIZES AND DETAILS. ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    TRA-603. PRECAST CONCRETE PANELS FOR SIDING. STANDARD SIZES AND DETAILS. NOTE CONSTRUCTION WITH WIRE MESH AND CELLULAR GLASS INSULATION. BLAW-KNOX BKC-3150-803-13, 8/1950. INL INDEX NO. 53-0603-62-098-100572, REV. 7. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  20. CONVECTION REACTOR

    DOEpatents

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

    1960-03-22

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

  1. The detection of global convection on the sun by an analysis of line shift data of the John M. Wilcox Solar Observatory at Stanford University

    NASA Technical Reports Server (NTRS)

    Yoshimura, Hirokazu

    1987-01-01

    Signatures of the existence of the global convection in the sun were found in the absorption line shift data of the John M. Wilcox Solar Observatory at Stanford University. The signatures are characterized by persistent periodic time variations in the east-west component of the velocity fields defined by fitting a slope to the line shift data in a certain longitude window at a specified latitude and longitude by a least square method. The variations indicate that the amplitude of the velocity fields is about 100 m/s. It is suggested that several modes of global convection are coexisting in the solar convection zone.

  2. PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.

    SciTech Connect

    Cheng, L.; Diamond, D.; Xu, J.; Carew, J.; Rorer, D.

    2004-03-31

    Detailed reactor physics and safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron and photon transport calculations were performed with the MCNP code to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim safety arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim safety arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model that includes the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the

  3. Austenitic alloy and reactor components made thereof

    DOEpatents

    Bates, John F.; Brager, Howard R.; Korenko, Michael K.

    1986-01-01

    An austenitic stainless steel alloy is disclosed, having excellent fast neutron irradiation swelling resistance and good post irradiation ductility, making it especially useful for liquid metal fast breeder reactor applications. The alloy contains: about 0.04 to 0.09 wt. % carbon; about 1.5 to 2.5 wt. % manganese; about 0.5 to 1.6 wt. % silicon; about 0.030 to 0.08 wt. % phosphorus; about 13.3 to 16.5 wt. % chromium; about 13.7 to 16.0 wt. % nickel; about 1.0 to 3.0 wt. % molybdenum; and about 0.10 to 0.35 wt. % titanium.

  4. 23 CFR 625.4 - Standards, policies, and standard specifications.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 23 Highways 1 2011-04-01 2011-04-01 false Standards, policies, and standard specifications. 625.4 Section 625.4 Highways FEDERAL HIGHWAY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION ENGINEERING AND TRAFFIC OPERATIONS DESIGN STANDARDS FOR HIGHWAYS § 625.4 Standards, policies, and standard...

  5. Reactor design and integration into a nuclear electric spacecraft

    NASA Technical Reports Server (NTRS)

    Phillips, W. M.; Koenig, D. R.

    1978-01-01

    One of the well-defined applications for nuclear power in space is nuclear electric propulsion (NEP). Mission studies have identified the optimum power level (400 kWe). A single Shuttle launch requirement and science-package integration have added additional constraints to the design. A reactor design which will meet these constraints has been studied. The reactor employs 90 fuel elements, each heat pipe cooled. Reactor control is obtained with BeO/B4C drums in a BeO reflector. The balance of the spacecraft is shielded from the reactor with LiH. Power conditioning and reactor control drum drives are located behind the LiH with the power conditioning. Launch safety, mechanical design and integration with the power conversion subsystem are discussed.

  6. Advanced Plasma Pyrolysis Assembly (PPA) Reactor and Process Development

    NASA Technical Reports Server (NTRS)

    Wheeler, Richard R., Jr.; Hadley, Neal M.; Dahl, Roger W.; Abney, Morgan B.; Greenwood, Zachary; Miller, Lee; Medlen, Amber

    2012-01-01

    Design and development of a second generation Plasma Pyrolysis Assembly (PPA) reactor is currently underway as part of NASA's Atmosphere Revitalization Resource Recovery effort. By recovering up to 75% of the hydrogen currently lost as methane in the Sabatier reactor effluent, the PPA helps to minimize life support resupply costs for extended duration missions. To date, second generation PPA development has demonstrated significant technology advancements over the first generation device by doubling the methane processing rate while, at the same time, more than halving the required power. One development area of particular interest to NASA system engineers is fouling of the PPA reactor with carbonaceous products. As a mitigation plan, NASA MSFC has explored the feasibility of using an oxidative plasma based upon metabolic CO2 to regenerate the reactor window and gas inlet ports. The results and implications of this testing are addressed along with the advanced PPA reactor development.

  7. Licensed reactor nuclear safety criteria applicable to DOE reactors

    SciTech Connect

    Not Available

    1993-11-01

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

  8. Nitrification of industrial and domestic saline wastewaters in moving bed biofilm reactor and sequencing batch reactor.

    PubMed

    Bassin, João P; Dezotti, Marcia; Sant'anna, Geraldo L

    2011-01-15

    Nitrification of saline wastewaters was investigated in bench-scale moving-bed biofilm reactors (MBBR). Wastewater from a chemical industry and domestic sewage, both treated by the activated sludge process, were fed to moving-bed reactors. The industrial wastewater contained 8000 mg Cl(-)/L and the salinity of the treated sewage was gradually increased until that level. Residual substances present in the treated industrial wastewater had a strong inhibitory effect on the nitrification process. Assays to determine inhibitory effects were performed with the industrial wastewater, which was submitted to ozonation and carbon adsorption pretreatments. The latter treatment was effective for dissolved organic carbon (DOC) removal and improved nitrification efficiency. Nitrification percentage of the treated domestic sewage was higher than 90% for all tested chloride concentrations up to 8000 mg/L. Results obtained in a sequencing batch reactor (SBR) were consistent with those attained in the MBBR systems, allowing tertiary nitrification and providing adequate conditions for adaptation of nitrifying microorganisms even under stressing and inhibitory conditions.

  9. 78 FR 28896 - Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-16

    ... COMMISSION Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components... Combinations for Metal Primary Reactor Containment System Components,'' in which there are no substantive... loading combinations for metal primary reactor containment system components. ADDRESSES: Please refer...

  10. 76 FR 39922 - Office of New Reactors; Proposed Revision 4 to Standard Review Plan Section 8.1 on Electric Power...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-07-07

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Office of New Reactors; Proposed Revision 4 to Standard Review Plan Section 8.1 on Electric Power... 8.1 on ``Electric Power-- Introduction.'' The notice period for this notice closes on June 30,...

  11. 75 FR 36126 - Office of New Reactors; Proposed Revision to Standard Review Plan Section 13.6.3, Revision 1 on...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-06-24

    ... information in their comments that they do not want publicly disclosed. Federal Rulemaking Web site: Go to... COMMISSION Office of New Reactors; Proposed Revision to Standard Review Plan Section 13.6.3, Revision 1 on Physical Security--Early Site Permit AGENCY: Nuclear Regulatory Commission (NRC). ACTION: Solicitation...

  12. Globalization of ASME Nuclear Codes and Standards

    SciTech Connect

    Swayne, Rick; Erler, Bryan A.

    2006-07-01

    With the globalization of the nuclear industry, it is clear that the reactor suppliers are based in many countries around the world (such as United States, France, Japan, Canada, South Korea, South Africa) and they will be marketing their reactors to many countries around the world (such as US, China, South Korea, France, Canada, Finland, Taiwan). They will also be fabricating their components in many different countries around the world. With this situation, it is clear that the requirements of ASME Nuclear Codes and Standards need to be adjusted to accommodate the regulations, fabricating processes, and technology of various countries around the world. It is also very important for the American Society of Mechanical Engineers (ASME) to be able to assure that products meeting the applicable ASME Code requirements will provide the same level of safety and quality assurance as those products currently fabricated under the ASME accreditation process. To do this, many countries are in the process of establishing or changing their regulations, and it is important for ASME to interface with the appropriate organizations in those countries, in order to ensure there is effective use of ASME Codes and standards around the world. (authors)

  13. Emergency Management Standards and Schools

    ERIC Educational Resources Information Center

    National Clearinghouse for Educational Facilities, 2009

    2009-01-01

    This publication discusses emergency management standards for school use and lists standards recommended by FEMA's National Incident Management System (NIMS). Schools are encouraged to review these standards carefully and to adopt, where applicable, those that meet their needs. The lists of standards, resources, and references contained herein…

  14. NEW SOLID FUELS FROM COAL AND BIOMASS WASTE

    SciTech Connect

    Hamid Farzan

    2001-09-24

    Under DOE sponsorship, McDermott Technology, Inc. (MTI), Babcock and Wilcox Company (B and W), and Minergy Corporation developed and evaluated a sludge derived fuel (SDF) made from sewage sludge. Our approach is to dry and agglomerate the sludge, combine it with a fluxing agent, if necessary, and co-fire the resulting fuel with coal in a cyclone boiler to recover the energy and to vitrify mineral matter into a non-leachable product. This product can then be used in the construction industry. A literature search showed that there is significant variability of the sludge fuel properties from a given wastewater plant (seasonal and/or day-to-day changes) or from different wastewater plants. A large sewage sludge sample (30 tons) from a municipal wastewater treatment facility was collected, dried, pelletized and successfully co-fired with coal in a cyclone-equipped pilot. Several sludge particle size distributions were tested. Finer sludge particle size distributions, similar to the standard B and W size distribution for sub-bituminous coal, showed the best combustion and slagging performance. Up to 74.6% and 78.9% sludge was successfully co-fired with pulverized coal and with natural gas, respectively. An economic evaluation on a 25-MW power plant showed the viability of co-firing the optimum SDF in a power generation application. The return on equity was 22 to 31%, adequate to attract investors and allow a full-scale project to proceed. Additional market research and engineering will be required to verify the economic assumptions. Areas to focus on are: plant detail design and detail capital cost estimates, market research into possible project locations, sludge availability at the proposed project locations, market research into electric energy sales and renewable energy sales opportunities at the proposed project location. As a result of this program, wastes that are currently not being used and considered an environmental problem will be processed into a renewable

  15. Pressurized reactor system and a method of operating the same

    DOEpatents

    Isaksson, J.M.

    1996-06-18

    A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Super-atmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gasification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor. 2 figs.

  16. Pressurized reactor system and a method of operating the same

    DOEpatents

    Isaksson, Juhani M.

    1996-01-01

    A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Superatmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gassification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor.

  17. Paper and board mill effluent treatment with the combined biological-coagulation-filtration pilot scale reactor.

    PubMed

    Afzal, Muhammad; Shabir, Ghulam; Hussain, Irshad; Khalid, Zafar M

    2008-10-01

    Pilot scale reactor based on combined biological-coagulation-filtration treatments was designed and evaluated for the treatment of effluent from a paper and board mill. Biological treatment by fed batch reactor (FBR) followed by coagulation and sand filtration (SF) resulted in a total COD and BOD reduction of 93% and 96.5%, respectively. A significant reduction in both COD (90%) and BOD (92%) was also observed by sequencing batch reactor (SBR) process followed by coagulation and filtration. Untreated effluent was found to be toxic, whereas the treated effluents by either of the above two processes were found to be non-toxic when exposed to the fish for 72h. The resultant effluent from FBR-coagulation-sand filtration system meets National Environmental Quality Standards (NEQS) of Pakistan and can be discharged into the environment without any risks.

  18. The temperature effect and safety of the TOPAZ reactor

    NASA Astrophysics Data System (ADS)

    Artiukhov, G. Ia.; Zelentsov, S. N.; Ionkin, V. I.; Kudriavtsev, V. P.; Makarenkov, Iu. D.; Marin, S. N.; Pupko, V. Ia.; Raskach, F. P.

    The TOPAZ reactor-converter is a thermionic space nuclear-power system of less than 50 kW capacity; its reactor employs a high-temperature moderator containing hydrogen and highly enriched fuel. The designers of TOPAZ gave special attention to the determination of temperature effects (TEs) and temperature reactivity coefficient (TRC). A critical assembly that was heated isothermally by an electric furnace was used to conduct measurements of TEs, TRC, energy-release fields, control and safety system performance, and reactor spectrum and dynamic characteristics.

  19. Tritium Formation and Mitigation in High Temperature Reactors

    SciTech Connect

    Piyush Sabharwall; Carl Stoots

    2012-08-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. In order to prevent the tritium contamination of proposed reactor buildings and surrounding sites, this paper examines the root causes and potential solutions for the production of this radionuclide, including materials selection and inert gas sparging. A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750°C. Results of the diffusion model are presented for one steadystate value of tritium production in the reactor.

  20. Effect of reactor's positions on polymerization and degradation in an ultrasonic field.

    PubMed

    Kobayashi, Daisuke; Matsumoto, Hideyuki; Kuroda, Chiaki

    2008-03-01

    Ultrasonic generators are used as emulsifiers and efficient alternative initiators in polymerization processes. In this study, the effects of reactor's position on the emulsion polymerization of styrene under indirect ultrasonic irradiation were investigated, along with the effects of reactor's position on chemical and physical degradation. Both polymer yield and molecular weight were influenced by the position of the reactor. The ultrasonic irradiation could be divided into three stages, and the molecular weight of the polymer was influenced by polymerization and degradation processes. It was found that the extent of radical generation estimated by KI oxidation dosimetry and the shock wave index obtained from studies of degradation of standard polymer were useful for controlling the characteristics of the polymer generated.

  1. Preliminary evaluation of the coal resources for part of the Wilcox Group (Paleocene through Eocene), central Texas

    USGS Publications Warehouse

    Warwick, Peter D.; Aubourg, Claire E.; Suitt, Stephen E.; Podwysocki, Steven M.; Schultz, Adam C.

    2002-01-01

    The Wilcox Group of central Texas contains shallow (<500 ft) coal deposits that are mined for use in mine-mouth electric power generating plants. These coal deposits range in apparent rank from lignite to sub-bituminous (Tewalt, 1986), and are similar in rank and composition to shallow coal deposits in the northeast and south Texas areas (fig. 1). The coal zones and associated strata in the central Texas study area generally dip to the southeast toward the Gulf of Mexico coastline and basin center. The central Texas resource assessment area includes parts of eight counties (fig. 2). The assessment area was selected to encompass current mining areas and because of the availability of subsurface stratigraphic data in the area. The assessment area is roughly 160 miles long and 5 to 25 miles wide and generally follows the outcrop of the Paleocene - Eocene Wilcox Group in central Texas (figs. 1 and 2). Approximately 1,800 subsurface stratigraphic records from rotary and core drill holes were used to assess the resources of the central Texas assessment area. Of the 1,800 drill holes, only 168 are public data points and are primarily located in the areas that have been permitted for surface mining (fig. 2; Appendix 1). The remaining 1632 drill holes, which are distributed throughout the assessment area, were provided to the U.S. Geological Survey (USGS) on a confidential basis by various coal companies for use in regional studies. Nine coal zones were identified and assessed in the central Texas assessment area. Several other coal zones (as many as 9 unassessed zones) were identified but were not assessed due to the thinness of the coal beds or the lack of deep stratigraphic data (fig. 3). A total of 7.7 billion short tons of coal was identified in this assessment that excluded the resources within current coal mine lease areas (fig. 2). Corresponding maps were constructed to show the overburden, structure contour of the top of the coal zone, and cumulative coal

  2. Process and apparatus for adding and removing particles from pressurized reactors

    DOEpatents

    Milligan, John D.

