Structural framework and sand genesis of Wilcox group, Travis Ward field, Jim Hogg County, Texas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rolf, E.G.

1987-09-01

Since its discovery in 1983, there have been eight deep Wilcox and eight Queen City wells drilled in the Travis Ward field area. Of the eight Wilcox wells, four are producing gas from deep sands; three, that are capable of production, have been junked and abandoned, and one produces from the Hinnant sand at the top of the Wilcox. Only five of the eight Queen city wells have been completed; three are considered commercial. Wilcox gas reserve estimates range from 80 to 300 bcf. To date, Wilcox and Queen City production is related to normal faulting associated with a deepmore » salt and/or shale ridge within the Rio Grande interior salt basin. Growth of the ridge has resulted in the Wilcox being as much as 2000 ft structurally higher than the areas immediately north and south of Travis Ward field. Knowledge of the ancestral development of ridge closure prior to faulting may be critical to successful completions at Travis Ward field. Ridge-associated sea floor topography, shelf currents, sediment source proximity, and rate of sedimentation have combined for local development of high quality clean reservoir sands.« less

CHARACTERISTIC QUALITIES OF SOME ATOMIC POWER STATIONS (in Hungarian)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ligeti, G.

1962-04-01

Mostly as the result of economic factors, the current rate of construction of public atomic power stations has slowed down. The use of atomic energy is considered economical only in a few special cases, such as ship propulsion or supplying power to remote regions. For this reason, many reactors were designed especially for the construction of such midget'' power stations, operating at power levels ranging from 10 to 70 Mw. Technical details are given of such already-built or proposed systems, including the following: pressurized- water reactors such as the Babcock and Wilcox 60-Mw reactor, using 2.4% U/sup 235/ fuel; themore » Humphrey-Glasow Company's 20 Mw reactor; the gascooled system of the de Havilland Company; the organicmoderated reactor of the English Electric Company; the organic-moderated system of the Hawker-Siddeley Nuclear Power Company; the boiling-water reactor of the Mitchell Engineering Company and the steam-cooled, heavy-water reactor of the Rolls-Royce & Vickers Company. (TTT)« less

Paleocene coal deposits of the Wilcox group, central Texas

USGS Publications Warehouse

Hook, Robert W.; Warwick, Peter D.; SanFilipo, John R.; Schultz, Adam C.; Nichols, Douglas J.; Swanson, Sharon M.; Warwick, Peter D.; Karlsen, Alexander K.; Merrill, Matthew D.; Valentine, Brett J.

2011-01-01

Coal deposits in the Wilcox Group of central Texas have been regarded as the richest coal resources in the Gulf Coastal Plain. Although minable coal beds appear to be less numerous and generally higher in sulfur content (1 percent average, as-received basis; table 1) than Wilcox coal deposits in the Northeast Texas and Louisiana Sabine assessment areas (0.5 and 0.6 percent sulfur, respectively; table 1), net coal thickness in coal zones in central Texas is up to 32 ft thick and more persistent along strike (up to 15 mi) at or near the surface than coals of any other Gulf Coast assessment area. The rank of the coal beds in central Texas is generally lignite (table 1), but some coal ranks as great as subbituminous C have been reported (Mukhopadhyay, 1989). The outcrop of the Wilcox Group in central Texas strikes northeast, extends for approximately 140 mi between the Trinity and Colorado Rivers, and covers parts of Bastrop, Falls, Freestone, Lee, Leon, Limestone, Milam, Navarro, Robertson, and Williamson Counties (Figure 1). Three formations, in ascending order, the Hooper, Simsboro, and Calvert Bluff, are recognized in central Texas (Figure 2). The Wilcox Group is underlain conformably by the Midway Group, a mudstone-dominated marine sequence, and is overlain and scoured locally by the Carrizo Sand, a fluvial unit at the base of the Claiborne Group.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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. Topicmore » 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.« less

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.

Hydrogeologic characteristics and water levels of Wilcox aquifer in southwestern and northeastern Arkansas

USGS Publications Warehouse

Pugh, Aaron L.; Schrader, Tony P.

2009-01-01

The Wilcox Group of Eocene and Paleocene age is located throughout most of southern and eastern Arkansas. The Wilcox Group in southern Arkansas is undifferentiated, while in northeastern Arkansas, the Wilcox Group is subdivided into three units: Flour Island, Fort Pillow Sand, and Old Breastworks Formation. The Wilcox Group crops out in southwestern Arkansas in discontinuous, 1 to 3 mi wide bands. In northeastern Arkansas, the Wilcox Group crops out along a narrow, discontinuous, band along the western edge of Crowleys Ridge. The Wilcox aquifer provides sources of groundwater in southwestern and northeastern Arkansas. In 2005, reported withdrawals from the Wilcox aquifer in Arkansas totaled 27.0 million gallons per day, most of which came from the northeastern area. Major withdrawals from the aquifer were for public supplies with lesser but locally important withdrawals for commercial, domestic, and industrial uses. 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 levels associated with the Wilcox aquifer in southwestern and northeastern Arkansas. During February 2009, 58 water-level measurements were made in wells completed in the Wilcox aquifer. The results from this study and previous studies are presented as potentiometric-surface maps, water-level difference maps, and long-term hydrographs. The direction of groundwater flow in the southwestern area is affected by two potentiometric-surface mounds, one in the north and the other in the southwest, and a cone of depression in the center. The direction of water flowing off of the northern mound of water is generally to the south and east with some to the north. The direction of water flowing off of the southwestern mound is generally to the south and east. The direction of water flowing into the cone of depression is generally from the north, west, and south. The direction of groundwater flow

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

Lithofacies control of lignite distribution and ground-water quality, Wilcox group (Eocene), east-central Texas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ayers, W.B. Jr.; Lewis, A.H.

1984-04-01

Deep lignite resources (200-2000 ft; 61-610 m) were evaluated regionally using 1470 geophysical well logs to interpret lithofacies, lignite occurrence, and resistivity (water quality). The regional distribution of lithofacies indicates that in the region, the Wilcox Group is a fluvial-deltaic system. The primary fluvial system entered the Wilcox coastal plain west of Waco, Texas, trended southeast, and supplied a 75-mi (120-km) wide fluvial-deltaic system comparable in size to the Mississippi system. Lignites are most abundant in the Calvert Bluff Formation (upper Wilcox). Lower Calvert Bluff lignites are thickest and most extensive southwest of the Navasota River, whereas those of themore » upper Calvert Bluff are thickest northeast of the Brazos River. In the shallow subsurface, Calvert Bluff lignites are found in dip-elongate low-sand areas (flood plains) between channel-sand belts. Basinward, laterally continuous lignites coincide with high net sand areas comprise of distributary channel sands indicative of a delta-plain setting. The wilcox Group is a major aquifer. Maps of resistivity values show that Wilcox channel sands are conduits for ground-water flow. High values of formation resistivity (low total dissolved solids) exist in recharge areas at outcrop and around salt domes. Elongate trends of high resistivity values extend tens of miles basinward and coincide with axes of major sands. Resistivity values decrease basinward and the 20 ohm-m contour delineates the downdip limit of fresh water. Lithofacies and lignite occurrence maps are guides to exploration for deep lignite. Resistivity maps can be used to explore for ground-water resources.« less

FIELD EVALUATION OF LOW-EMISSION COAL BURNER TECHNOLOGY ON UTILITY BOILERS VOLUME III. FIELD EVALUATIONS

EPA Science Inventory

The report gives results of field tests conducted to determine the emission characteristics of a Babcock and Wilcox Circular burner and Dual Register burner (DRB). The field tests were performed at two utility boilers, generally comparable in design and size except for the burner...

77. Detail view looking east showing Dovel horizontal gas washer ...

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

77. Detail view looking east showing Dovel horizontal gas washer in foreground, Rust Co. boilers and blowing engine house at left, and Babcock & Wilcox type boilers at right. - Sloss-Sheffield Steel & Iron, First Avenue North Viaduct at Thirty-second Street, Birmingham, Jefferson County, AL

Additions to the flora of the Wilcox group

USGS Publications Warehouse

Berry, Edward Wilber

1923-01-01

A rather full account of the extensive flora contained in the lower Eocene strata of the Mississippi embayment which are referred to the Wilcox group was published in 1916. At that time it was not possible to obtain sections of the numerous specimens of petrified wood that had been collected from these beds. These woods have since been sectioned and studied, and it seems eminently desirable to place the results of this study on record, for although much of the material had suffered greatly from decay before silicification, some of it is fairly well preserved and shows, among other results, that conifers were individually much more plentiful during Wilcox time than would be inferred from the almost total absence of their foliage in the very large collections of remains of this class that have been studied.

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).

76. General view looking east showing Rust Co. boiler stacks ...

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

76. General view looking east showing Rust Co. boiler stacks at left, Babcock & Wilcox type boiler stacks at right, Dovel horizontal gas washer in foreground, and No. 1 Furnace in distance. - Sloss-Sheffield Steel & Iron, First Avenue North Viaduct at Thirty-second Street, Birmingham, Jefferson County, AL

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.

Paleocene coal deposits of the Wilcox Group, Northeast Texas

USGS Publications Warehouse

Hook, Robert W.; Warwick, Peter D.; SanFilipo, John R.; Nichols, Douglas J.; Swanson, Sharon M.; Warwick, Peter D.; Karlsen, Alexander K.; Merrill, Matthew D.; Valentine, Brett J.

2011-01-01

The surface exposure of the Paleocene Wilcox Group in northeast Texas varies in width from 9 to 27 mi along an arcuate outcrop that extends southwest approximately 156 mi from the Texas-Arkansas State line to 32° latitude. Parts of Bowie, Camp, Cass, Franklin, Henderson, Hopkins, Morris, Navarro, Rains, Titus, Van Zandt, and Wood Counties are included in this outcrop belt (Figure 1). This area forms the northwestern flank of the East Texas Basin (Figure 2), the axis of which separates northeast Texas from the Sabine uplift structural area. The Wilcox Group dips south and southeast at 2° or less toward the axis of the East Texas Basin, with the exception of local salt-dome structures and a transcurrent structural high that extends from Monticello to Martin Lake (Figure 1).

50. View looking east down rail rightofway traveled by ladle ...

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

50. View looking east down rail right-of-way traveled by ladle car taking molten iron from furnaces to pig machine; gas main at left, Babcock & Wilcox boilers at right. - Sloss-Sheffield Steel & Iron, First Avenue North Viaduct at Thirty-second Street, Birmingham, Jefferson County, AL

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

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 anymore » 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.« less

Dr Amos G Babcock - fact or fiction?

PubMed

Smith, Douglas

2014-11-01

The War of 1812-14 between the United States of America and Great Britain gave rise to several journals relating the sufferings of prisoners of war confined in prison ships and gaols in England. One of these is A Journal of a Young Man from Massachusetts, said to have been written by Dr Amos G Babcock, an American ship's surgeon, and first published in 1816. This article sets out arguments for and against the truth of this assertion. © The Author(s) 2013 Reprints and permissions: sagepub.co.uk/journalsPermissions.nav.

Thermal maturation history of the Wilcox group (Paleocene-Eocene), Texas: Results of regional-scale multi-1D modeling

USGS Publications Warehouse

Rowan, E.L.; Warwick, Peter D.; Pitman, Janet K.; Kennan, Lorcan; Pindell, James; Rosen, Norman C.

2007-01-01

The thermal maturation history of the Paleocene-Eocene Wilcox Group has been reconstructed based on burial history models of 53 wells in the Texas coastal plain. This modeling study has been conducted in conjunction with a geologically based assessment of the oil and gas resources in Cenozoic strata of the Gulf of Mexico coastal plain and state waters. In the onshore Texas coastal plain, coals and organic-rich shales, predominantly of terrestrial origin, within the Wilcox Group are the primary source of oil (Wenger et al., 1994) as well as a source of gas. The Wilcox, however, is modeled as a single unit, without subdivision into source rock and non-source rock intervals.Generation of oil from Type III kerogen within the Wilcox Group is modeled using hydrous pyrolysis reaction kinetic parameters (Lewan, M.D., written communication, 2006). Gas generation from Type III kerogen is represented using calculated Ro values. The models are calibrated with bottom hole temperature (BHT), and vitrinite reflectance (Ro %) data for the Wilcox Group. Ro data from near-coastal sites have been selected to minimize the possible effects of uplift and erosion and then composited to give a regional Rodepth trend.Model calculations for the study area, the onshore Texas coastal plain, indicate that downdip portions of the basal Wilcox had reached sufficient thermal maturity to generate hydrocarbons by early Eocene (~50 Ma). This relatively early maturation is explained by rapid sediment accumulation in the early Tertiary combined with the reaction kinetic parameters used in the models. Thermal maturation increases through time with increasing burial depth and temperature, gradually moving the maturation front updip. At present day, hydrocarbon generation is complete in the downdip Wilcox within the study area but is currently ongoing in the updip portions of the formation.

Good Engineering + Poor Communication = Three Mile Island.

ERIC Educational Resources Information Center

Mathes, J. C.

The accident at the Three Mile Island nuclear power plant resulted from a communication failure. Following an incident at an Ohio plant a year and a half earlier, B. M. Dunn, manager of Emergency Core Cooling Systems Analysis at Babcock and Wilcox (engineers), wrote a memorandum making specific recommendations on written instructions for nuclear…

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.