    1983-01-01

    A method for adding and removing fine particles from a pressurized reactor is provided, which comprises connecting the reactor to a container, sealing the container from the reactor, filling the container with particles and a liquid material compatible with the reactants, pressurizing the container to substantially the reactor pressure, removing the seal between the reactor and the container, permitting particles to fall into or out of the reactor, and resealing the container from the reactor. An apparatus for adding and removing particles is also disclosed.

  3. Comparison of Irradiation Conditions of VVER-1000 Reactor Pressure Vessel and Surveillance Specimens for Various Core Loadings

    NASA Astrophysics Data System (ADS)

    Bukanov, V. N.; Diemokhin, V. L.; Grytsenko, O. V.; Vasylieva, O. G.; Pugach, S. M.

    2009-08-01

    The comparative analysis of irradiation conditions of surveillance specimens and pressure vessel of VVER-1000 reactor has been carried out for various configurations of the core. It is proved the fluences onto specimens and a pressure vessel don't correlate with each other but only the spectral indexes do. It is revealed that in the case of the specimen reconstitution technique application the data on the assembly orientation to the reactor core is sufficient to complete four representative groups from the samples of any container assembly. It is shown that the standard surveillance program of VVER-1000 allows obtaining reliable information on the reactor pressure vessel state.

  4. Integral experiment information for fast reactors: Sensitivity and uncertainty analysis of reactor performance parameters

    SciTech Connect

    Collins, P.J.

    1982-01-01

    This chapter offers a detailed analysis of uncertainties in experimental parameters for the ZPR benchmark cores. Discusses the critical facilities and measurements; the need for well documented data; the relevance of data for reactor design; uses of integral data; benchmark data; mockup cores; accuracy of experimental data; critical mass; reaction rate ratios; covariance matrices; selection of reliable integral data; cavity measurements; and the SCHERZO 556 core. Points out that substantial revisions of data in the CSEWG benchmark book have resulted from a reevaluation of analytical corrections using modern methods and codes. Concludes that the integral data presently being utilized represent a very limited base, which will be enlarged considerably before application to a wider range of power reactor parameters.

  5. Renewing Liquid Fueled Molten Salt Reactor Research and Development

    NASA Astrophysics Data System (ADS)

    Towell, Rusty; NEXT Lab Team

    2016-09-01

    Globally there is a desperate need for affordable, safe, and clean energy on demand. More than anything else, this would raise the living conditions of those in poverty around the world. An advanced reactor that utilizes liquid fuel and molten salts is capable of meeting these needs. Although, this technology was demonstrated in the Molten Salt Reactor Experiment (MSRE) at ORNL in the 60's, little progress has been made since the program was cancelled over 40 years ago. A new research effort has been initiated to advance the technical readiness level of key reactor components. This presentation will explain the motivation and initial steps for this new research initiative.

  6. Instrumentation and control improvements at Experimental Breeder Reactor II

    SciTech Connect

    Christensen, L.J.; Planchon, H.P.

    1993-01-01

    The purpose of this paper is to describe instrumentation and control (I C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I C systems of the next generation of liquid metal reactor (LMR) plants.

  7. Instrumentation and control improvements at Experimental Breeder Reactor II

    SciTech Connect

    Christensen, L.J.; Planchon, H.P.

    1993-03-01

    The purpose of this paper is to describe instrumentation and control (I&C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I&C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I&C systems of the next generation of liquid metal reactor (LMR) plants.

  8. Autonomous Control and Diagnostics of Space Reactor Systems

    SciTech Connect

    Upadhyaya, B.R.; Xu, X.; Perillo, S.R.P.; Na, M.G.

    2006-07-01

    This paper describes three key features of the development of an autonomous control strategy for space reactor systems. These include the development of a reactor simulation model for transient analysis, development of model-predictive control as part of the autonomous control strategy, and a fault detection and isolation module. The latter is interfaced with the control supervisor as part of a hierarchical control system. The approach has been applied to the nodal model of the SP-100 reactor with a thermo-electric generator. The results of application demonstrate the effectiveness of the control approach and its ability to reconfigure the control mode under fault conditions. (authors)

  9. PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.

    SciTech Connect

    Carew, J.; Hanson, A.; Xu, J.; Rorer, D.; Diamond, D.

    2003-08-26

    Detailed reactor physics and safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional MCNP Monte Carlo neutron and photon transport calculations were performed to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model including the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the core power transient is terminated

  10. Increasing Fuel Utilization of Breed and Burn Reactors

    NASA Astrophysics Data System (ADS)

    Di Sanzo, Christian Diego

    Breed and Burn reactors (B&B), also referred to Traveling Wave Reactors, are fast spectrum reactors that can be fed indefinitely with depleted uranium only, once criticality is achieved without the need for fuel reprocessing. Radiation damage to the fuel cladding limits the fuel utilization of B&B reactors to ˜ 18-20% FIMA (Fissions of Initial Metal Atoms) -- the minimum burnup required for sustaining the B&B mode of operation. The fuel discharged from this type of cores contain ˜ 10% fissile plutonium. Such a high plutonium content poses environmental and proliferation concerns, but makes it possible to utilize the fuel for further energy production. The objectives of the research reported in this dissertation are to analyze the fuel cycle of B&B reactors and study new strategies to extend the fuel utilization beyond ˜ 18-20% FIMA. First, the B&B reactor physics is examined while recycling the fuel every 20% FIMA via a limited separation processing, using either the melt refining or AIROX dry processes. It was found that the maximum attainable burnup varies from 54% to 58% FIMA -- depending on the recycling process and on the fraction of neutrons lost via leakage and reactivity control. In Chapter 3 the discharge fuel characteristics of B&B reactors operating at 20% FIMA and 55% FIMA is analyzed and compared. It is found that the 20% FIMA reactor discharges a fuel with about ˜ 80% fissile plutonium over total plutonium content. Subsequently a new strategy of minimal reconditioning, called double cladding is proposed to extend the fuel utilization in specifically designed second-tier reactors. It is found that with this strategy it is possible to increase fuel utilization to 30% in a sodium fast reactor and up to 40% when a subcritical B&B core is driven by an accelerator-driven spallation neutron source. Lastly, a fuel cycle using Pressurized Water Reactors (PWR) to reduce the plutonium content of discharged B&B reactors is analyzed. It was found that it is

  11. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    SciTech Connect

    Carbajo, Juan; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Schmidt, Rodney Cannon; Thomas, Justin; Wei, Tom; Sofu, Tanju; Ludewig, Hans; Tobita, Yoshiharu; Ohshima, Hiroyuki; Serre, Frederic

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  12. Decontamination and decommissioning of Shippingport commercial reactor

    SciTech Connect

    Schreiber, J.

    1989-11-01

    To a certain degree, the decontamination and decommissioning (D and D) of the Shippingport reactor was a joint venture with Duquesne Light Company. The structures that were to be decommissioned were to be removed to at least three feet below grade. Since the land had been leased from Duquesne Light, there was an agreement with them to return the land to them in a radiologically safe condition. The total enclosure volume for the steam and nuclear containment systems was about 1.3 million cubic feet, more than 80% of which was below ground. Engineering plans for the project were started in July of 1980 and the final environmental impact statement (EIS) was published in May of 1982. The plant itself was shut down in October of 1982 for end-of-life testing and defueling. The engineering services portion of the decommissioning plans was completed in September of 1983. DOE moved onto the site and took over from the Navy in September of 1984. Actual physical decommissioning began after about a year of preparation and was completed about 44 months later in July of 1989. This paper describes the main parts of D and D.

  13. Neutron flux spectra and radiation damage parameters for the Russian Bor-60 and SM-2 reactors

    SciTech Connect

    Karasiov, A.V.; Greenwood, L.R.

    1995-04-01

    The objective is to compare neutron irradiation conditions in Russian reactors and similar US facilities. Neutron fluence and spectral information and calculated radiation damage parameters are presented for the BOR-60 (Fast Experimental Reactor - 60 MW) and SM-2 reactors in Russia. Their neutron exposure characteristics are comparable with those of the Experimental Breeder Reactor (ERB-II), the Fast Flux Test Facility (FFTF), and the High Flux Isotope Reactor (HFIR) in the United States.

  14. Numerical study of the effects of lamp configuration and reactor wall roughness in an open channel water disinfection UV reactor.

    PubMed

    Sultan, Tipu

    2016-07-01

    This article describes the assessment of a numerical procedure used to determine the UV lamp configuration and surface roughness effects on an open channel water disinfection UV reactor. The performance of the open channel water disinfection UV reactor was numerically analyzed on the basis of the performance indictor reduction equivalent dose (RED). The RED values were calculated as a function of the Reynolds number to monitor the performance. The flow through the open channel UV reactor was modelled using a k-ε model with scalable wall function, a discrete ordinate (DO) model for fluence rate calculation, a volume of fluid (VOF) model to locate the unknown free surface, a discrete phase model (DPM) to track the pathogen transport, and a modified law of the wall to incorporate the reactor wall roughness effects. The performance analysis was carried out using commercial CFD software (ANSYS Fluent 15.0). Four case studies were analyzed based on open channel UV reactor type (horizontal and vertical) and lamp configuration (parallel and staggered). The results show that lamp configuration can play an important role in the performance of an open channel water disinfection UV reactor. The effects of the reactor wall roughness were Reynolds number dependent. The proposed methodology is useful for performance optimization of an open channel water disinfection UV reactor.

  15. SP-100 Program: space reactor system and subsystem investigations

    SciTech Connect

    Harty, R.B.

    1983-09-30

    For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs.

  16. Standards and Standard Setting and the Post School Curriculum

    ERIC Educational Resources Information Center

    Young, Michael

    2014-01-01

    This paper is concerned with the role of standards and standard setting in shaping the expansion of post school education in highly unequal society. It draws on an account of the debates and policies on standards in the UK from the 1980's to today and the wider lessons that can be learned from them. It argues that relying on any type of standards…

  17. Design and testing of integrated circuits for reactor protection channels

    SciTech Connect

    Battle, R.E.; Vandermolen, R.I.; Jagadish, U.; Swail, B.K.; Naser, J.

    1995-06-01

    Custom and semicustom application-specific integrated circuit design and testing methods are investigated for use in research and commercial nuclear reactor safety systems. The Electric Power Research Institute and Oak Ridge National Laboratory are working together through a cooperative research and development agreement to apply modern technology to a nuclear reactor protection system. The purpose of this project is to demonstrate to the nuclear industry an alternative approach for new or upgrade reactor protection and safety system signal processing and voting logic. Motivation for this project stems from (1) the difficulty of proving that software-based protection systems are adequately reliable, (2) the obsolescence of the original equipment, and (3) the improved performance of digital processing. A demonstration model for protection system of PWR reactor has been designed and built.

  18. ESDIS Standards Office (ESO): Requirements, Standards and Practices

    NASA Technical Reports Server (NTRS)

    Mitchell, Andrew E.; Mcinerney, Mark Allen; Enloe, Yonsok K.; Conover, Helen T.; Doyle, Allan

    2016-01-01

    The ESDIS Standards Office assists the ESDIS Project in formulating standards policy for NASA Earth Science Data Systems (ESDS), coordinates standards activities within ESDIS, and provides technical expertise and assistance with standards related tasks within the NASA Earth Science Data System Working Groups (ESDSWG). This poster summarizes information found on the earthdata.nasa.gov site that describes the ESO.

  19. Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles

    SciTech Connect

    Yin, Shengjun; Bass, Bennett Richard; Stevens, Gary; Kirk, Mark

    2011-01-01

    This paper describes stress analysis and fracture mechanics work performed to assess boiling water reactor (BWR) and pressurized water reactor (PWR) nozzles located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Various RPV nozzle geometries were investigated: 1. BWR recirculation outlet nozzle; 2. BWR core spray nozzle3 3. PWR inlet nozzle; ; 4. PWR outlet nozzle; and 5. BWR partial penetration instrument nozzle. The above nozzle designs were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-license (EOL) to require evaluation as part of establishing the allowed limits on heatup, cooldown, and hydrotest (leak test) conditions. These nozzles analyzed represent one each of the nozzle types potentially requiring evaluation. The purpose of the analyses performed on these nozzle designs was as follows: To model and understand differences in pressure and thermal stress results using a two-dimensional (2-D) axi-symmetric finite element model (FEM) versus a three-dimensional (3-D) FEM for all nozzle types. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated; To verify the accuracy of a selected linear elastic fracture mechanics (LEFM) hand solution for stress intensity factor for a postulated nozzle corner crack for both thermal and pressure loading for all nozzle types; To assess the significance of attached piping loads on the stresses in the nozzle corner region; and To assess the significance of applying pressure on the crack face with respect to the stress intensity factor for a postulated nozzle corner crack.

  20. A reload and startup plan for conversion of the NIST research reactor

    SciTech Connect

    D. J. Diamond

    2016-03-31

    The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts. The reload portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.

  1. Biodiesel production from palm oil using combined mechanical stirred and ultrasonic reactor.

    PubMed

    Choedkiatsakul, I; Ngaosuwan, K; Cravotto, G; Assabumrungrat, S

    2014-07-01

    This paper investigates the production of biodiesel from palm oil using a combined mechanical stirred and ultrasonic reactor (MS-US). The incorporation of mechanical stirring into the ultrasonic reactor explored the further improvement the transesterification of palm oil. Initial reaction rate values were 54.1, 142.9 and 164.2 mmol/L min for the mechanical-stirred (MS), ultrasonic (US) and MS-US reactors, respectively. Suitable methanol to oil molar ratio and the catalyst loading values were found to be 6 and 1 of oil, respectively. The effect of ultrasonic operating parameters; i.e. frequency, location, and number of transducer, has been investigated. Based on the conversion yield at the reactor outlet after 1 h, the number of transducers showed a relevant role in the reaction rate. Frequency and transducer location would appear to have no significant effect. The properties of the obtained biodiesel (density, viscosity, pour point, and flash point) satisfy the ASTM standard. The combined MS-US reactors improved the reaction rate affording the methyl esters in higher yield.

  2. The concerted calculation of the BN-600 reactor for the deterministic and stochastic codes

    NASA Astrophysics Data System (ADS)

    Bogdanova, E. V.; Kuznetsov, A. N.

    2017-01-01

    The solution of the problem of increasing the safety of nuclear power plants implies the existence of complete and reliable information about the processes occurring in the core of a working reactor. Nowadays the Monte-Carlo method is the most general-purpose method used to calculate the neutron-physical characteristic of the reactor. But it is characterized by large time of calculation. Therefore, it may be useful to carry out coupled calculations with stochastic and deterministic codes. This article presents the results of research for possibility of combining stochastic and deterministic algorithms in calculation the reactor BN-600. This is only one part of the work, which was carried out in the framework of the graduation project at the NRC “Kurchatov Institute” in cooperation with S. S. Gorodkov and M. A. Kalugin. It is considering the 2-D layer of the BN-600 reactor core from the international benchmark test, published in the report IAEA-TECDOC-1623. Calculations of the reactor were performed with MCU code and then with a standard operative diffusion algorithm with constants taken from the Monte - Carlo computation. Macro cross-section, diffusion coefficients, the effective multiplication factor and the distribution of neutron flux and power were obtained in 15 energy groups. The reasonable agreement between stochastic and deterministic calculations of the BN-600 is observed.

  3. Stability and harmonics in thyristor controlled reactors

    SciTech Connect

    Bohmann, L.J.; Lasseter, R.H. )

    1990-04-01

    Harmonics that arise from the interaction of thyristor controlled reactors (TCRs) and power systems can sometimes cause stability problems that are difficult to analyze since the harmonics are affected by the power system. The classical method for calculating harmonics is to calculate the harmonic current assuming an infinite bus at the high side of the TCR transformer. This current is then used as a harmonic current source on the ac system. The basic problem with this method is that many of the interactions between the ac system and the TCR are neglected. In this paper two methods for studying the neglected interactions are described. The first uses state variables to analyze the circuit containing the TCR. The resulting equations are linear differential equations with periodic coefficients. This formulation allows the study of stability, periodic operation, and resonance which can not be achieved by other methods. The second method uses a fourier matrix description of the TCR. In this model the coupling between the different harmonics due to the switching is clearly shown.

  4. Code System for Reactor Physics and Fuel Cycle Simulation.

    SciTech Connect

    TEUCHERT, E.

    1999-04-21

    Version 00 VSOP94 (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shut-down features, in-core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. In addition to its use in research and development work for the High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors.

  5. Ceramic oxygen transport membrane array reactor and reforming method

    DOEpatents

    Kelly, Sean M.; Christie, Gervase Maxwell; Rosen, Lee J.; Robinson, Charles; Wilson, Jamie R.; Gonzalez, Javier E.; Doraswami, Uttam R.

    2016-09-27

    A commercially viable modular ceramic oxygen transport membrane reforming reactor for producing a synthesis gas that improves the thermal coupling of reactively-driven oxygen transport membrane tubes and catalyst reforming tubes required to efficiently and effectively produce synthesis gas.

  6. Modeling of Reactor Kinetics and Dynamics

    SciTech Connect

    Matthew Johnson; Scott Lucas; Pavel Tsvetkov

    2010-09-01

    In order to model a full fuel cycle in a nuclear reactor, it is necessary to simulate the short time-scale kinetic behavior of the reactor as well as the long time-scale dynamics that occur with fuel burnup. The former is modeled using the point kinetics equations, while the latter is modeled by coupling fuel burnup equations with the kinetics equations. When the equations are solved simultaneously with a nonlinear equation solver, the end result is a code with the unique capability of modeling transients at any time during a fuel cycle.

  7. Energy Codes and Standards: Facilities

    SciTech Connect

    Bartlett, Rosemarie; Halverson, Mark A.; Shankle, Diana L.

    2007-01-01

    Energy codes and standards play a vital role in the marketplace by setting minimum requirements for energy-efficient design and construction. They outline uniform requirements for new buildings as well as additions and renovations. This article covers basic knowledge of codes and standards; development processes of each; adoption, implementation, and enforcement of energy codes and standards; and voluntary energy efficiency programs.

  8. Standard and Regional Standard Speech: Distinctions and Similarities.

    ERIC Educational Resources Information Center

    Edwards, J.; Jacobsen, M.

    1987-01-01

    Differential evaluations of speech usually occur along the standard-nonstandard dimension. Standard accent rates highly in regard to status and competence but low on the dimensions of integrity and attractiveness. In a Canadian context, however, a regional standard (mainland Nova Scotia) compared favorably or equally in all dimensions to other…

  9. 75 FR 14643 - Office of New Reactors; Proposed Standard Review Plan, Branch Technical Position 7-19 on Guidance...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-03-26

    ... Nuclear Reactor Regulation (NRR) staff in the review of applications for license amendments in currently..., Office of New Reactors, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone at 301... From the Federal Register Online via the Government Publishing Office NUCLEAR...

  10. Design and Performance Validation of a Conductively Heated Sealed-Vessel Reactor for Organic Synthesis.

    PubMed

    Obermayer, David; Znidar, Desiree; Glotz, Gabriel; Stadler, Alexander; Dallinger, Doris; Kappe, C Oliver

    2016-12-02

    A newly designed robust and safe laboratory scale reactor for syntheses under sealed-vessel conditions at 250 °C maximum temperature and 20 bar maximum pressure is presented. The reactor employs conductive heating of a sealed glass vessel via a stainless steel heating jacket and implements both online temperature and pressure monitoring in addition to magnetic stirring. Reactions are performed in 10 mL borosilicate vials that are sealed with a silicone cap and Teflon septum and allow syntheses to be performed on a 2-6 mL scale. This conductively heated reactor is compared to a standard single-mode sealed-vessel microwave instrument with respect to heating and cooling performance, stirring efficiency, and temperature and pressure control. Importantly, comparison of the reaction outcome for a number of different synthetic transformations performed side by side in the new device and a standard microwave reactor suggest that results obtained using microwave conditions can be readily mimicked in the operationally much simpler and smaller conventionally heated device.

  11. Embedding Materials and Economy for a Deep Underground Reactor

    SciTech Connect

    Hiroshi Takahashi

    2002-07-01

    I proposed embedding the high-conversion LWR, studied in the NERI program, about 500-1000 meters deep underground. At such depths, the earth's gravity force passively removes heat using the natural circulation of the reactor coolant; then, even a nuclear-power plant with very tight-lattice fuel assembly can be operated safely. Safety is ensured by embedding the reactor vessel and other components, such as coolant ducts, in casing containers and filling the space between the container and the vessel with embedding material. I describe suitable embedding materials that can be easily removed to allow access to the reactor and coolant components. Finally, I discuss the key economic aspects of building a reactor deep underground. (author)

  12. Secrecy, Simultaneous Discovery, and the Theory of Nuclear Reactors

    ERIC Educational Resources Information Center

    Weart, Spencer

    1977-01-01

    Discusses the simultaneous discovery of the four-factor formula in various countries, the influence of secrecy in preventing the sharing of discovery, and the resultant direction in the development of nuclear reactor theory. (SL)

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

    SciTech Connect

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

    2012-07-30

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

  14. Looking North at Reactor Number One and Air Vent on ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    Looking North at Reactor Number One and Air Vent on Fourth Floor of Oxide Building - Hematite Fuel Fabrication Facility, Oxide Building & Oxide Loading Dock, 3300 State Road P, Festus, Jefferson County, MO

  15. Uranium ARC Fission Reactor for Space Power and Propulsion

    DTIC Science & Technology

    1992-03-01

    thruster or MHD accelerator/generator. Uranium arc technology is being developed for use in space nuclear thermal and electric propulsion reactors. In...specific impulse propulsion or ultrahigh temperature power conversion. Fission events in the nuclear arc plasma provide for additional dissociation and...I Technical Objectives 3 2. URANIUM ARC FISSION REACTOR CONCEPT AND NUCLEAR -AUGMENTED THRUSTER CONCEPT 4 2.1 Physics Basis 4 2.2 Uranium Arc

  16. Fatigue and environmentally assisted cracking in light water reactors

    SciTech Connect

    Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Shack, W.J.

    1991-12-01

    Fatigue and environmentally assisted cracking of piping, pressure vessels, and core components in light water reactors (LWRs) are important concerns as extended reactor lifetimes are envisaged. The degradation processes include intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or SCC cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Similar cracking has also occurred in upper shell-to-transition cone girth welds in pressurized water reactor (PWR) steam generator vessels. Another concern is failure of reactor-core internal components after accumulation of relatively high fluence, which has occurred in both BWRs and PWRs. Research during the past year focused on (1) fatigue and SCC of ferritic steels used in piping and in steam generator and reactor pressure vessels, (2) role of chromate and sulfate in simulated BWR water in SCC of sensitized Type 304 SS, and (3) irradiation-assisted SCC in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs. Failure after accumulation of relatively high fluence has been attributed to radiation-induced segregation (RIS) of elements such as Si, P, Ni, and Cr. This document provides a summary of research progress in these areas.

  17. METHOD AND APPARATUS FOR REACTOR SAFETY CONTROL

    DOEpatents

    Huston, N.E.

    1961-06-01

    A self-contained nuclear reactor fuse controlled device tron absorbing material, normally in a compact form but which can be expanded into an extended form presenting a large surface for neutron absorption when triggered by an increase in neutron flux, is described.

  18. A study on naphtha catalytic reforming reactor simulation and analysis.

    PubMed

    Liang, Ke-min; Guo, Hai-yan; Pan, Shi-wei

    2005-06-01

    A naphtha catalytic reforming unit with four reactors in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reactions characteristics based on idealizing the complex naphtha mixture by representing the paraffin, naphthene, and aromatic groups by single compounds. The simulation results based above models agree very well with actual operation unit data.

  19. A study on naphtha catalytic reforming reactor simulation and analysis

    PubMed Central

    Liang, Ke-min; Guo, Hai-yan; Pan, Shi-wei

    2005-01-01

    A naphtha catalytic reforming unit with four reactors in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reactions characteristics based on idealizing the complex naphtha mixture by representing the paraffin, naphthene, and aromatic groups by single compounds. The simulation results based above models agree very well with actual operation unit data. PMID:15909350

  20. Fast Breeder Reactors in Sweden: Vision and Reality.

    PubMed

    Fjaestad, Maja

    2015-01-01

    The fast breeder is a type of nuclear reactor that aroused much attention in the 1950s and '60s. Its ability to produce more nuclear fuel than it consumes offered promises of cheap and reliable energy. Sweden had advanced plans for a nuclear breeder program, but canceled them in the middle of the 1970s with the rise of nuclear skepticism. The article investigates the nuclear breeder as a technological vision. The nuclear breeder reactor is an example of a technological future that did not meet its industrial expectations. But that does not change the fact that the breeder was an influential technology. Decisions about the contemporary reactors were taken with the idea that in a foreseeable future they would be replaced with the efficient breeder. The article argues that general themes in the history of the breeder reactor can deepen our understanding of the mechanisms behind technological change.

  1. Reactor Antineutrino Flux and Spectrum Shape from Daya Bay

    NASA Astrophysics Data System (ADS)

    Napolitano, Jim; Daya Bay Collaboration

    2017-01-01

    The Daya Bay Reactor Neutrino Experiment has collected very large samples of νe p ->e+ n events, where the νe are from the cores of six power plant reactors that undergo regular refueling. With 621 days of data, more than 1.2 million events of this type were detected. The collaboration has analyzed these data in terms of the absolute flux (addressing the ``Reactor Neutrino Anomaly''), the spectrum shape (including the excess in the region of 5 MeV prompt energy), and other effects. This talk will summarize the results from our most recent analyses, and discuss new initiatives aimed at continuing to understand the fine detail of the reactor νe spectrum.

  2. NEUTRONIC REACTOR

    DOEpatents

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

    1961-11-21

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

  3. Nuclear Reactors. Revised.

    ERIC Educational Resources Information Center

    Hogerton, John F.

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

  4. REACTOR COOLING

    DOEpatents

    Quackenbush, C.F.

    1959-09-29

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

  5. Status and Value of International Standards for Nuclear Criticality Safety

    SciTech Connect

    Hopper, Calvin Mitchell

    2011-01-01

    This presentation provides an update to the author's standards report provided at the ICNC-2007 meeting. It includes a discussion about the difference between, and the value of participating in, the development of international 'consensus' standards as opposed to nonconsensus standards. Standards are developed for a myriad of reasons. Generally, standards represent an agreed upon, repeatable way of doing something as defined by an individual or group of people. They come in various types. Examples include personal, family, business, industrial, commercial, and regulatory such as military, community, state, federal, and international standards. Typically, national and international 'consensus' standards are developed by individuals and organizations of diverse backgrounds representing the subject matter users and developers of a service or product and other interested parties or organizations. Within the International Organization for Standardization (ISO), Technical Committee 85 (TC85) on nuclear energy, Subcommittee 5 (SC5) on nuclear fuel technology, there is a Working Group 8 (WG8) on standardization of calculations, procedures, and practices related to criticality safety. WG8 has developed, and is developing, ISO standards within the category of nuclear criticality safety of fissionable materials outside of reactors (i.e., nonreactor fissionable material nuclear fuel cycle facilities). Since the presentation of the ICNC-2007 report, WG8 has issued three new finalized international standards and is developing two more new standards. Nearly all elements of the published WG8 ISO standards have been incorporated into IAEA nonconsensus guides and standards. The progression of consensus standards development among international partners in a collegial environment establishes a synergy of different concepts that broadens the perspectives of the members. This breadth of perspectives benefits the working group members in their considerations of consensus standards

  6. A reactor for high-temperature pyrolysis and oxygen isotopic analysis of cellulose via induction heating.

    PubMed

    Evans, Michael N

    2008-07-01

    A reactor for converting cellulose into carbon monoxide for subsequent oxygen isotopic analysis via continuous flow isotope ratio mass spectrometry is described. The system employs an induction heater to produce temperatures >or=1500 degrees C within a molybdenum foil crucible positioned by boron nitride (BN) spacers within a quartz outer sleeve. For samples of a homogeneous working standard cellulose between 300 and 400 microg in size, the blank/signal ratio is <5%, and the long-term precision is 0.30 per thousand (N = 232). For samples of 30 to 100 microg in size, a gas pressure sintered silicon nitride (Si(3)N(4)) outer sleeve replaces the quartz sleeve, the BN spacers are not used, and 6.0-grade carrier He must be used to minimize the blank signal. With these modifications a blank/sample ratio of <5% and long-term precision of 0.30 per thousand (N = 144) are obtained. These results are similar to those achieved using standard high-temperature furnaces, but the reactor is simpler to pack, the system is more economical to run, and samples as small as 30 microg cellulose may be measured. For both reactors memory is significant in the subsequent sample and is believed to be due to exchange with reactor oxygen at temperatures above 1000 degrees C. Further applications might include online preparation of other materials requiring temperatures of 1500-2600 degrees C.

  7. Molten metal reactor and method of forming hydrogen, carbon monoxide and carbon dioxide using the molten alkaline metal reactor

    DOEpatents

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

    2012-11-13

    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.