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.

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.

Qualification of submerged-arc narrow strip cladding process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ayres, P.S.; Gottschling, J.D.; Jeffers, G.K.

1976-03-01

Babcock and Wilcox has developed an unique narrow strip cladding process for use on both plate and forging material for nuclear components. The qualification testing of this low-heat input process for cladding nuclear components is described, including those of SA508 Class 2 material. The theory that explains the acceptable results of these tests is also given.

Geologic assessment of undiscovered conventional oil and gas resources in the Lower Paleogene Midway and Wilcox Groups, and the Carrizo Sand of the Claiborne Group, of the Northern Gulf coast region

USGS Publications Warehouse

Warwick, Peter D.

2017-09-27

The U.S. Geological Survey (USGS) recently conducted an assessment of the undiscovered, technically recoverable oil and gas potential of Tertiary strata underlying the onshore areas and State waters of the northern Gulf of Mexico coastal region. The assessment was based on a number of geologic elements including an evaluation of hydrocarbon source rocks, suitable reservoir rocks, and hydrocarbon traps in an Upper Jurassic-Cretaceous-Tertiary Composite Total Petroleum System defined for the region by the USGS. Five conventional assessment units (AUs) were defined for the Midway (Paleocene) and Wilcox (Paleocene-Eocene) Groups, and the Carrizo Sand of the Claiborne Group (Eocene) interval including: (1) the Wilcox Stable Shelf Oil and Gas AU; (2) the Wilcox Expanded Fault Zone Gas and Oil AU; (3) the Wilcox-Lobo Slide Block Gas AU; (4) the Wilcox Slope and Basin Floor Gas AU; and (5) the Wilcox Mississippi Embayment AU (not quantitatively assessed).The USGS assessment of undiscovered oil and gas resources for the Midway-Wilcox-Carrizo interval resulted in estimated mean values of 110 million barrels of oil (MMBO), 36.9 trillion cubic feet of gas (TCFG), and 639 million barrels of natural gas liquids (MMBNGL) in the four assessed units. The undiscovered oil resources are almost evenly divided between fluvial-deltaic sandstone reservoirs within the Wilcox Stable Shelf (54 MMBO) AU and deltaic sandstone reservoirs of the Wilcox Expanded Fault Zone (52 MMBO) AU. Greater than 70 percent of the undiscovered gas and 66 percent of the natural gas liquids (NGL) are estimated to be in deep (13,000 to 30,000 feet), untested distal deltaic and slope sandstone reservoirs within the Wilcox Slope and Basin Floor Gas AU.

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.

Composite Golf Clubs

NASA Technical Reports Server (NTRS)

1976-01-01

Babcock & Wilcox Co. under a partnership with Marshall Space Flight Center, produced composite materials, originally from the shuttle program, for improving golf clubs. Company used Marshall Space Flight Center's data summary file summarizing typical processing techniques and mechanical and physical properties of graphite and boron- reinforced composite materials. Reinforced composites provide combination of shaft rigidity and flexibility that provide maximum distance.

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.

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.

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…

CONVECTIVE BABCOCK-LEIGHTON DYNAMO MODELS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miesch, Mark S.; Brown, Benjamin P., E-mail: miesch@ucar.edu

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 unclearmore » 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.« less

A Babcock-Leighton solar dynamo model with multi-cellular meridional circulation in advection- and diffusion-dominated regimes

NASA Astrophysics Data System (ADS)

Belucz, B.; Dikpati, M.; Forgacs-Dajka, E.

2014-12-01

Babcock-Leighton type solar dynamo models with single cell meridional circulation are successful in reproducing many solarcycle features, and recently such a model was applied for solarcycle 24 amplitude prediction. It seems that cycle 24 amplitudeforecast may not be validated. One of the reasons is the assumption of a single cell meridional circulation. Recent observations andtheoretical models of meridional circulation do not indicate a single-celledflow pattern. So it is nessecary to examine the role of complexmulti-cellular circulation patterns in a Babcock-Leighton solar dynamo model in the advection and diffusion dominated regimes.By simulating a Babcock-Leighton solar dynamo model with multi-cellularflow, we show that the presence of a weak, second, high-latitudereverse cell speeds up the cycle and slighty enhances the poleward branch in the butterfly diagram, whereas the presence of a second cellin depth reverses the tilt of the butterfly wing and leads to ananti-solar type feature. If, instead, the butterfly diagram isconstructed from the middle of the convection zone in that case, a solar-like pattern can be retrieved. All the above cases behavequalitatively similar in advection and diffusion-dominated regimes.However, our dynamo with a meridional circulation containing fourcells in latitude behaves distinctly different in the two regimes, producing a solar-like butterfly diagram with fast cycles indiffusion-dominated regime, and a complex branches in the butterflydiagram in the advection-dominated regime. Another interestingfinding from our studies is that a four-celled flow pattern containing two in radius and two in latitude always producesquadrupolar parity as the relaxed solution.

Environmental geology of the Wilcox Group Lignite Belt, east Texas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Henry, C.D.; Basciano, J.M.

This report provides a data base for decisions about lignite mining and reclamation in the Wilcox Group of East Texas. A set of environmental geologic maps, which accompanies this report, depicts the character of the land that will be affected by mining. The environmental geologic maps of the East Texas lignite belt provide an accurate inventory of land resources. The maps identify areas where mining is most likely to occur, areas of critical natural resources that could be affected by mining, such as aquifer recharge areas, and areas of natural hazards, such as floodplains. Principal areas of both active andmore » planned surface mining are also located. The seven environmental geologic maps cover the outcrop area of the Wilcox Group, the major lignite host, and adjacent geologic units from Bastrop County to Texarkana. This report begins with a discussion of various physical aspects of the lignite belt, including geology, hydrology, soils, climate, and land use, to aid in understanding the maps. The criteria and methodology used to delineate the environmental geologic units are discussed. Varied applications of the environmental geologic maps are considered. 23 references, 9 figures, 3 tables.« less

Great Lakes Steel -- PCI facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eichinger, F.T.; Dake, S.H.; Wagner, E.D.

1997-12-31

This paper discusses the planning, design, and start-up of the 90 tph PCI facility for National Steel`s Great Lakes Steel Division in River Rouge, MI. This project is owned and operated by Edison Energy Services, and was implemented on a fast-track basis by Raytheon Engineers and Constructors, Babcock Material Handling, and Babcock and Wilcox. This paper presents important process issues, basic design criteria, an the challenges of engineering and building a state-of-the-art PCI facility in two existing plants. Pulverized coal is prepared at the River Rouge Power Plant of Detroit Edison, is pneumatically conveyed 6,000 feet to a storage silomore » at Great Lakes Steel, and is injected into three blast furnaces.« less

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.