  8. NEUTRONIC REACTOR WITH ACCESSIBLE THIMBLE AND EMERGENCY COOLING FEATURES

    DOEpatents

    McCorkle, W.H.

    1960-02-23

    BS>A safety system for a water-moderated reactor is described. The invention comprises a reservoir system for spraying the fuel elements within a fuel assembly with coolant and keeping them in a continuous bath even if the coolant moderator is lost from the reactor vessel. A reservoir gravity feeds one or more nozzels positioned within each fuel assembly which continually forces water past the fuel elements.

  9. Inverse-square law violation and reactor antineutrino anomaly

    NASA Astrophysics Data System (ADS)

    Naumov, D. V.; Naumov, V. A.; Shkirmanov, D. S.

    2017-01-01

    We discuss a possibility that the so-called reactor antineutrino anomaly can be, at least in part, explained by applying a quantum field-theoretical approach to neutrino oscillations, which in particular predicts a small deviation from the classical inverse-square law at short but macroscopic distances between the neutrino source and detector. An extensive statistical analysis of the reactor data is performed to examine this speculation.

  10. High Temperature Gas-Cooled Test Reactor Options Status Report

    SciTech Connect

    Sterbentz, James William; Bayless, Paul David

    2015-08-01

    Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

  11. Standards and Assessment. IDRA Focus.

    ERIC Educational Resources Information Center

    IDRA Newsletter, 1997

    1997-01-01

    This newsletter includes three articles, two of which focus on standards for student evaluation and for admission to higher education. "A Measuring Stick for Standards and TEKS: Meeting the Needs of Second Language Learners" (Laura Chris Green, Adela Solis) examines beliefs embodied in the notion of standards; defines content,…

  12. Microchannel CO Methanation Reactors for Martian and Lunar ISRU

    SciTech Connect

    Dagle, Robert A.; Wegeng, Robert S.

    2008-07-05

    [Abstract] Microhannel technology offer significant advantages as a highly effective route to process intensification. NASA applications where size, weight, and process efficiency are key performance metrics, are very suitable applications. In situ resource utilization (ISRU) technologies is one such example. For use in a lunar carbothermal reduction process system a catalytic microchannel CO methanation reactor was developed. Design characteristics and operating performance evaluation for a multichannel reactor, approximately 4.5” X 0.5” X 1.0” in size, is discussed. Temperature, throughput, and mode of operation are variables explored. Two modes of operation were investigated: 1) under adiabatic conditions and 2) utilizing counter-current air cooling. The latter scenario offered a differential temperature profile which helped to improve performance. Demonstration of this concept is provided. Maintaining the inlet reactor temperature at 430oC the methane effluent composition output increased from 66.2% to 79.9% by introducing counter-flow air cooling. This allowed the exiting reactor temperature to decrease by ~ 100oC. Operating under such temperature differential conditions offeres exploitation of high kinetics at the hot front-end of the reactor while utilizing favorable thermodynamics at the cooler back-end. The highly efficient and compact nature of microchannel reactors make them uniquely suitable for such reaction engineering applications.

  13. Safety and control of modular liquid-metal reactors

    SciTech Connect

    Sackett, J.I. ); Sevy, R.H.; Wei, T.Y.C. )

    1989-01-01

    As part of recent development efforts on advanced reactor designs Argonne National Laboratory has proposed the integral fast reactor (IFR) concept. The IFR concept is currently being applied to modular-sized reactors, which would be built in multiple power packs together with an integrated fuel-cycle facility. It has been amply demonstrated that the concept, as applied to modular designs, has significant advantages in regard to anticipated transients without scram (ATWS). Attention is now focused on whether or not those advantages derived from the IFR traits can be translated to the operational/design-basis-accident class of transients. Inherent operability, in which reactor power control is effected through the use of primary pumps and balance-of-plant (BOP) swings rather than through the active motion of control rods, is a proposal to utilize the enhanced inherent feedback response of the IFR to improve the operating characteristics of liquid-metal reactors (LMRs). The scheme has associated with it potential advantages in the areas of plant control and design simplification. This study on inherent operability in modular LMRs therefore has implications for both operational and ATWS events. Current intentions are to analytically explore possibilities of applying various schemes to advanced LMRs with the aid of the SASSYS system code and then to test viable alternatives in the Experimental Breeder Reactor II (EBR-II) plant under the auspices of the inherent safety operability testing program.

  14. Integrated Bacillus sp. immobilized cell reactor and Synechocystis sp. algal reactor for the treatment of tannery wastewater.

    PubMed

    Sekaran, G; Karthikeyan, S; Nagalakshmi, C; Mandal, A B

    2013-01-01

    The wastewater discharged from leather industries lack biodegradability due to the presence of xenobiotic compounds. The primary clarification and aerobic treatment in Bacillus sp. immobilized Chemo Autotrophic Activated Carbon Oxidation (CAACO) reactor removed considerable amount of pollution parameters. The residual untreated organics in the wastewater was further treated in algal batch reactor inoculated with Synechocystis sp. Sodium nitrate, K(2)HPO(4), MgSO(4).7H(2)O, NH(4)Cl, CaCl(2)·2H(2)O, FeCl(3) (anhydrous), and thiamine hydrochloride, rice husk based activated carbon (RHAC), immobilization of Bacillus sp. in mesoporous activated carbon, sand filter of dimensions diameter, 6 cm and height, 30 cm; and the CAACO reactor of dimensions diameter, 5.5 cm and height, 30 cm with total volume 720 ml, and working volume of 356 ml. In the present investigation, the CAACO treated tannery wastewater was applied to Synechocystis sp. inoculated algal batch reactor of hydraulic residence time 24 h. The BOD(5), COD, and TOC of treated wastewater from algal batch reactor were 20 ± 7, 167 ± 29, and 78 ± 16 mg/l respectively. The integrated CAACO system and Algal batch reactor was operated for 30 days and they accomplished a cumulative removal of BOD(5),COD, TOC, VFA and sulphide as 98 %, 95 %, 93 %, 86 %, and 100 %, respectively. The biokinetic constants for the growth of algae in the batch reactor were specific growth rate, 0.095(day(-1)) and yield coefficient, 3.15 mg of algal biomass/mg of COD destructed. The degradation of xenobiotic compounds in the algal batch reactor was confirmed through HPLC and FT-IR techniques. The integrated CAACO-Algal reactor system established a credible reduction in pollution parameters in the tannery wastewater. The removal mechanism is mainly due to co-metabolism between algae and bacterial species and the organics were completely metabolized rather than by adsorption.

  15. Streamlined one-pass modelling

    SciTech Connect

    Jones, J.H.; Harne, R.L.; Firth, K.J.; Meyer, G.A.

    1986-01-01

    Streamlined one-pass models provide an alternative to the standard type of one-pass models for departure from nucleate boiling ratio (DNBR) studies, which require a large number of computer simulations. The standard one-pass models simulate the limiting subchannel (subchannel in which minimum DNBR occurs), a ring of subchannels around the limiting subchannel, the remainder of the limiting bundle, and the remainder of the core. Typically, these standard models use from 8 to 12 channels and approx.20 cross-flow gaps. Babcock and Wilcox (B and W) has developed streamlined one-pass models that require approximately one-half the number of channels and approx.20% of the computer resources as the standard one-pass models. These streamlined models are applicable in one-pass analysis codes such as the B and W code LYNXT and the Electric Power Research Institute VIPRE code. Streamlined one-pass models are particularly suited to extensive studies wherein computer cost and turnaround are critical, without sacrificing DNBR accuracy. Examples of such studies are reactor protection operating limit determination, limiting transient selection, and scoping studies. Moreover, streamlined one-pass models make it possible to perform DNBR studies on personal computers (PCs) using PC-adapted codes such as COBRAPC.

  16. Neutron Capture and the Antineutrino Yield from Nuclear Reactors.

    PubMed

    Huber, Patrick; Jaffke, Patrick

    2016-03-25

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ∼0.9% of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle.

  17. Fiber-Optical Sensors: Basics and Applications in Multiphase Reactors

    PubMed Central

    Li, Xiangyang; Yang, Chao; Yang, Shifang; Li, Guozheng

    2012-01-01

    This work presents a brief introduction on the basics of fiber-optical sensors and an overview focused on the applications to measurements in multiphase reactors. The most commonly principle utilized is laser back scattering, which is also the foundation for almost all current probes used in multiphase reactors. The fiber-optical probe techniques in two-phase reactors are more developed than those in three-phase reactors. There are many studies on the measurement of gas holdup using fiber-optical probes in three-phase fluidized beds, but negative interference of particles on probe function was less studied. The interactions between solids and probe tips were less studied because glass beads etc. were always used as the solid phase. The vision probes may be the most promising for simultaneous measurements of gas dispersion and solids suspension in three-phase reactors. Thus, the following techniques of the fiber-optical probes in multiphase reactors should be developed further: (1) online measuring techniques under nearly industrial operating conditions; (2) corresponding signal data processing techniques; (3) joint application with other measuring techniques.

  18. 77 FR 43542 - Cost Accounting Standards: Cost Accounting Standards 412 and 413-Cost Accounting Standards...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-07-25

    ... BUDGET Office of Federal Procurement Policy 48 CFR Part 9904 Cost Accounting Standards: Cost Accounting Standards 412 and 413--Cost Accounting Standards Pension Harmonization Rule AGENCY: Cost Accounting... correcting amendments. SUMMARY: The Office of Federal Procurement Policy (OFPP), Cost Accounting...

  19. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... advance notification of shipment of irradiated reactor fuel or nuclear waste must contain the following... irradiated reactor fuel or nuclear waste shipment; (2) A description of the irradiated reactor fuel...

  20. Method for passive cooling liquid metal cooled nuclear reactors, and system thereof

    DOEpatents

    Hunsbedt, Anstein; Busboom, Herbert J.

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel.

  1. ETR, TRA642. REACTOR FLOOR. TOP PLATES AND PLUGS ARE REMOVED. ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    ETR, TRA-642. REACTOR FLOOR. TOP PLATES AND PLUGS ARE REMOVED. CAMERA LOOKS INTO ETR TANK. TEST LOOP APPARATUS DESCENDS THROUGH REACTOR'S TEST HOLES. INL NEGATIVE NO. 60-1930. Unknown Photographer, ca. 1960 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  2. Gas core reactor concepts and technology - Issues and baseline strategy

    NASA Technical Reports Server (NTRS)

    Diaz, Nils J.; Dugan, Edward T.; Kahook, Samer; Maya, Isaac

    1991-01-01

    Results of a research program including phenomenological studies, conceptual design, and systems analysis of a series of gaseous/vapor fissile fuel driven engines for space power platforms and for thermal and electric propulsion are reviewed. It is noted that gas and vapor phase reactors provide the path for minimum mass in orbit and trip times, with a specific impulse from 1020 sec at the lowest technololgical risk to 5200 sec at the highest technological risk. The discussion covers various configurations of gas core reactors and critical technologies and the nuclear vapor thermal rocket engine.

  3. Reactor pressure vessel head vents and methods of using the same

    SciTech Connect

    Gels, John L; Keck, David J; Deaver, Gerald A

    2014-10-28

    Internal head vents are usable in nuclear reactors and include piping inside of the reactor pressure vessel with a vent in the reactor upper head. Piping extends downward from the upper head and passes outside of the reactor to permit the gas to escape or be forcibly vented outside of the reactor without external piping on the upper head. The piping may include upper and lowers section that removably mate where the upper head joins to the reactor pressure vessel. The removable mating may include a compressible bellows and corresponding funnel. The piping is fabricated of nuclear-reactor-safe materials, including carbon steel, stainless steel, and/or a Ni--Cr--Fe alloy. Methods install an internal head vent in a nuclear reactor by securing piping to an internal surface of an upper head of the nuclear reactor and/or securing piping to an internal surface of a reactor pressure vessel.

  4. A Note on Standard Deviation and Standard Error

    ERIC Educational Resources Information Center

    Hassani, Hossein; Ghodsi, Mansoureh; Howell, Gareth

    2010-01-01

    Many students confuse the standard deviation and standard error of the mean and are unsure which, if either, to use in presenting data. In this article, we endeavour to address these questions and cover some related ambiguities about these quantities.

  5. Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

    SciTech Connect

    Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.; Powers, Jeffrey J.

    2016-08-31

    Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

  6. Reactor design for minimizing product inhibition during enzymatic lignocellulose hydrolysis: II. Quantification of inhibition and suitability of membrane reactors.

    PubMed

    Andrić, Pavle; Meyer, Anne S; Jensen, Peter A; Dam-Johansen, Kim

    2010-01-01

    Product inhibition of cellulolytic enzymes affects the efficiency of the biocatalytic conversion of lignocellulosic biomass to ethanol and other valuable products. New strategies that focus on reactor designs encompassing product removal, notably glucose removal, during enzymatic cellulose conversion are required for alleviation of glucose product inhibition. Supported by numerous calculations this review assesses the quantitative aspects of glucose product inhibition on enzyme-catalyzed cellulose degradation rates. The significance of glucose product inhibition on dimensioning of different ideal reactor types, i.e. batch, continuous stirred, and plug-flow, is illustrated quantitatively by modeling different extents of cellulose conversion at different reaction conditions. The main operational challenges of membrane reactors for lignocellulose conversion are highlighted. Key membrane reactor features, including system set-up, dilution rate, glucose output profile, and the problem of cellobiose are examined to illustrate the quantitative significance of the glucose product inhibition and the total glucose concentration on the cellulolytic conversion rate. Comprehensive overviews of the available literature data for glucose removal by membranes and for cellulose enzyme stability in membrane reactors are given. The treatise clearly shows that membrane reactors allowing continuous, complete, glucose removal during enzymatic cellulose hydrolysis, can provide for both higher cellulose hydrolysis rates and higher enzyme usage efficiency (kg(product)/kg(enzyme)). Current membrane reactor designs are however not feasible for large scale operations. The report emphasizes that the industrial realization of cellulosic ethanol requires more focus on the operational feasibility within the different hydrolysis reactor designs, notably for membrane reactors, to achieve efficient enzyme-catalyzed cellulose degradation.

  7. An atmospheric pressure flow reactor: Gas phase kinetics and mechanism in tropospheric conditions without wall effects

    NASA Technical Reports Server (NTRS)

    Koontz, Steven L.; Davis, Dennis D.; Hansen, Merrill

    1988-01-01

    A new type of gas phase flow reactor, designed to permit the study of gas phase reactions near 1 atm of pressure, is described. A general solution to the flow/diffusion/reaction equations describing reactor performance under pseudo-first-order kinetic conditions is presented along with a discussion of critical reactor parameters and reactor limitations. The results of numerical simulations of the reactions of ozone with monomethylhydrazine and hydrazine are discussed, and performance data from a prototype flow reactor are presented.

  8. System and method for air temperature control in an oxygen transport membrane based reactor

    DOEpatents

    Kelly, Sean M

    2016-09-27

    A system and method for air temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

  9. System and method for temperature control in an oxygen transport membrane based reactor

    DOEpatents

    Kelly, Sean M.