TRAC-PF1 code verification with data from the OTIS test facility. [Once-Through Intergral System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Childerson, M.T.; Fujita, R.K.

1985-01-01

A computer code (TRAC-PF1/MOD1) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the One-Through Integral System (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and loop saturation, intermittent reactor coolant system circulation, boiler-condenser mode, and the initial stages of refill. The TRAC code wasmore » successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool and auxiliary-feedwater initiated boiler-condenser mode heat transfer.« less

Paleocene Wilcox cross-shelf channel-belt history and shelf-margin growth: Key to Gulf of Mexico sediment delivery

NASA Astrophysics Data System (ADS)

Zhang, Jinyu; Steel, Ronald; Ambrose, William

2017-12-01

Shelf margins prograde and aggrade by the incremental addition of deltaic sediments supplied from river channel belts and by stored shoreline sediment. This paper documents the shelf-edge trajectory and coeval channel belts for a segment of Paleocene Lower Wilcox Group in the northern Gulf of Mexico based on 400 wireline logs and 300 m of whole cores. By quantitatively analyzing these data and comparing them with global databases, we demonstrate how varying sediment supply impacted the Wilcox shelf-margin growth and deep-water sediment dispersal under greenhouse eustatic conditions. The coastal plain to marine topset and uppermost continental slope succession of the Lower Wilcox shelf-margin sediment prism is divided into eighteen high-frequency ( 300 ky duration) stratigraphic sequences, and further grouped into 5 sequence sets (labeled as A-E from bottom to top). Sequence Set A is dominantly muddy slope deposits. The shelf edge of Sequence Sets B and C prograded rapidly (> 10 km/Ma) and aggraded modestly (< 80 m/Ma). The coeval channel belts are relatively large (individually averaging 11-13 m thick) and amalgamated. The water discharge of Sequence Sets B and C rivers, estimated by channel-belt thickness, bedform type, and grain size, is 7000-29,000 m3/s, considered as large rivers when compared with modern river databases. In contrast, slow progradation (< 10 km/Ma) and rapid aggradation (> 80 m/Ma) characterizes Sequence Sets D and E, which is associated with smaller (9-10 m thick on average) and isolated channel belts. This stratigraphic trend is likely due to an upward decreasing sediment supply indicated by the shelf-edge progradation rate and channel size, as well as an upward increasing shelf accommodation indicated by the shelf-edge aggradation rate. The rapid shelf-edge progradation and large rivers in Sequence Sets B and C confirm earlier suggestions that it was the early phase of Lower Wilcox dispersal that brought the largest deep-water sediment

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

A reexamination of the North American Crepis agamic complex and comparison with the findings of Babcock and Stebbins' classic biosystematic monograph.

PubMed

Sears, Christopher J; Whitton, Jeannette

2016-07-01

Babcock and Stebbins coined the term agamic complex in their 1938 monograph of the North American Crepis agamic complex. Despite the historical role that this complex holds in the evolutionary literature, it has not been reexamined in over 75 years. We present a thorough reevaluation of the complex to test hypotheses proposed by Babcock and Stebbins about its origins and spread, the relationships of diploids, and the nature and origins of polyploids. We used flow cytometry to infer ploidy of roughly 600 samples spanning the morphological and taxonomic diversity of the complex and a phylogenetic analysis of plastid DNA variation to infer maternal relationships among diploids and to infer maternal origins of polyploids. We identified populations of all seven recognized diploids plus one new lineage. Phylogenetic analysis of plastid DNA variation in diploids revealed a well-resolved, but moderately supported phylogeny, with evidence for monophyly of the North America Crepis agamic complex and no evidence of widespread homoploid hybridization. Polyploids showed evidence of multiple origins and a pattern of frequent local co-occurrence consistent with repeated colonization of suitable sites. Our findings agree broadly with the distribution and variation of ploidy within and among species described by Babcock and Stebbins. One key difference is finding support for monophyly of North American species, and refuting their hypothesis of polyphyly. Our results provide an explicit phylogenetic framework for further study of this classic agamic complex. © 2016 Botanical Society of America.

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.

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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'smore » 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)« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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 validationmore » 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.« less

Stochastic Fluctuations in a Babcock-Leighton Model of the Solar Cycle

NASA Astrophysics Data System (ADS)

Charbonneau, Paul; Dikpati, Mausumi

2000-11-01

We investigate the effect of stochastic fluctuations on a flux transport model of the solar cycle based on the Babcock-Leighton mechanism. Specifically, we make use of our recent flux transport model (Dikpati & Charbonneau) to investigate the consequences of introducing large-amplitude stochastic fluctuations in either or both the meridional flow and poloidal source term in the model. Solar cycle-like oscillatory behavior persists even for fluctuation amplitudes as high as 300%, thus demonstrating the inherent robustness of this class of solar cycle models. We also find that high-amplitude fluctuations lead to a spread of cycle amplitude and duration showing a statistically significant anticorrelation, comparable to that observed in sunspot data. This is a feature of the solar cycle that is notoriously difficult to reproduce with dynamo models based on mean field electrodynamics and relying only on nonlinearities associated with the back-reaction of the Lorentz force to produce amplitude modulation. Another noteworthy aspect of our flux transport model is the fact that meridional circulation in the convective envelope acts as a ``clock'' regulating the tempo of the solar cycle; shorter-than-average cycles are typically soon followed by longer-than-average cycles. In other words, the oscillation exhibits good phase locking, a property that also characterizes the solar activity cycle. This shows up quite clearly in our model, but we argue that it is in fact a generic property of flux transport models based on the Babcock-Leighton mechanism, and relies on meridional circulation as the primary magnetic field transport agent.

78 FR 73898 - Operator Licensing Examination Standards for Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-09

... Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft NUREG; request for comment. SUMMARY: The U.S..., Revision 10, ``Operator Licensing Examination Standards for Power Reactors.'' DATES: Submit comments [email protected] . Both of the Office of New Reactors; or Timothy Kolb, Office of Nuclear Reactor Regulation, U...

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

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.

Standard interface files and procedures for reactor physics codes, version III

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carmichael, B.M.

Standards and procedures for promoting the exchange of reactor physics codes are updated to Version-III status. Standards covering program structure, interface files, file handling subroutines, and card input format are included. The implementation status of the standards in codes and the extension of the standards to new code areas are summarized. (15 references) (auth)

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Belles, Randy; Poore, III, Willis P.; Brown, Nicholas R.

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-basedmore » description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.« less

Chemical evolution of groundwater in the Wilcox aquifer of the northern Gulf Coastal Plain, USA

NASA Astrophysics Data System (ADS)

Haile, Estifanos; Fryar, Alan E.