    2017-02-21

    A system and method for temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

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

    SciTech Connect

    Jung, Y. S.; Joo, H. G.; Yoon, J. I.

    2013-07-01

    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)

  11. Hybrid adsorptive membrane reactor

    NASA Technical Reports Server (NTRS)

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

    2011-01-01

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

  12. Hybrid adsorptive membrane reactor

    SciTech Connect

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

    2011-03-01

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

  13. Advanced Light Water Reactor Program: Program management and staff review methodology

    SciTech Connect

    Moran, D.H.

    1986-12-01

    This report summarizes the NRC/EPRI coordinated effort to develop design requirements for a standardized advanced light water reactor (ALWR) and the procedures for screening and applying new generic safety issues to this program. The end-product will be an NRC-approved ALWR Requirements Document for use by the nuclear industry in generating designs of LWRs to be constructed for operation in the 1990s and beyond.

  14. Developments in Molten Salt and Liquid-Salt-Cooled Reactors

    SciTech Connect

    Forsberg, Charles W.

    2006-07-01

    In the last 5 years, there has been a rapid growth in interest in the use of high-temperature (700 to 1000 deg C) molten and liquid fluoride salts as coolants in nuclear systems. This renewed interest is a consequence of new applications for high-temperature heat and the development of new reactor concepts. Fluoride salts have melting points between 350 and 500 deg C; thus, they are of use only in high-temperature systems. Historically, steam cycles with temperature limits of {approx}550 deg C have been the only efficient method to convert heat to electricity. This limitation produced few incentives to develop high-temperature reactors for electricity production. However, recent advances in Brayton gas turbine technology now make it possible to convert higher-temperature heat efficiency into electricity on an industrial scale and thus have created the enabling technology for more efficient nuclear reactors. Simultaneously, there is a growing interest in using high-temperature nuclear heat for the production of hydrogen and shale oil. Five nuclear-related applications are being investigated: (1) liquid-salt heat-transport systems in hydrogen and shale oil production systems; (2) the advanced high-temperature reactor, which uses a graphite-matrix coated-particle fuel and a liquid salt coolant; (3) the liquid-salt-cooled fast reactor which uses metal-clad fuel and a liquid salt coolant; (4) the molten salt reactor, with the fuel dissolved in the molten salt coolant; and (5) fusion energy systems. The reasons for the new interest in liquid salt coolants, the reactor concepts, and the relevant programs are described. (author)

  15. Telecommunications Policy Research Conference. Standards and Standardization Section. Papers.

    ERIC Educational Resources Information Center

    Telecommunications Policy Research Conference, Inc., Washington, DC.

    These three papers consider various models and mechanisms for the achievement of industrial standardization. The first, "Duopoly Compatibility Standards with Partial Cooperation and Standards Leadership" (Sanford V. Berg, University of Florida) presents a model of duopolists producing differentiated substitutes, and considers the…

  16. 77 FR 43196 - Minimum Internal Control Standards and Technical Standards

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-07-24

    ... COMMISSION 25 CFR Parts 543 and 547 Minimum Internal Control Standards and Technical Standards AGENCY... . SUPPLEMENTARY INFORMATION: Part 543 addresses minimum internal control standards (MICS) for Class II gaming operations. The regulations require tribes to establish controls and implement procedures at least...

  17. University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor

    SciTech Connect

    Eric C. Woolstenhulme; Dana M. Hewit

    2008-09-01

    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.

  18. Improved Nuclear Reactor and Shield Mass Model for Space Applications

    NASA Technical Reports Server (NTRS)

    Robb, Kevin

    2004-01-01

    New technologies are being developed to explore the distant reaches of the solar system. Beyond Mars, solar energy is inadequate to power advanced scientific instruments. One technology that can meet the energy requirements is the space nuclear reactor. The nuclear reactor is used as a heat source for which a heat-to-electricity conversion system is needed. Examples of such conversion systems are the Brayton, Rankine, and Stirling cycles. Since launch cost is proportional to the amount of mass to lift, mass is always a concern in designing spacecraft. Estimations of system masses are an important part in determining the feasibility of a design. I worked under Michael Barrett in the Thermal Energy Conversion Branch of the Power & Electric Propulsion Division. An in-house Closed Cycle Engine Program (CCEP) is used for the design and performance analysis of closed-Brayton-cycle energy conversion systems for space applications. This program also calculates the system mass including the heat source. CCEP uses the subroutine RSMASS, which has been updated to RSMASS-D, to estimate the mass of the reactor. RSMASS was developed in 1986 at Sandia National Laboratories to quickly estimate the mass of multi-megawatt nuclear reactors for space applications. In response to an emphasis for lower power reactors, RSMASS-D was developed in 1997 and is based off of the SP-100 liquid metal cooled reactor. The subroutine calculates the mass of reactor components such as the safety systems, instrumentation and control, radiation shield, structure, reflector, and core. The major improvements in RSMASS-D are that it uses higher fidelity calculations, is easier to use, and automatically optimizes the systems mass. RSMASS-D is accurate within 15% of actual data while RSMASS is only accurate within 50%. My goal this summer was to learn FORTRAN 77 programming language and update the CCEP program with the RSMASS-D model.

  19. NEUTRONIC REACTOR

    DOEpatents

    Metcalf, H.E.; Johnson, H.W.

    1961-04-01

    BS>A nuclear reactor incorporating fuel rods passing through a moderator and including tubes of a material of higher Thermal conductivity than the fuel in contact with the fuel is described. The tubes extend beyond the active portion of the reactor into contant with a fiuld coolant.

  20. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1958-04-22

    A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

  1. 10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... light water nuclear power reactors which result from exposure of these materials to neutron irradiation..., “Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels”; ASTM E 185-79, “Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels”; and...

  2. 10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... light water nuclear power reactors which result from exposure of these materials to neutron irradiation..., “Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels”; ASTM E 185-79, “Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels”; and...

  3. 10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... light water nuclear power reactors which result from exposure of these materials to neutron irradiation..., “Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels”; ASTM E 185-79, “Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels”; and...

  4. Control Means for Reactor

    DOEpatents

    Manley, J. H.

    1961-06-27

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

  5. The combined hybrid system: A symbiotic thermal reactor/fast reactor system for power generation and radioactive waste toxicity reduction

    SciTech Connect

    Hollaway, W.R.

    1991-08-01

    If there is to be a next generation of nuclear power in the United States, then the four fundamental obstacles confronting nuclear power technology must be overcome: safety, cost, waste management, and proliferation resistance. The Combined Hybrid System (CHS) is proposed as a possible solution to the problems preventing a vigorous resurgence of nuclear power. The CHS combines Thermal Reactors (for operability, safety, and cost) and Integral Fast Reactors (for waste treatment and actinide burning) in a symbiotic large scale system. The CHS addresses the safety and cost issues through the use of advanced reactor designs, the waste management issue through the use of actinide burning, and the proliferation resistance issue through the use of an integral fuel cycle with co-located components. There are nine major components in the Combined Hybrid System linked by nineteen nuclear material mass flow streams. A computer code, CHASM, is used to analyze the mass flow rates CHS, and the reactor support ratio (the ratio of thermal/fast reactors), IFR of the system. The primary advantages of the CHS are its essentially actinide-free high-level radioactive waste, plus improved reactor safety, uranium utilization, and widening of the option base. The primary disadvantages of the CHS are the large capacity of IFRs required (approximately one MW{sub e} IFR capacity for every three MW{sub e} Thermal Reactor) and the novel radioactive waste streams produced by the CHS. The capability of the IFR to burn pure transuranic fuel, a primary assumption of this study, has yet to be proven. The Combined Hybrid System represents an attractive option for future nuclear power development; that disposal of the essentially actinide-free radioactive waste produced by the CHS provides an excellent alternative to the disposal of intact actinide-bearing Light Water Reactor spent fuel (reducing the toxicity based lifetime of the waste from roughly 360,000 years to about 510 years).

  6. Reactor protection system with automatic self-testing and diagnostic

    DOEpatents

    Gaubatz, D.C.

    1996-12-17

    A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically ``identical`` values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic. 16 figs.

  7. Reactor protection system with automatic self-testing and diagnostic

    DOEpatents

    Gaubatz, Donald C.

    1996-01-01

    A reactor protection system having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically "identical" values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic.

  8. Programmable Automated Welding System (PAWS)

    NASA Technical Reports Server (NTRS)

    Kline, Martin D.

    1994-01-01

    An ambitious project to develop an advanced, automated welding system is being funded as part of the Navy Joining Center with Babcock & Wilcox as the prime integrator. This program, the Programmable Automated Welding System (PAWS), involves the integration of both planning and real-time control activities. Planning functions include the development of a graphical decision support system within a standard, portable environment. Real-time control functions include the development of a modular, intelligent, real-time control system and the integration of a number of welding process sensors. This paper presents each of these components of the PAWS and discusses how they can be utilized to automate the welding operation.

  9. Pyrolysis reactor and fluidized bed combustion chamber

    DOEpatents

    Green, Norman W.

    1981-01-06

    A solid carbonaceous material is pyrolyzed in a descending flow pyrolysis reactor in the presence of a particulate source of heat to yield a particulate carbon containing solid residue. The particulate source of heat is obtained by educting with a gaseous source of oxygen the particulate carbon containing solid residue from a fluidized bed into a first combustion zone coupled to a second combustion zone. A source of oxygen is introduced into the second combustion zone to oxidize carbon monoxide formed in the first combustion zone to heat the solid residue to the temperature of the particulate source of heat.

  10. Granular flow in pebble bed reactors: Dust generation and scaling

    SciTech Connect

    Rycroft, C. H.; Lind, T.; Guentay, S.; Dehbi, A.

    2012-07-01

    In experimental prototypes of pebble bed reactors, significant quantities of graphite dust have been observed due to rubbing between pebbles as they flow through the core. At the high temperatures and pressures in these reactors, little data is available to understand the frictional properties of the pebble surfaces, and as a result, the Paul Scherrer Institut (Switzerland) proposes a conceptual design of a scaled-down version of a pebble bed reactor to investigate this issue in detail. In this paper, simulations of granular flow in pebble bed reactors using the discrete-element method are presented. Simulations in the full geometry (using 440,000 pebbles) are compared to those in geometries scaled down by 3:1 and 6:1. The simulations show complex behavior due to discrete pebble packing effects, meaning that pebble flow and dust generation in a scaled-down facility may be significantly different. The differences between velocity profiles, packing geometry, and pebble wear at the different scales are discussed. The results can aid in the design of the prototypical facility to more accurately reproduce the flow in a full-size reactor. (authors)

  11. Titer-plate formatted continuous flow thermal reactors: Design and performance of a nanoliter reactor.

    PubMed

    Chen, Pin-Chuan; Park, Daniel S; You, Byoung-Hee; Kim, Namwon; Park, Taehyun; Soper, Steven A; Nikitopoulos, Dimitris E; Murphy, Michael C

    2010-08-06

    Arrays of continuous flow thermal reactors were designed, configured, and fabricated in a 96-device (12 × 8) titer-plate format with overall dimensions of 120 mm × 96 mm, with each reactor confined to a 8 mm × 8 mm footprint. To demonstrate the potential, individual 20-cycle (740 nL) and 25-cycle (990 nL) reactors were used to perform the continuous flow polymerase chain reaction (CFPCR) for amplification of DNA fragments of different lengths. Since thermal isolation of the required temperature zones was essential for optimal biochemical reactions, three finite element models, executed with ANSYS (v. 11.0, Canonsburg, PA), were used to characterize the thermal performance and guide system design: (1) a single device to determine the dimensions of the thermal management structures; (2) a single CFPCR device within an 8 mm × 8 mm area to evaluate the integrity of the thermostatic zones; and (3) a single, straight microchannel representing a single loop of the spiral CFPCR device, accounting for all of the heat transfer modes, to determine whether the PCR cocktail was exposed to the proper temperature cycling. In prior work on larger footprint devices, simple grooves between temperature zones provided sufficient thermal resistance between zones. For the small footprint reactor array, 0.4 mm wide and 1.2 mm high fins were necessary within the groove to cool the PCR cocktail efficiently, with a temperature gradient of 15.8°C/mm, as it flowed from the denaturation zone to the renaturation zone. With temperature tolerance bands of ±2°C defined about the nominal temperatures, more than 72.5% of the microchannel length was located within the desired temperature bands. The residence time of the PCR cocktail in each temperature zone decreased and the transition times between zones increased at higher PCR cocktail flow velocities, leading to less time for the amplification reactions. Experiments demonstrated the performance of the CFPCR devices as a function of flow

  12. Titer-plate formatted continuous flow thermal reactors: Design and performance of a nanoliter reactor

    PubMed Central

    Chen, Pin-Chuan; Park, Daniel S.; You, Byoung-Hee; Kim, Namwon; Park, Taehyun; Soper, Steven A.; Nikitopoulos, Dimitris E.; Murphy, Michael C.

    2010-01-01

    Arrays of continuous flow thermal reactors were designed, configured, and fabricated in a 96-device (12 × 8) titer-plate format with overall dimensions of 120 mm × 96 mm, with each reactor confined to a 8 mm × 8 mm footprint. To demonstrate the potential, individual 20-cycle (740 nL) and 25-cycle (990 nL) reactors were used to perform the continuous flow polymerase chain reaction (CFPCR) for amplification of DNA fragments of different lengths. Since thermal isolation of the required temperature zones was essential for optimal biochemical reactions, three finite element models, executed with ANSYS (v. 11.0, Canonsburg, PA), were used to characterize the thermal performance and guide system design: (1) a single device to determine the dimensions of the thermal management structures; (2) a single CFPCR device within an 8 mm × 8 mm area to evaluate the integrity of the thermostatic zones; and (3) a single, straight microchannel representing a single loop of the spiral CFPCR device, accounting for all of the heat transfer modes, to determine whether the PCR cocktail was exposed to the proper temperature cycling. In prior work on larger footprint devices, simple grooves between temperature zones provided sufficient thermal resistance between zones. For the small footprint reactor array, 0.4 mm wide and 1.2 mm high fins were necessary within the groove to cool the PCR cocktail efficiently, with a temperature gradient of 15.8°C/mm, as it flowed from the denaturation zone to the renaturation zone. With temperature tolerance bands of ±2°C defined about the nominal temperatures, more than 72.5% of the microchannel length was located within the desired temperature bands. The residence time of the PCR cocktail in each temperature zone decreased and the transition times between zones increased at higher PCR cocktail flow velocities, leading to less time for the amplification reactions. Experiments demonstrated the performance of the CFPCR devices as a function of flow

  13. STEAM FORMING NEUTRONIC REACTOR AND METHOD OF OPERATING IT

    DOEpatents

    Untermyer, S.

    1960-05-10

    The heterogeneous reactor is liquid moderated and cooled by a steam forming coolant and is designed to produce steam from the coolant directly within the active portion of the reactor while avoiding the formation of bubbles in the liquid moderator. This reactor achieves inherent stability as a result of increased neutron leakage and increased neutron resonance absorption in the U/sup 238/ fuel with the formation of bubbles. The invention produces certain conditions under which the formation of vapor bubbles as a result of a neutron flux excursion from the injection of a reactivity increment into the reactor will operate to nullify the reactivity increment within a sufficiently short period of time to prevent unsafe reactor operating conditions from developing. This is obtained by disposing a plurality of fuel elements within a mass of steam forming coolant in the core with the ratio of the volume of steam forming coolant to the volume of fissionable isotopes being within the range yielding a multiplication factor greater than unity and a negative reactivity to core void coefficient at the boiling temperature of the coolant.

  14. Review of High Temperature Water and Steam Cooled Reactor Concepts

    SciTech Connect

    Oka, Yoshiaki

    2002-07-01

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

  15. The Virtual Environment for Reactor Applications (VERA). Design and architecture☆

    NASA Astrophysics Data System (ADS)

    Turner, John A.; Clarno, Kevin; Sieger, Matt; Bartlett, Roscoe; Collins, Benjamin; Pawlowski, Roger; Schmidt, Rodney; Summers, Randall

    2016-12-01

    VERA, the Virtual Environment for Reactor Applications, is the system of physics capabilities being developed and deployed by the Consortium for Advanced Simulation of Light Water Reactors (CASL). CASL was established for the modeling and simulation of commercial nuclear reactors. VERA consists of integrating and interfacing software together with a suite of physics components adapted and/or refactored to simulate relevant physical phenomena in a coupled manner. VERA also includes the software development environment and computational infrastructure needed for these components to be effectively used. We describe the architecture of VERA from both software and numerical perspectives, along with the goals and constraints that drove major design decisions, and their implications. We explain why VERA is an environment rather than a framework or toolkit, why these distinctions are relevant (particularly for coupled physics applications), and provide an overview of results that demonstrate the use of VERA tools for a variety of challenging applications within the nuclear industry.

  16. Basis for Interim Operation for the K-Reactor in Cold Standby

    SciTech Connect

    Shedrow, B.

    1998-10-19

    The Basis for Interim Operation (BIO) document for K Reactor in Cold Standby and the L- and P-Reactor Disassembly Basins was prepared in accordance with the draft DOE standard for BIO preparation (dated October 26, 1993).

  17. Assessment of light water reactor accident management programs and experience

    SciTech Connect

    Hammersley, R.J.

    1992-03-01

    The objective of this report is to provide an assessment of the current light water reactor experience regarding accident management programs and associated technology developments. This assessment for light water reactor (LWR) designs is provided as a resource and reference for the development of accident management capabilities for the production reactors at the Savannah River Site. The specific objectives of this assessment are as follows: 1. Perform a review of the NRC, utility, and industry (NUMARC, EPRI) accident management programs and implementation experience. 2. Provide an assessment of the problems and opportunities in developing an accident management program in conjunction or following the Individual Plant Examination process. 3. Review current NRC, utility, and industry technological developments in the areas of computational tools, severe accident predictive tools, diagnostic aids, and severe accident training and simulation.

  18. Design and testing of integrated circuits for reactor protection channels

    SciTech Connect

    Battle, R.E.; Vandermolen, R.I.; Jagadish, U.; Swail, B.K.; Naser, J.; Rana, I.

    1995-06-01

    Custom and semicustom application-specific integrated circuit design and testing methods are investigated for use in research and commercial nuclear reactor safety systems. The Electric Power Research Institute and Oak Ridge National Laboratory are working together through a cooperative research and development agreement to apply modern technology to a nuclear reactor protection system. Purpose of this project is to demonstrate to the nuclear industry an alternative approach for new or upgrade reactor protection and safety system signal processing and voting logic. Motivation for this project stems from (1) the difficulty of proving that software-based protection systems are adequately reliable, (2) the obsolescence of the original equipment, and (3) the improved performance of digital processing.

  19. 40 CFR 63.11149 - What are the standards and compliance requirements for new sources?

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... Smelting Area Sources Standards and Compliance Requirements § 63.11149 What are the standards and... include, but are not limited to, copper concentrate dryers, smelting flash furnaces, smelting bath furnaces, converting vessels, combined smelting and converting reactors, anode refining furnaces, and...

  20. Yttrium and rare earth stabilized fast reactor metal fuel

    DOEpatents

    Guon, Jerold; Grantham, LeRoy F.; Specht, Eugene R.

    1992-01-01

    To increase the operating temperature of a reactor, the melting point and mechanical properties of the fuel must be increased. For an actinide-rich fuel, yttrium, lanthanum and/or rare earth elements can be added, as stabilizers, to uranium and plutonium and/or a mixture of other actinides to raise the melting point of the fuel and improve its mechanical properties. Since only about 1% of the actinide fuel may be yttrium, lanthanum, or a rare earth element, the neutron penalty is low, the reactor core size can be reduced, the fuel can be burned efficiently, reprocessing requirements are reduced, and the nuclear waste disposal volumes reduced. A further advantage occurs when yttrium, lanthanum, and/or other rare earth elements are exposed to radiation in a reactor, they produce only short half life radioisotopes, which reduce nuclear waste disposal problems through much shorter assured-isolation requirements.

  1. BISON and MARMOT Development for Modeling Fast Reactor Fuel Performance

    SciTech Connect

    Gamble, Kyle Allan Lawrence; Williamson, Richard L.; Schwen, Daniel; Zhang, Yongfeng; Novascone, Stephen Rhead; Medvedev, Pavel G.

    2015-09-01

    BISON and MARMOT are two codes under development at the Idaho National Laboratory for engineering scale and lower length scale fuel performance modeling. It is desired to add capabilities for fast reactor applications to these codes. The fast reactor fuel types under consideration are metal (U-Pu-Zr) and oxide (MOX). The cladding types of interest include 316SS, D9, and HT9. The purpose of this report is to outline the proposed plans for code development and provide an overview of the models added to the BISON and MARMOT codes for fast reactor fuel behavior. A brief overview of preliminary discussions on the formation of a bilateral agreement between the Idaho National Laboratory and the National Nuclear Laboratory in the United Kingdom is presented.

  2. Reactor modeling in heterogeneous photocatalysis: toxicity and biodegradability assessment.

    PubMed

    Satuf, M L; José, S; Paggi, J C; Brandi, R J; Cassano, A E; Alfano, O M

    2010-01-01

    Photocatalysis employing titanium dioxide is a useful method to degrade a wide variety of organic and inorganic pollutants from water and air. However, the application of this advanced oxidation process at industrial scale requires the development of mathematical models to design and scale-up photocatalytic reactors. In the present work, intrinsic kinetic expressions previously obtained in a laboratory reactor are employed to predict the performance of a bench scale reactor of different configuration and operating conditions. 4-Chlorophenol was chosen as the model pollutant. The toxicity and biodegradability of the irradiated mixture in the bench photoreactor was also assessed. Good agreement was found between simulation and experimental data. The root mean square error of the estimations was 9.9%. The photocatalytic process clearly enhances the biodegradability of the reacting mixture, and the initial toxicity of the pollutant was significantly reduced by the treatment.

  3. Standards on Ethics and Integrity.

    ERIC Educational Resources Information Center

    Burns, Janet Z.; Dean, Peter J.; Hatcher, Tim; Otte, Fred L.; Preskill, Hallie; Russ-Eft, Darlene

    1999-01-01

    The purpose of the standards described in this article are to provide guidance and a common set of values for human resource development/human performance training (HRD/HPT) professionals. General Standards; Research and Evaluation; Advertising and Other Public Statements; Publication of Work; Privacy and Confidentiality; Teaching and…

  4. Collecting and recirculating condensate in a nuclear reactor containment

    DOEpatents

    Schultz, T.L.

    1993-10-19

    An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank. 3 figures.

  5. Software reliability and safety in nuclear reactor protection systems

    SciTech Connect

    Lawrence, J.D.

    1993-11-01

    Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor.

  6. Collecting and recirculating condensate in a nuclear reactor containment

    DOEpatents

    Schultz, Terry L.

    1993-01-01

    An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank.

  7. Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide

    SciTech Connect

    Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

    2012-09-01

    This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors’ spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

  8. Progress in reliability of fast reactor operation and new trends to increased inherent safety

    SciTech Connect

    Merk, Bruno; Stanculescu, Alexander; Chellapandi, Perumal; Hill, Robert

    2015-06-01

    The reasons for the renewed interest in fast reactors and an overview of the progress in sodium cooled fast reactor operation in the last ten years are given. The excellent operational performance of sodium cooled fast reactors in this period is highlighted as a sound basis for the development of new fast reactors. The operational performance of the BN-600 is compared and evaluated against the performance of German light water reactors to assess the reliability. The relevance of feedback effects for safe reactor design is described, and a new method for the enhancement of feedback effects in fast reactors is proposed. Experimental reactors demonstrating the inherent safety of advanced sodium cooled fast reactor designs are described and the potential safety improvements resulting from the use of fine distributed moderating material are discussed.

  9. PROSPECT: The Precision Reactor Oscillation and Spectrum Experiment

    NASA Astrophysics Data System (ADS)

    Langford, Thomas; Prospect Collaboration

    2016-09-01

    PROSPECT is a phased experiment consisting of segmented 6Li-loaded liquid scintillator antineutrino detectors designed to probe short-baseline neutrino oscillations and precisely measure the reactor antineutrino spectrum. The experiment will be located at the High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab. The first phase is a movable 3 tonne detector located 7-12 m from the compact, highly enriched uranium core. Over the past three years, PROSPECT has deployed multiple detectors at HFIR to understand the local background environment and demonstrate active and passive background rejection. Measuring the neutrino spectrum from 235U will give insight to the recent spectral discrepancies and provide an important benchmark for future reactor experiments. PROSPECT will probe the sterile neutrino best-fit region at 3 σ within one year of operation at HFIR. We will discuss the design, experimental program, and discovery potential of the experiment.

  10. Reactor concepts for bioelectrochemical syntheses and energy conversion.

    PubMed

    Krieg, Thomas; Sydow, Anne; Schröder, Uwe; Schrader, Jens; Holtmann, Dirk

    2014-12-01

    In bioelectrochemical systems (BESs) at least one electrode reaction is catalyzed by microorganisms or isolated enzymes. One of the existing challenges for BESs is shifting the technology towards industrial use and engineering reactor systems at adequate scales. Due to the fact that most BESs are usually deployed in the production of large-volume but low-value products (e.g., energy, fuels, and bulk chemicals), investment and operating costs must be minimized. Recent advances in reactor concepts for different BESs, in particular biofuel cells and electrosynthesis, are summarized in this review including electrode development and first applications on a technical scale. A better understanding of the impact of reactor components on the performance of the reaction system is an important step towards commercialization of BESs.

  11. Electroweak interaction: Standard and beyond

    SciTech Connect

    Harari, H.

    1987-02-01

    Several important topics within the standard model raise questions which are likely to be answered only by further theoretical understanding which goes beyond the standard model. In these lectures we present a discussion of some of these problems, including the quark masses and angles, the Higgs sector, neutrino masses, W and Z properties and possible deviations from a pointlike structure. 44 refs.

  12. SPERTI plot plan, showing reactor and control areas after 1956 ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    SPERT-I plot plan, showing reactor and control areas after 1956 addition to PER-601. Includes reactor-area buildings PER-605, -606, and -607; Terminal Building (PER-604), and control-area buildings PER-601, -602, -603 along with associated parking areas and fencing. Vicinity map shows relationship of SPERT-I to SPERT-II, SPERT-III, central facilities area (at west end of E. Portland Avenue) and Highways 20 and 26. Idaho Operations Office PER-103-IDO-1. Date: December 1955. INEEL index no. 760-0103-396-109112 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  13. PROSPECT: the Precision Reactor Oscillation and Spectrum Experiment

    NASA Astrophysics Data System (ADS)

    Jones, Donald; Prospect Collaboration

    2016-03-01

    PROSPECT is a phased experiment consisting of segmented Li-loaded liquid scintillator antineutrino detectors designed to probe short-baseline neutrino oscillations and precisely measure the reactor antineutrino spectrum. The experiment will be located at the High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab. The first phase is a movable 3 tonne detector located 7-9 m from the compact, highly enriched uranium core. Over the past three years, PROSPECT has deployed multiple prototype detectors at HFIR to understand the local background environment and demonstrate active and passive background rejection. A two-segment prototype has been developed that demonstrates critical subsystems of the full detector. Measuring the neutrino spectrum from 235U will give insight to the recent spectral discrepancies and provide an important benchmark for future reactor experiments. As a high statistics experiment, PROSPECT will probe the sterile neutrino best-fit region within one year of operation at HFIR.

  14. Performance and safety parameters for the high flux isotope reactor

    SciTech Connect

    Ilas, G.; Primm III, T.

    2012-07-01

    A Monte Carlo depletion model for the High Flux Isotope Reactor (HFIR) Cycle 400 and its use in calculating parameters of relevance to the reactor performance and safety during the reactor cycle are presented in this paper. This depletion model was developed to serve as a reference for the design of a low-enriched uranium (LEU) fuel for an ongoing study to convert HFIR from high-enriched uranium (HEU) to LEU fuel; both HEU and LEU depletion models use the same methodology and ENDF/B-VII nuclear data as discussed in this paper. The calculated HFIR Cycle 400 parameters, which are compared with measurement data from critical experiments performed at HFIR, data included in the HFIR Safety Analysis Report (SAR), or data reported by previous calculations, provide a basis for verification or updating of the corresponding SAR data. (authors)

  15. Coal hydrogenation and deashing in ebullated bed catalytic reactor

    DOEpatents

    Huibers, Derk T. A.; Johanson, Edwin S.

    1983-01-01

    An improved process for hydrogenation of coal containing ash with agglomeration and removal of ash from an ebullated bed catalytic reactor to produce deashed hydrocarbon liquid and gas products. In the process, a flowable coal-oil slurry is reacted with hydrogen in an ebullated catalyst bed reaction zone at elevated temperature and pressure conditions. The upward velocity and viscosity of the reactor liquid are controlled so that a substantial portion of the ash released from the coal is agglomerated to form larger particles in the upper portion of the reactor above the catalyst bed, from which the agglomerated ash is separately withdrawn along with adhering reaction zone liquid. The resulting hydrogenated hydrocarbon effluent material product is phase separated to remove vapor fractions, after which any ash remaining in the liquid fraction can be removed to produce substantially ash-free coal-derived liquid products.