2017-12-01

The Wilcox aquifer is a major groundwater resource in the northern Gulf Coastal Plain (lower Mississippi Valley) of the USA, yet the processes controlling water chemistry in this clastic aquifer have received relatively little attention. The current study combines analyses of solutes and stable isotopes in groundwater, petrography of core samples, and geochemical modeling to identify plausible reactions along a regional flow path ˜300 km long. The hydrochemical facies evolves from Ca-HCO3 upgradient to Na-HCO3 downgradient, with a sequential zonation of terminal electron-accepting processes from Fe(III) reduction through SO4 2- reduction to methanogenesis. In particular, decreasing SO4 2- and increasing δ34S of SO4 2- along the flow path, as well as observations of authigenic pyrite in core samples, provide evidence of SO4 2- reduction. Values of δ13C in groundwater suggest that dissolved inorganic carbon is contributed both by oxidation of sedimentary organic matter and calcite dissolution. Inverse modeling identified multiple plausible sets of reactions between sampled wells, which typically involved cation exchange, pyrite precipitation, CH2O oxidation, and dissolution of amorphous Fe(OH)3, calcite, or siderite. These reactions are consistent with processes identified in previous studies of Atlantic Coastal Plain aquifers. Contrasts in groundwater chemistry between the Wilcox and the underlying McNairy and overlying Claiborne aquifers indicate that confining units are relatively effective in limiting cross-formational flow, but localized cross-formational mixing could occur via fault zones. Consequently, increased pumping in the vicinity of fault zones could facilitate upward movement of saline water into the Wilcox.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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. 32 refs.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burge, S.W.

Erosion has been identified as one of the significant design issues in fluid beds. A cooperative R&D venture of industry, research, and government organizations was recently formed to meet the industry need for a better understanding of erosion in fluid beds. Research focussed on bed hydrodynamics, which are considered to be the primary erosion mechanism. As part of this work, ANL developed an analytical model (FLUFIX) for bed hydrodynamics. Partial validation was performed using data from experiments sponsored by the research consortium. Development of a three-dimensional fluid bed hydrodynamic model was part of Asea-Babcock`s in-kind contribution to the R&D venture.more » This model, FORCE2, was developed by Babcock & Wilcox`s Research and Development Division existing B&W program and on the gas-solids modeling and was based on an existing B&W program and on the gas-solids modeling technology developed by ANL and others. FORCE2 contains many of the features needed to model plant size beds and, therefore can be used along with the erosion technology to assess metal wastage in industrial equipment. As part of the development efforts, FORCE2 was partially validated using ANL`s two-dimensional model, FLUFIX, and experimental data. Time constraints as well as the lack of good hydrodynamic data, particularly at the plant scale, prohibited a complete validation of FORCE2. This report describes this initial validation of FORCE2.« less

Standardized reactors for the study of medical biofilms: a review of the principles and latest modifications.

PubMed

Gomes, Inês B; Meireles, Ana; Gonçalves, Ana L; Goeres, Darla M; Sjollema, Jelmer; Simões, Lúcia C; Simões, Manuel

2018-08-01

Biofilms can cause severe problems to human health due to the high tolerance to antimicrobials; consequently, biofilm science and technology constitutes an important research field. Growing a relevant biofilm in the laboratory provides insights into the basic understanding of the biofilm life cycle including responses to antibiotic therapies. Therefore, the selection of an appropriate biofilm reactor is a critical decision, necessary to obtain reproducible and reliable in vitro results. A reactor should be chosen based upon the study goals and a balance between the pros and cons associated with its use and operational conditions that are as similar as possible to the clinical setting. However, standardization in biofilm studies is rare. This review will focus on the four reactors (Calgary biofilm device, Center for Disease Control biofilm reactor, drip flow biofilm reactor, and rotating disk reactor) approved by a standard setting organization (ASTM International) for biofilm experiments and how researchers have modified these standardized reactors and associated protocols to improve the study and understanding of medical biofilms.

Acceptance Test Data for BWXT Coated Particle Batch 93164A Defective IPyC Fraction and Pyrocarbon Anisotropy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Helmreich, Grant W.; Hunn, John D.; Skitt, Darren J.

2017-02-01

Coated particle fuel batch J52O-16-93164 was produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), or may be used as demonstration production-scale coated particle fuel for other experiments. The tristructural-isotropic (TRISO) coatings were deposited in a 150-mm-diameter production-scale fluidizedbed chemical vapor deposition (CVD) furnace onto 425-μm-nominal-diameter spherical kernels from BWXT lot J52L-16-69316. Each kernel contained a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO) and was coated with four consecutive CVD layers:more » a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. The TRISO-coated particle batch was sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batch was designated by appending the letter A to the end of the batch number (i.e., 93164A).« less

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…

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…

A Babcock-Leighton Solar Dynamo Model with Multi-cellular Meridional Circulation in Advection- and Diffusion-dominated Regimes

NASA Astrophysics Data System (ADS)

Belucz, Bernadett; Dikpati, Mausumi; Forgács-Dajka, Emese

2015-06-01

Babcock-Leighton type-solar dynamo models with single-celled meridional circulation are successful in reproducing many solar cycle features. Recent observations and theoretical models of meridional circulation do not indicate a single-celled flow pattern. We examine the role of complex multi-cellular circulation patterns in a Babcock-Leighton solar dynamo in advection- and diffusion-dominated regimes. We show from simulations that the presence of a weak, second, high-latitude reverse cell speeds up the cycle and slightly enhances the poleward branch in the butterfly diagram, whereas the presence of a second cell in depth reverses the tilt of the butterfly wing to an antisolar type. A butterfly diagram constructed from the middle of convection zone yields a solar-like pattern, but this may be difficult to realize in the Sun because of magnetic buoyancy effects. Each of the above cases behaves similarly in higher and lower magnetic diffusivity regimes. However, our dynamo with a meridional circulation containing four cells in latitude behaves distinctly differently in the two regimes, producing solar-like butterfly diagrams with fast cycles in the higher diffusivity regime, and complex branches in butterfly diagrams in the lower diffusivity regime. We also find that dynamo solutions for a four-celled pattern, two in radius and two in latitude, prefer to quickly relax to quadrupolar parity if the bottom flow speed is strong enough, of similar order of magnitude as the surface flow speed.

Air Sampling Logbook of Region 4 Yellow Bluff Air Study Wilcox County, Alabama SESD Project Identification Number:11-0068

EPA Pesticide Factsheets

Contains the Air Sampling Logbook between 1-24-2011 thru 1-28-2011 from the Region 4 Yellow Bluff Air Study Wilcox County, Alabama SESD Project Identification Number:11-0068 November 2010-December 2010

Coal resources for part of the Wilcox group (Paleocene-Eocene), northeast Texas

USGS Publications Warehouse

Warwick, Peter D.; Aubourg, Claire E.; Podwysocki, Steven M.; Schultz, Adam C.; Warwick, Peter D.; Karlsen, Alexander K.; Merrill, Matthew D.; Valentine, Brett J.