  16. Continuous-flow high pressure hydrogenation reactor for optimization and high-throughput synthesis.

    PubMed

    Jones, Richard V; Godorhazy, Lajos; Varga, Norbert; Szalay, Daniel; Urge, Laszlo; Darvas, Ferenc

    2006-01-01

    This paper reports on a novel continuous-flow hydrogenation reactor and its integration with a liquid handler to generate a fully automated high-throughput hydrogenation system for library synthesis. The reactor, named the H-Cube, combines endogenous hydrogen generation from the electrolysis of water with a continuous flow-through system. The system makes significant advances over current batch hydrogenation reactors in terms of safety, reaction validation efficiency, and rates of reaction. The hydrogenation process is described along with a detailed description of the device's main parts. The reduction of a series of functional groups, varying in difficulty up to 70 degrees C and 70 bar are also described. The paper concludes with the integration of the device into an automated liquid handler followed by the reduction of a nitro compound in a high throughput manner. The system is fully automated and can conduct 5 reactions in the time it takes to perform and workup one reaction manually on a standard batch reactor.

  17. Reduced enrichment for research and test reactors: Proceedings

    SciTech Connect

    Not Available

    1988-05-01

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

  18. Corrosion and Corrosion Control in Light Water Reactors

    NASA Astrophysics Data System (ADS)

    Gordon, Barry M.

    2013-08-01

    Serious corrosion problems have plagued the light water reactor (LWR) industry for decades. The complex corrosion mechanisms involved and the development of practical engineering solutions for their mitigation will be discussed in this article. After a brief overview of the basic designs of the boiling water reactor (BWR) and pressurized water reactor (PWR), emphasis will be placed on the general corrosion of LWR containments, flow-accelerated corrosion of carbon steel components, intergranular stress corrosion cracking (IGSCC) in BWRs, primary water stress corrosion cracking (PWSCC) in PWRs, and irradiation-assisted stress corrosion cracking (IASCC) in both systems. Finally, the corrosion future of both plants will be discussed as plants extend their period of operation for an additional 20 to 40 years.

  19. Options: the JADE reactor and heat transfer by heat pipes

    SciTech Connect

    Simpson, J.E.; Massey, J.V.

    1981-08-10

    The JADE reactor is a new Lawrence Livermore National Laboratory (LLNL) concept which maintains advantages of liquid metal walls and addresses some of their problems. The concept envisions a porous medium, called the jade, of specific geometry lining the reactor cavity. The jade is designed to convert the kinetic energy of the fluid to thermal energy before it reaches the first wall. Finally, its particular geometric shape is used to minimize reaction forces on the first wall due to blow-off caused by soft x-rays and debris, to provide empty spaces for fluid expansion after neutron energy deposition where droplets collide with droplets cancelling their kinetic energies, and to provide large surface areas for rapid condensation of vapor. LLNL also suggested that heat pipes might be used to eliminate portions of the primary or secondary coolant loops, thereby reducing pumping requirements found in current reactor designs.

  20. A Pebble-Bed Breed-and-Burn Reactor

    SciTech Connect

    Greenspan, Ehud

    2016-03-31

    The primary objective of this project is to use three-dimensional fuel shuffling in order to reduce the minimum peak radiation damage of ~550 dpa present Breed-and-Burn (B&B) fast nuclear reactor cores designs (they feature 2-D fuel shuffling) call for to as close as possible to the presently accepted value of 200 dpa thereby enabling earlier commercialization of B&B reactors which could make substantial contribution to energy sustainability and economic stability without need for fuel recycling. Another objective is increasing the average discharge burnup for the same peak discharge burnup thereby (1) increasing the fuel utilization of 2-D shuffled B&B reactors and (2) reducing the reprocessing capacity required to support a given capacity of FRs that are to recycle fuel.

  1. Method and apparatus for a catalytic firebox reactor

    DOEpatents

    Smith, Lance L.; Etemad, Shahrokh; Ulkarim, Hasan; Castaldi, Marco J.; Pfefferle, William C.

    2001-01-01

    A catalytic firebox reactor employing an exothermic catalytic reaction channel and multiple cooling conduits for creating a partially reacted fuel/oxidant mixture. An oxidation catalyst is deposited on the walls forming the boundary between the multiple cooling conduits and the exothermic catalytic reaction channel, on the side of the walls facing the exothermic catalytic reaction channel. This configuration allows the oxidation catalyst to be backside cooled by any fluid passing through the cooling conduits. The heat of reaction is added to both the fluid in the exothermic catalytic reaction channel and the fluid passing through the cooling conduits. After discharge of the fluids from the exothermic catalytic reaction channel, the fluids mix to create a single combined flow. A further innovation in the reactor incorporates geometric changes in the exothermic catalytic reaction channel to provide streamwise variation of the velocity of the fluids in the reactor.

  2. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOEpatents

    Schreiber, Roger B.; Fero, Arnold H.; Sejvar, James

    1997-01-01

    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 to form 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 valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

  3. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOEpatents

    Schreiber, R.B.; Fero, A.H.; Sejvar, J.

    1997-12-16

    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 to form 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 valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

  4. Development and Optimization of Modular Hybrid Plasma Reactor

    SciTech Connect

    N /A

    2013-01-02

    INL developed a bench–scale, modular hybrid plasma system for gas-phase nanomaterials synthesis. The system was optimized for WO{sub 3} nanoparticle production and scale-model projection to a 300 kW pilot system. During the course of technology development, many modifications were made to the system to resolve technical issues that surfaced and also to improve performance. All project tasks were completed except two optimization subtasks. Researchers were unable to complete these two subtasks, a four-hour and an eight-hour continuous powder production run at 1 lb/hr powder-feeding rate, due to major technical issues developed with the reactor system. The 4-hour run was attempted twice, and on both occasions, the run was terminated prematurely. The termination was due to (1) heavy material condensation on the modular electrodes, which led to system operational instability, and (2) pressure buildup in the reactor due to powder clogging of the exhaust gas and product transfer line. The modular electrode for the plasma system was significantly redesigned to address the material condensation problem on the electrodes. However, the cause for product powder clogging of the exhaust gas and product transfer line led to a pressure build up in the reactor that was undetected. Fabrication of the redesigned modular electrodes and additional components was completed near the end of the project life. However, insufficient resource was available to perform tests to evaluate the performance of the new modifications. More development work would be needed to resolve these problems prior to scaling. The technology demonstrated a surprising capability of synthesizing a single phase of meta-stable {delta}- Al{sub 2}O{sub 3} from pure {alpha}-phase large Al{sub 2}O{sub 3} powder. The formation of {delta} -Al{sub 2}O{sub 3} was surprising because this phase is meta-stable and only formed between 973–1073 K, and {delta} -Al{sub 2}O{sub 3} is very difficult to synthesize as a single

  5. Tritium Formation and Mitigation in High-Temperature Reactors

    SciTech Connect

    Piyush Sabharwall; Carl Stoots

    2012-10-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

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

    SciTech Connect

    Piyush Sabharwall; Carl Stoots; Hans A. Schmutz

    2013-03-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

  7. Bioaugmentation of a sequencing batch biofilm reactor with Comamonas testosteroni and Bacillus cereus and their impact on reactor bacterial communities.

    PubMed

    Cheng, Zhongqin; Chen, Mei; Xie, Liqun; Peng, Lin; Yang, Maohua; Li, Mengying

    2015-02-01

    The immobilization of microorganisms is essential for efficient bioaugmentation systems. The performance of Bacillus cereus G5 as biofilm-forming bacteria and Comamonas testosteroni A3 a 3,5 dinitrobenzoic acid (DNB)-degrading strain] in laboratory-scale sequencing batch biofilm reactors (SBBRs) treating DNB synthetic wastewater has been examined. The microbial diversity in the reactors was also explored. The reactor R3 inoculated with B. cereus G5 and C. testosteroni A3 together not only improved the removal of contaminants, but also exhibited obvious resistance to shock loading with DNB during later operations. Pyrosequencing was used to evaluate bacterial communities in three reactors. Comamonas was predominant in the reactor R3, indicating the effect of G5 in promoting immobilization of A3 cells in biofilms. Those microbial resources, e.g.G5, which can stimulate the self-immobilization of the degrading bacteria offer a novel strategy for immobilization of degraders in bioaugmentation systems and show broader application prospects.

  8. Deuterium and tritium separation in a tokamak reactor divertor layer

    NASA Astrophysics Data System (ADS)

    Tokar', M. Z.

    1989-04-01

    It's shown that the plasma isotope composition in a tokamak reactor divertor layer changes along the magnetic field and can notable differ from the gas composition in a pumping chamber. Heavier tritium must concentrate in the hot plasma far from the divertor plate due to thermal force stipulated by mutial collisions of deuterium and tritium ions. This circumstance is favourable from the point of view of tritium cycle optimization and must facilitate solution of the problem of tritium accumulation in the reactor construction elements.

  9. Neutron and gamma characterization within the FFTF reactor cavity

    SciTech Connect

    Bunch, W.L.; Carter, L.L.; Moore, F.S.; Werner, E.J.; Wilcox, A.D.; Wood, M.R.

    1980-08-01

    Neutron and gamma ray measurements were made within the reactor cavity of the Fast Flux Test Facility (FFTF) to establish the operating characteristics of the Ex-Vessel Flux Monitoring (EVFM) system as a function of reactor power level. A significant effort was made to obtain absolute flux values in order that the measurements could be compared directly with shield design calculations. Good agreement was achieved for neutrons and for both the prompt and delayed components of the gamma ray field. 8 figures, 3 tables.

  10. Alloying of steel and graphite by hydrogen in nuclear reactor

    NASA Astrophysics Data System (ADS)

    Krasikov, E.

    2017-02-01

    In traditional power engineering hydrogen may be one of the first primary source of equipment damage. This problem has high actuality for both nuclear and thermonuclear power engineering. Study of radiation-hydrogen embrittlement of the steel raises the question concerning the unknown source of hydrogen in reactors. Later unexpectedly high hydrogen concentrations were detected in irradiated graphite. It is necessary to look for this source of hydrogen especially because hydrogen flakes were detected in reactor vessels of Belgian NPPs. As a possible initial hypothesis about the enigmatical source of hydrogen one can propose protons generation during beta-decay of free neutrons поскольку inasmuch as protons detected by researches at nuclear reactors as witness of beta-decay of free neutrons.

  11. Non-Proliferative, Thorium-Based, Core and Fuel Cycle for Pressurized Water Reactors

    SciTech Connect

    Todosow M.; Todosow M.; Raitses, G. Galperin, A.

    2009-07-12

    amount of discharged spent fuel, for a given energy production, compared with standard VVER/PWR. The total Pu production rate of RTF cycles is only 30 % of standard reactor. In addition, the isotopic compositions of the RTF's and standard reactor grade Pu are markedly different due to the very high burnup accumulated by the RTF spent fuel.

  12. Uncertainty and instream flow standards

    USGS Publications Warehouse

    Castleberry, D.; Cech, J.; Erman, D.; Hankin, D.; Healey, M.; Kondolf, M.; Mengel, M.; Mohr, M.; Moyle, P.; Nielsen, Jennifer; Speed, T.; Williams, J.

    1996-01-01

    Several years ago, Science published an important essay (Ludwig et al. 1993) on the need to confront the scientific uncertainty associated with managing natural resources. The essay did not discuss instream flow standards explicitly, but its arguments apply. At an April 1995 workshop in Davis, California, all 12 participants agreed that currently no scientifically defensible method exists for defining the instream flows needed to protect particular species of fish or aquatic ecosystems (Williams, in press). We also agreed that acknowledging this fact is an essential step in dealing rationally and effectively with the problem.Practical necessity and the protection of fishery resources require that new instream flow standards be established and that existing standards be revised. However, if standards cannot be defined scientifically, how can this be done? We join others in recommending the approach of adaptive management. Applied to instream flow standards, this approach involves at least three elements.

  13. Induced Radioactivity and Waste Classification of Reactor Zone Components of the Chernobyl Nuclear Power Plant Unit 1 After Final Shutdown

    SciTech Connect

    Bylkin, Boris K.; Davydova, Galina B.; Zverkov, Yuri A.; Krayushkin, Alexander V.; Neretin, Yuri A.; Nosovsky, Anatoly V.; Seyda, Valery A.; Short, Steven M.

    2001-10-15

    The dismantlement of the reactor core materials and surrounding structural components is a major technical concern for those planning closure and decontamination and decommissioning of the Chernobyl Nuclear Power Plant (NPP). Specific issues include when and how dismantlement should be accomplished and what the radwaste classification of the dismantled system would be at the time it is disassembled. Whereas radiation levels and residual radiological characteristics of the majority of the plant systems are directly measured using standard radiation survey and radiochemical analysis techniques, actual measurements of reactor zone materials are not practical due to high radiation levels and inaccessibility. For these reasons, neutron transport analysis was used to estimate induced radioactivity and radiation levels in the Chernobyl NPP Unit 1 reactor core materials and structures.Analysis results suggest that the optimum period of safe storage is 90 to 100 yr for the Unit 1 reactor. For all of the reactor components except the fuel channel pipes (or pressure tubes), this will provide sufficient decay time to allow unlimited worker access during dismantlement, minimize the need for expensive remote dismantlement, and allow for the dismantled reactor components to be classified as low- or medium-level radioactive waste. The fuel channel pipes will remain classified as high-activity waste requiring remote dismantlement for hundreds of years due to the high concentration of induced {sup 63}Ni in the Zircaloy pipes.

  14. THERMAL FISSION REACTOR COMPOSITIONS AND METHOD OF FABRICATING SAME

    DOEpatents

    Blainey, A.

    1959-10-01

    A body is presented for use in a thermal fission reactor comprising a sintered compressed mass of a substance of the group consisting of uranium, thorium, and oxides and carbides of uranium and thorium, enclosed in an envelope of a sintered, compacted, heat-conductive material of the group consisting of beryllium, zirconium, and oxides and carbides of beryllium and zirconium.

  15. Standards and Access.

    ERIC Educational Resources Information Center

    Fox, Tom

    1993-01-01

    Argues that easy claims about the relationship between language mastery and academic or economic access (made by both conservative commentators on education and mainstream writing teachers) are false and obscure real social and political boundaries, such as racism, sexism, elitism, and homophobia, that really do prevent access. (SR)

  16. The detection of global convection on the sun by an analysis of line shift data of the John M. Wilcox Solar Observatory at Stanford University

    NASA Technical Reports Server (NTRS)

    Yoshimura, Hirokazu

    1987-01-01

    An analysis of the absorption line shift data of the John M. Wilcox Solar Observatory at Stanford University has yielded signatures of the existence of global convection on the sun. These include persistent periodic time variations in the east-west component of the velocity fields defined by fitting a slope to the line shift data in a certain longitude window at a specified latitude and longitude by the least squares method. The amplitude of the velocity fields estimated from these variations is of the order of 100 m/s. The results of the analysis also suggest that several modes of global convection coexist in the solar convection zone. Details of the analysis are given.

  17. Electroformed Nanocrystalline Coatings An Advanced Alternative to Hard-Chrome Electroplating

    DTIC Science & Technology

    2003-04-02

    E. Lee, Babcock & Wilcox Canada Dr. Jonathan L. McCrea, Integran Technologies Dr. Uwe Erb, University of Toronto HCAT Meeting, San Diego, California...5e. TASK NUMBER 5f. WORK UNIT NUMBER 7. PERFORMING ORGANIZATION NAME(S) AND ADDRESS(ES) Babcock & Wilcox Canada,581 Coronation Boulevard...Through Environmental Research SM SERDP Program 1152 DACA72-00-C-003 (9/26/02) 14 H2 Embrittlement Retest Nano Co 2-3 wt% P Careful NDE of test specimens

  18. Advanced Test Reactor -- Testing Capabilities and Plans AND Advanced Test Reactor National Scientific User Facility -- Partnerships and Networks

    SciTech Connect

    Frances M. Marshall

    2008-07-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world’s premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner “lobes” to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. For future research, some ATR modifications and enhancements are currently planned. In 2007 the US Department of Energy designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR for material testing research by a broader user community. This paper provides more details on some of the ATR capabilities, key design features, experiments, and plans for the NSUF.

  19. Reference and Standard Atmosphere Models

    NASA Technical Reports Server (NTRS)

    Johnson, Dale L.; Roberts, Barry C.; Vaughan, William W.; Parker, Nelson C. (Technical Monitor)

    2002-01-01

    This paper describes the development of standard and reference atmosphere models along with the history of their origin and use since the mid 19th century. The first "Standard Atmospheres" were established by international agreement in the 1920's. Later some countries, notably the United States, also developed and published "Standard Atmospheres". The term "Reference Atmospheres" is used to identify atmosphere models for specific geographical locations. Range Reference Atmosphere Models developed first during the 1960's are examples of these descriptions of the atmosphere. This paper discusses the various models, scopes, applications and limitations relative to use in aerospace industry activities.

  20. Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation

    SciTech Connect

    Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri

    2005-09-27

    The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

  1. NUCLEAR REACTOR

    DOEpatents

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

    1958-12-01

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

  2. Design and analysis of megawatt-class heat-pipe reactor concepts

    SciTech Connect

    Poston, D.; Kapernick, R.

    2012-07-01

    There is growing interest in finding an alternative to diesel-powered systems at locations removed from a reliable electrical grid. One promising option is a 1- to 10-MW mobile reactor system, that could provide robust, self-contained, and long-term ({>=} 5 years) power in any environment. The reactor and required infrastructure could be transported to any location within one or a few standard transport containers. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than 'traditional' reactors that rely on pumped coolant through the core. This paper examines a heat pipe reactor that is fabricated and shipped as six identical core segments. Each core segment includes a heat-pipe-to-gas heat exchanger that is coupled to the condenser end of the heat pipes. The reference power conversion system is a CO{sub 2}-Brayton system. The segments by themselves are deeply subcritical during transport, and they would be locked into an operating configuration (with control inserted) at the final destination. Two design options are considered: a near-term option and an advanced option. The near-term option is a 5-MWt concept that uses uranium-dioxide fuel, a stainless-steel structure, and potassium as the heat-pipe working fluid. The advanced option is a 15-MWt concept that uses uranium-nitride fuel, a molybdenum/TZM structure, and sodium as the heat-pipe working fluid. The materials used in the advanced option allow for higher temperatures and power densities, and enhanced power throughput in the heat pipes. Higher powers can be obtained from both concepts by increasing the core size and the number of heat pipes. (authors)

  3. Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors.

    PubMed

    Mohammadi, A; Hassanzadeh, M; Gharib, M

    2016-02-01

    In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified.

  4. International academic program in technologies of light-water nuclear reactors. Phases of development and implementation

    NASA Astrophysics Data System (ADS)

    Geraskin, N. I.; Glebov, V. B.

    2017-01-01

    The results of implementation of European educational projects CORONA and CORONA II dedicated to preserving and further developing nuclear knowledge and competencies in the area of technologies of light-water nuclear reactors are analyzed. Present article addresses issues of design and implementation of the program for specialized training in the branch of technologies of light-water nuclear reactors. The systematic approach has been used to construct the program for students of nuclear specialties, which corresponding to IAEA standards and commonly accepted nuclear principles recognized in the European Union. Possibilities of further development of the international cooperation between countries and educational institutions are analyzed. Special attention is paid to e-learning/distance training, nuclear knowledge preservation and interaction with European Nuclear Education Network.

  5. Reduced enrichment for research and test reactors: Proceedings

    SciTech Connect

    Not Available

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

  6. The temperature effect and safety of the Topaz reactor

    SciTech Connect

    Artyukhov, G.Y.; Zelentsov, S.N.; Ionkin, V.I.; Kudryavtsev, V.P.; Makarenkov, Y.D.; Marin, S.N.; Pupko, V.Y.; Raskach, F.P. )

    1991-01-01

    This paper presents the results of a design calculation to compute the 3-dimensional geometry of the TOPAZ reactor. Emphasis was given to the calculation of the temperature effects and temperature reactivity coefficients. Performance of the control and safety systems, spectral and dynamic characteristics were also evaluated.(AIP)

  7. Blue Ribbon Commission, Yucca Mountain Closure, Court Actions - Future of Decommissioned Reactors, Operating Reactors and Nuclear Power - 13249

    SciTech Connect

    Devgun, Jas S.

    2013-07-01

    Issues related to back-end of the nuclear fuel cycle continue to be difficult for the commercial nuclear power industry and for the decision makers at the national and international level. In the US, the 1982 NWPA required DOE to develop geological repositories for SNF and HLW but in spite of extensive site characterization efforts and over ten billion dollars spent, a repository opening is nowhere in sight. There has been constant litigation against the DOE by the nuclear utilities for breach of the 'standard contract' they signed with the DOE under the NWPA. The SNF inventory continues to rise both in the US and globally and the nuclear industry has turned to dry storage facilities at reactor locations. In US, the Blue Ribbon Commission on America's Nuclear Future issued its report in January 2012 and among other items, it recommends a new, consent-based approach to siting of facilities, prompt efforts to develop one or more geologic disposal facilities, and prompt efforts to develop one or more consolidated storage facilities. In addition, the March 2011 Fukushima Daiichi accident had a severe impact on the future growth of nuclear power. The nuclear industry is focusing on mitigation strategies for beyond design basis events and in the US, the industry is in the process of implementing the recommendations from NRC's Near Term Task Force. (authors)

  8. Biological processing in oscillatory baffled reactors: operation, advantages and potential

    PubMed Central

    Abbott, M. S. R.; Harvey, A. P.; Perez, G. Valente; Theodorou, M. K.

    2013-01-01

    The development of efficient and commercially viable bioprocesses is essential for reducing the need for fossil-derived products. Increasingly, pharmaceuticals, fuel, health products and precursor compounds for plastics are being synthesized using bioprocessing routes as opposed to more traditional chemical technologies. Production vessels or reactors are required for synthesis of crude product before downstream processing for extraction and purification. Reactors are operated either in discrete batches or, preferably, continuously in order to reduce waste, cost and energy. This review describes the oscillatory baffled reactor (OBR), which, generally, has a niche application in performing ‘long’ processes in plug flow conditions, and so should be suitable for various bioprocesses. We report findings to suggest that OBRs could increase reaction rates for specific bioprocesses owing to low shear, good global mixing and enhanced mass transfer compared with conventional reactors. By maintaining geometrical and dynamic conditions, the technology has been proved to be easily scaled up and operated continuously, allowing laboratory-scale results to be easily transferred to industrial-sized processes. This is the first comprehensive review of bioprocessing using OBRs. The barriers facing industrial adoption of the technology are discussed alongside some suggested strategies to overcome these barriers. OBR technology could prove to be a major aid in the development of commercially viable and sustainable bioprocesses, essential for moving towards a greener future. PMID:24427509

  9. Biological processing in oscillatory baffled reactors: operation, advantages and potential.

    PubMed

    Abbott, M S R; Harvey, A P; Perez, G Valente; Theodorou, M K

    2013-02-06

    The development of efficient and commercially viable bioprocesses is essential for reducing the need for fossil-derived products. Increasingly, pharmaceuticals, fuel, health products and precursor compounds for plastics are being synthesized using bioprocessing routes as opposed to more traditional chemical technologies. Production vessels or reactors are required for synthesis of crude product before downstream processing for extraction and purification. Reactors are operated either in discrete batches or, preferably, continuously in order to reduce waste, cost and energy. This review describes the oscillatory baffled reactor (OBR), which, generally, has a niche application in performing 'long' processes in plug flow conditions, and so should be suitable for various bioprocesses. We report findings to suggest that OBRs could increase reaction rates for specific bioprocesses owing to low shear, good global mixing and enhanced mass transfer compared with conventional reactors. By maintaining geometrical and dynamic conditions, the technology has been proved to be easily scaled up and operated continuously, allowing laboratory-scale results to be easily transferred to industrial-sized processes. This is the first comprehensive review of bioprocessing using OBRs. The barriers facing industrial adoption of the technology are discussed alongside some suggested strategies to overcome these barriers. OBR technology could prove to be a major aid in the development of commercially viable and sustainable bioprocesses, essential for moving towards a greener future.

  10. Operation and performance of the Supercritical Fluids Reactor (SFR)

    SciTech Connect

    Hanush, R.G.; Rice, S.F.; Hunter, T.B.; Aiken, J.D.

    1995-11-01

    The Supercritical Fluids Reactor (SFR) at Sandia National Laboratories, CA has been developed to examine and solve engineering, process, and fundamental chemistry issues regarding the development of supercritical water oxidation (SCWO). This report details the experimental apparatus, procedures, analytical methods used in these experiments, and performance characteristics of the reactor. The apparatus consists of pressurization, feed, preheat, reactor, cool down, and separation subsystems with ancillary control and data acquisition hardware and software. Its operating range is from 375 - 650{degrees} at 3250 - 6300 psi with resident times from 0.09 to 250 seconds. Procedures required for experimental operations are described. They include maintenance procedures conducted between experiments, optical alignment for acquisition of spectroscopic data, setup of the experiment, reactor start up, experimental operations, and shutdown of apparatus. Analytical methods used are Total Organic Carbon analysis, Gas Chromatography, ion probes, pH probes, turbidity measurements and in situ Raman spectroscopy. Experiments conducted that verify the accuracy of measurement and sampling methods are described.

  11. Selecting and using materials for a nuclear rocket engine reactor

    NASA Astrophysics Data System (ADS)

    Lanin, Anatolii G.; Fedik, Ivan I.

    2011-03-01

    This paper provides a historical account of how the nuclear rocket engine reactor was created and discusses the problem of selecting materials for a gas environment at a temperature of up to 3100 K and energy release of 30 MW per liter.

  12. Yttrium and rare earth stabilized fast reactor metal fuel

    SciTech Connect

    Guon, J.; Grantham, L.F.; Specht, E.R.

    1992-05-12

    This patent describes an improved metal alloy reactor fuel consisting essentially of uranium, plutonium, and at least one element from the group consisting of yttrium, lanthanum, cerium, praseodymium, neodymium, promethium, samarium, europium, gadolinium, terbium, dysprosium, holmium, erbium, thulium, ytterbium and lutetium.

  13. Plasma Physics Lab and the Tokamak Fusion Test Reactor, 1989

    ScienceCinema

    None

    2016-07-12

    From the Princeton University Archives: Promotional video about the Plasma Physics Lab and the new Tokamak Fusion Test Reactor (TFTR), with footage of the interior, machines, and scientists at work. This film is discussed in the audiovisual blog of the Seeley G. Mudd Manuscript Library, which holds the archives of Princeton University.

  14. Semiconductor Chemical Reactor Engineering and Photovoltaic Unit Operations.

    ERIC Educational Resources Information Center

    Russell, T. W. F.

    1985-01-01

    Discusses the nature of semiconductor chemical reactor engineering, illustrating the application of this engineering with research in physical vapor deposition of cadmium sulfide at both the laboratory and unit operations scale and chemical vapor deposition of amorphous silicon at the laboratory scale. (JN)

  15. Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

    SciTech Connect

    Adams, S.R.

    1985-10-01

    A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences.

  16. Reference Collections and Standards.

    ERIC Educational Resources Information Center

    Winkel, Lois

    1999-01-01

    Reviews six reference materials for young people: "The New York Public Library Kid's Guide to Research"; "National Audubon Society First Field Guide. Mammals"; "Star Wars: The Visual Dictionary"; "Encarta Africana"; "World Fact Book, 1998"; and "Factastic Book of 1001 Lists". Includes ordering information.(AEF)

  17. IQ and Standard English.

    ERIC Educational Resources Information Center

    Farrell, Thomas J.

    1983-01-01

    Points out the untestability of a heredity explanation for the differences in IQs of Black children and those of White children. Suggests an environmental explanation, arguing that Black children, simply have not had the opportunity to develop their potential for abstract thinking. (HTH)

  18. Unifying process control and optimization

    SciTech Connect

    Makansi, J.

    2005-09-01

    About 40% of US generation is now subject to wholesale competition. To intelligently bid into these new markets, real-time prices must be aligned with real-time costs. It is time to integrate the many advanced applications, sensors, and analyzers used for control, automation, and optimization into a system that reflects process and financial objectives. The paper reports several demonstration projects in the USA revealing what is being done in the area of advanced process optimization (by Alliant Energy, American Electric Power, PacifiCorp, Detroit Edison and Tennessee Valley Authority). In addition to these projects US DOE's NETL has funded the plant environment and cost optimization system, PECOS which combines physical models, neural networks and fuzzy logic control to provide operators with least cost setpoints for controllable variables. At Dynegy Inc's Baldwin station in Illinois the DOE is subsidizing a project where real time, closed-loop IT systems will optimize combustion, soot-blowing and SCR performance as well as unit thermal performance and plant economic performance. Commercial products such as Babcock and Wilcox's Flame Doctor, continuous emissions monitoring systems and various real-time predictive monitoring systems are also available. 4 figs.

  19. Biases and Standard Errors of Standardized Regression Coefficients

    ERIC Educational Resources Information Center

    Yuan, Ke-Hai; Chan, Wai

    2011-01-01

    The paper obtains consistent standard errors (SE) and biases of order O(1/n) for the sample standardized regression coefficients with both random and given predictors. Analytical results indicate that the formulas for SEs given in popular text books are consistent only when the population value of the regression coefficient is zero. The sample…

  20. Synfuels from fusion: producing hydrogen with the Tandem Mirror Reactor and thermochemical cycles

    SciTech Connect

    Werner, R.W.; Ribe, F.L.

    1981-01-21

    This volume contains the following sections: (1) the Tandem Mirror fusion driver, (2) the Cauldron blanket module, (3) the flowing microsphere, (4) coupling the reactor to the process, (5) the thermochemical cycles, and (6) chemical reactors and process units. (MOW)