2011-01-01

The Wilcox Group of northeast Texas contains shallow (less than 500 ft) coal deposits that are mined for use in mine-mouth electric power generating plants. The coal deposits, which are lignite in apparent rank (Pierce et al., 2011), are separated from similar shallow coal deposits in the Sabine uplift area by the East Texas Basin (Figure 1). The coal zones and associated strata in the northeast assessment area generally dip to the south and southeast at 28 or less toward the axis of the East Texas Basin. The northeast Texas resource assessment area includes parts of nine counties (Figure 2).

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.; Warwick, Peter D.; Karlsen, Alexander K.; Merrill, Matthew D.; Valentine, Brett J.

2011-01-01

The Wilcox Group of central Texas contains shallow (less than 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 (Pierce et al., 2011) and are similar in rank and composition to shallow coal deposits in the northeast and south Texas areas (Figure 1). The coal zones and associated strata in the central Texas assessment 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 (Figure 2). The assessment area was selected to encompass current mining areas and areas with available subsurface stratigraphic data. The assessment area is roughly 160 miles long and 5 to 25 miles wide and generally follows the outcrop of the Paleocene to Eocene Wilcox Group in central Texas (Figures 1, 2). Approximately 1800 subsurface stratigraphic records from rotary and core drill holes were used to assess the resources of the central Texas assessment area. Of the 1800 drill holes, only 167 are public data points and are primarily located in the areas that have been permitted for surface mining (Figure 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.

Grindability measurements on low-rank fuels. [Prediction of large pulverizer performance from small scale test equipment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peipho, R.R.; Dougan, D.R.

1981-01-01

Experience has shown that the grinding characteristics of low rank coals are best determined by testing them in a pulverizer. Test results from a small MPS-32 Babcock and Wilcox pulverizer to predict large, full-scale pulverizer performance are presented. The MPS-32 apparatus, test procedure and evaluation of test results is described. The test data show that the Hardgrove apparatus and the ASTM test method must be used with great caution when considering low-rank fuels. The MPS-32 meets the needs for real-machine simulation but with some disadvantages. A smaller pulverizer is desirable. 1 ref.

Technology could deliver 90% Hg reduction from coal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maize, K.

2009-07-15

Reducing mercury emissions at coal-fired power plants by 90% has been considered the holy grail of mercury control. A new technology promises to get used there, but at a price. This is a mixture of chemical approaches, including activated carbon injection into the gases coming off the combustor along with injection of trona or calcium carbonate to reduce sulfur trioxide in the exhaust gases. The trick according to Babcock and Wilcox's manager Sam Kumar, to 'capture the mercury as a particulate on the carbon and then capture the particulate' in an electrostatic precipitator or a fabric filter baghouse. 2 figs.

Estimation of weekly 99Mo production by AHR 200 kW

NASA Astrophysics Data System (ADS)

Siregar, I. H.; Suharyana; Khakim, A.; Siregar, D.; Frida, A. R.

2016-11-01

The estimation of weekly 99Mo production by AHR 200 kW fueled with Low Enriched Uranium Uranyl Nitrate solution has been simulated by using MCNPX computer code. We have employed the AHR design of Babcock & Wilcox Medical Isotope Production System with 9Be Reflector and Stainless steel vessel. We found that when the concentration of uranium in the fresh fuel was 108 gr U/L of UO2(NO3)2 fuel solution, the multiplication factor was 1.0517. The 99Mo concentration reached saturated at tenth day operation. The AHR can produce approximately 1.96×103 6-day-Ci weekly.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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)more » Condition Assessment of Unit RL-10 Boiler.« less

Acceptance Test Data for Candidate AGR-5/6/7 TRISO Particle Batches BWXT Coater Batches 93165 93172 Defective IPyC Fraction and Pyrocarbon Anisotropy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Helmreich, Grant W.; Hunn, John D.; Skitt, Darren J.

2017-03-01

Coated particle fuel batches J52O-16-93165, 93166, 93168, 93169, 93170, and 93172 were produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR). Some of these batches may alternately be used as demonstration coated particle fuel for other experiments. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture of 15.5%-enriched uranium carbide andmore » uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μmnominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93165A).« less

A Conversation with Allen J. Wilcox

PubMed Central

Jukic, Anne Marie Z.

2017-01-01

Allen James Wilcox was born on 30 September 1946 in Columbus, Ohio. He studied medicine at the University of Michigan, graduating in 1973, and after a rotating internship completed a master’s degree in maternal and child health (1976) and a PhD in epidemiology (1979) at the University of North Carolina in Chapel Hill. After graduation, he went to work at the National Institute of Environmental Health Sciences (NIEHS, one of the US National Institutes of Health) in Durham, North Carolina, where he has spent his career. He developed a research program in reproductive and perinatal epidemiology, a relatively unexplored area at the time. His studies include the Early Pregnancy Study, which documented the extent of subclinical pregnancy loss in humans and established the fertile days of a woman’s menstrual cycle. He served as Chief of the Epidemiology Branch from 1991 to 2001, and as Editor-in-Chief of the journal EPIDEMIOLOGY from 2001 to 2014. His textbook, Fertility and Pregnancy – An Epidemiologic Perspective, was published by Oxford University Press in 2010. He was elected to the American Epidemiological Society in 1989, and served as its president in 2003. He also served as president of the Society of Pediatric and Perinatal Epidemiological Research (1996) and the president of the Society of Epidemiological Research (1998). He holds adjunct teaching appointments at the University of North Carolina, Harvard University, and the University of Bergen (Norway), which awarded him an honorary doctoral degree in 2008. PMID:27482869

A THREE-DIMENSIONAL BABCOCK-LEIGHTON SOLAR DYNAMO MODEL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miesch, Mark S.; Dikpati, Mausumi, E-mail: miesch@ucar.edu

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)more » 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.« less

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

... Nuclear Operations Group, Inc.; Establishment of Atomic Safety and Licensing Board Pursuant to delegation... 2.106, 2.300, 2.313(a), and 2.318, notice is hereby given that an Atomic Safety and Licensing Board..., 2010. The Board is comprised of the following administrative judges: Alan S. Rosenthal, Chair, Atomic...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 followingmore » sections: (1) Condition Assessment of Unit RL-10 Turbine-Generator; (2) Condition Assessment of Unit RL-10 and RL-11 Boiler Feed Pumps.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 followingmore » sections: (1) Executive Summary; (2) Hydrogen Damage Assessment; (3) RL-17 Gas Conversion Proposal; (4) RL-10 and RL-11 Gas Conversion Proposals.« less

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.

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

2017-06-01

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.

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.

Howard Wilcox Haggard and the Institutionalization of Modern Alcohol Studies.

PubMed

Allred, Nicholas; Bejarano, William; Ward, Judit

2017-03-01

This biographical sketch and accompanying bibliography provide a new look at Howard Wilcox Haggard, M.D., Ph.D., whose career highlights the consolidation of alcohol studies as a field in twentieth-century America. The article relies in large part on the works of Haggard assembled for the bibliography project, supplemented by published and unpublished documents and records from collections at Rutgers University. Haggard began his career in respiratory physiology, influenced by his work on chemical weapons for the Army during the First World War. As his reputation grew, he moved into anesthesiology and supplemented his research with textbooks and popular science bestsellers. Haggard moved into the burgeoning field of alcohol studies after the repeal of National Prohibition and, in 1940, became the inaugural editor and president of the corporation of the Quarterly Journal of Studies on Alcohol, now the Journal of Studies on Alcohol and Drugs. Under the aegis of the Yale Laboratory for Applied Physiology, he also assembled and oversaw what would become the Center of Alcohol Studies. Haggard died in 1959, his legacy established as a central figure in the 20th-century transformation of alcohol studies in the United States. A prolific researcher with a talent for tapping into the public zeitgeist, Haggard helped provide the institutional infrastructure, academic credibility, and broad audience that made the renaissance of alcohol studies in post-Prohibition America possible.

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)

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.

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

USGS Publications Warehouse

Hackley, Paul C.; Warwick, Peter D.; Breland, F. Clayton

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 consistently are undersaturated with respect to CH4 adsorption isotherms; deeper (> 600 m) coal samples containing less moisture range from under- to oversaturated with respect to their CH4adsorption capacity.

Solar Cycle Variability Induced by Tilt Angle Scatter in a Babcock-Leighton Solar Dynamo Model

NASA Astrophysics Data System (ADS)

Karak, Bidya Binay; Miesch, Mark

2017-09-01

We present results from a three-dimensional Babcock-Leighton (BL) dynamo model that is sustained by the emergence and dispersal of bipolar magnetic regions (BMRs). On average, each BMR has a systematic tilt given by Joy’s law. Randomness and nonlinearity in the BMR emergence of our model produce variable magnetic cycles. However, when we allow for a random scatter in the tilt angle to mimic the observed departures from Joy’s law, we find more variability in the magnetic cycles. We find that the observed standard deviation in Joy’s law of {σ }δ =15^\\circ produces a variability comparable to the observed solar cycle variability of ˜32%, as quantified by the sunspot number maxima between 1755 and 2008. We also find that tilt angle scatter can promote grand minima and grand maxima. The time spent in grand minima for {σ }δ =15^\\circ is somewhat less than that inferred for the Sun from cosmogenic isotopes (about 9% compared to 17%). However, when we double the tilt scatter to {σ }δ =30^\\circ , the simulation statistics are comparable to the Sun (˜18% of the time in grand minima and ˜10% in grand maxima). Though the BL mechanism is the only source of poloidal field, we find that our simulations always maintain magnetic cycles even at large fluctuations in the tilt angle. We also demonstrate that tilt quenching is a viable and efficient mechanism for dynamo saturation; a suppression of the tilt by only 1°-2° is sufficient to limit the dynamo growth. Thus, any potential observational signatures of tilt quenching in the Sun may be subtle.

Taxonomic revision of Pseudorhopalia Wilcox & Papavero, 1971 (Insecta, Diptera, Mydidae, Rhopaliinae), with description of a new species from the Brazilian Amazon .

PubMed

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

2014-11-17

The previously monotypic genus Pseudorhopalia Wilcox & Papavero, 1971 is here revised, with description of a new species, Pseudorhopalia manauara, sp. n. The type-species, P. mirandai (d'Andretta & Carrera, 1951), is only found in northeastern Brazil, in the semiarid Caatinga biome. On the other hand, the new species is found in the humid Brazilian Amazonia, within dense ombrophilous forest and campina (scleromorphic scrub vegetation associated with white sand).

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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), Pacificmore » 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).« less

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

... under 10 CFR 2.315(c), must be filed in accordance with the NRC E-Filing rule (72 FR 49139, August 28, 2007). The E-Filing process requires participants to submit and serve all adjudicatory documents over... below. To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing...

A General Small-Scale Reactor To Enable Standardization and Acceleration of Photocatalytic Reactions.

PubMed

Le, Chi Chip; Wismer, Michael K; Shi, Zhi-Cai; Zhang, Rui; Conway, Donald V; Li, Guoqing; Vachal, Petr; Davies, Ian W; MacMillan, David W C

2017-06-28

Photocatalysis for organic synthesis has experienced an exponential growth in the past 10 years. However, the variety of experimental procedures that have been reported to perform photon-based catalyst excitation has hampered the establishment of general protocols to convert visible light into chemical energy. To address this issue, we have designed an integrated photoreactor for enhanced photon capture and catalyst excitation. Moreover, the evaluation of this new reactor in eight photocatalytic transformations that are widely employed in medicinal chemistry settings has confirmed significant performance advantages of this optimized design while enabling a standardized protocol.

Reservoir characterization and modeling of deltaic facies, Lower Wilcox, Concordia Parish, Louisiana

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schenewerk, P.; Goddard, D.; Echols, J.

Production decline in several fields in Concordia Parish, Louisiana, has sparked interest in the economic feasibility of producing the remaining bypassed oil in the lower Wilcox. One of these fields, the Bee Brake field, located in townships 4N, 6E and 4N, 7E, has been one of the more prolific oil-producing areas in east central Louisiana. The producing interval in the field, the Minter, typically consists of an upper Bee Brake sand and a lower Angelina sand. Cumulative production from the Angelina has been 2.1 mm STB of oil. A detailed study of a conventional core in the center of themore » field presented a 15-ft-thick Minter interval bounded above and below by sealing shales and lignites of lower delta plain marsh facies. The lower oil producing 3-ft thick Angelina consists of fine to medium sandstone of overbank bay fill facies. The upper 4-ft thick Bee Brake is a very fine silty sandstone with characteristics of a crevasse splay deposit. Special core analysis data (capillary pressure, relative permeability, and waterflood recovery) were obtained and have been used to develop a simulation model of the two reservoirs in the Minter. This model incorporates the geologic and engineering complexities noted during the first comprehensive evaluation of the field area. The model results will be used by the operators in the field to plan the optimal development for enhanced recovery. In addition, the production potential of the Bee Brake sand has been defined.« less

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

... NUCLEAR REGULATORY COMMISSION [NRC-2010-0228] Office of New Reactors; Proposed Revision to Standard Review Plan Section 13.6.1, Revision 1 on Physical Security--Combined License and Operating...), Section 13.6.1 on ``Physical Security--Combined License and Operating Reactors,'' (Agencywide Documents...

Reservoir characterization and preliminary modeling of deltaic facies, lower Wilcox, Concordia Parish, Louisiana

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schenewerk, P.; Goddard, D.; Echols, J.

The decline in production in several fields in Concordia Parish, Louisiana, has created interest in the economic feasibility of producing the remaining bypassed oil in the lower Wilcox Group. One of these fields, Bee Brake, has been one of the more prolific oil-producing fields in east-central Louisiana. The producing interval, the Minter sandstones, at a depth of about 6,775 ft typically consists of an upper Bee Brake sandstone and a lower Angelina sandstone. A detailed study of a conventional core in the center of the field reveals a 15-ft-thick Minter interval bounded above and below by sealing shales and lignitesmore » of lower delta plain marsh facies. The upper 4-ft-thick Bee Brake is a very fine silty sandstone with characteristics of a small overbank or crevasse splay deposit. The lower 3-ft-thick oil-producing Angelina sandstone consists of very fine and fine sandstone of probable overbank or crevasse facies. Cumulative production from the Angelina is about 1.8 million stock-tank barrels of oil. Special core analysis data (capillary pressure, relative permeability, and waterflood recovery) have been used to develop a simulation model of the two reservoirs in the Minter. This model incorporates the geologic and engineering complexities noted during evaluation of the field area. Operators can use the model results in this field to design an optimal development plan for enhanced recovery.« less

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

... Standard Review Plan, Section 1.0 on Introduction and Interfaces AGENCY: Nuclear Regulatory Commission (NRC... Revision 2 to Standard Review Plan (SRP), Section 1.0, ``Introduction and Interfaces'' (Agencywide Documents Access and Management System (ADAMS) Accession No. ML110110573). The Office of New Reactors (NRO...

Application of CFCC technology to hot gas filtration applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Richlen, S.

1995-06-01

Discussion will feature high temperature filter development under the DOE`s Office of Industrial Technologies Continuous Fiber Ceramic Composite (CFCC) Program. Within the CFCC Program there are four industry projects and a national laboratory technology support project. Atlantic Research, Babcock & Wilcox, DuPont Lanxide Composites, and Textron are developing processing methods to produce CFCC Components with various types of matrices and composites, along with the manufacturing methods to produce industrial components, including high temperature gas filters. The Oak Ridge National Laboratory is leading a National Laboratory/University effort to increase knowledge of such generic and supportive technology areas as environmental degradation, measurementmore » of mechanical properties, long-term performance, thermal shock and thermal cycling, creep and fatigue, and non-destructive characterization. Tasks include composite design, materials characterization, test methods, and performance-related phenomena, that will support the high temperature filter activities of industry and government.« less

Dual fuel gradients in uranium silicide plates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pace, B.W.

1997-08-01

Babcock & Wilcox has been able to achieve dual gradient plates with good repeatability in small lots of U{sub 3}Si{sub 2} plates. Improvements in homogeneity and other processing parameters and techniques have allowed the development of contoured fuel within the cladding. The most difficult obstacles to overcome have been the ability to evaluate the bidirectional fuel loadings in comparison to the perfect loading model and the different methods of instilling the gradients in the early compact stage. The overriding conclusion is that to control the contour of the fuel, a known relationship between the compact, the frames and final coremore » gradient must exist. Therefore, further development in the creation and control of dual gradients in fuel plates will involve arriving at a plausible gradient requirement and building the correct model between the compact configuration and the final contoured loading requirements.« less

Acceptance Test Data for BWXT Coated Particle Batches 93172B and 93173B—Defective IPyC and Pyrocarbon Anisotropy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunn, John D.; Helmreich, Grant W.; Dyer, John A.

Coated particle batches J52O-16-93172B and J52O-16-93173B were produced by Babcock and Wilcox Technologies (BWXT) as part of the production campaign for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), but were not used in the final fuel composite. However, these batches may be used as demonstration production-scale coated particle fuel for other experiments. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture ofmore » 15.5%-enriched uranium carbide and uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93172A). Secondary upgrading by sieving was performed on the A-designated batches to remove particles with missing or very-thin buffer layers that were identified during previous analysis of the individual batches for defective IPyC, as reported in the acceptance test data report for the AGR-5/6/7 production batches [Hunn et al. 2017b]. The additionally-upgraded batches were designated by appending the letter B to the end of the batch number (e.g., 93172B).« less

Plant maintenance and plant life extension issue, 2009

DOE Office of Scientific and Technical Information (OSTI.GOV)

Agnihotri, Newal

The focus of the March-April issue is on plant maintenance and plant life extension. Major articles include the following: Application of modeling and simulation to nuclear power plants, by Berry Gibson, IBM, and Rolf Gibbels, Dassault Systems; Steam generators with tight manufacturing procedures, by Ei Kadokami, Mitsubishi Heavy Industries; SG design based on operational experience and R and D, by Jun Tang, Babcock and Wilcox Canada; Confident to deliver reliable performance, by Bruce Bevilacqua, Westinghouse Nuclear; An evolutionary plant design, by Martin Parece, AREVA NP, Inc.; and, Designed for optimum production, by Danny Roderick, GE Hitachi Nuclear Energy. Industry Innovationmore » articles include: Controlling alloy 600 degradation, by John Wilson, Exelon Nuclear Corporation; Condensate polishing innovation, by Lewis Crone, Dominion Millstone Power Station; Reducing deposits in steam generators, by the Electric Power Research Institute; and, Minimizing Radiological effluent releases, by the Electric Power Research Institute. The plant profile article is titled 2008 - a year of 'firsts' for AmerenUE's Callaway plant, by Rick Eastman, AmerenUE.« less

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.

TRAC analyses for CCTF and SCTF tests and UPTF design/operation. [Cylindrical Core Test Facility; Slab Core Test Facility; Upper Plenum Test Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spore, J.W.; Cappiello, M.W.; Dotson, P.J.

The analytical support in 1985 for Cylindrical Core Test Facility (CCTF), Slab Core Test Facility (SCTF), and Upper Plenum Test Facility (UPTF) tests involves the posttest analysis of 16 tests that have already been run in the CCTF and the SCTF and the pretest analysis of 3 tests to be performed in the UPTF. Posttest analysis is used to provide insight into the detailed thermal-hydraulic phenomena occurring during the refill and reflood tests performed in CCTF and SCTF. Pretest analysis is used to ensure that the test facility is operated in a manner consistent with the expected behavior of anmore » operating full-scale plant during an accident. To obtain expected behavior of a plant during an accident, two plant loss-of-coolant-accident (LOCA) calculations were performed: a 200% cold-leg-break LOCA calculation for a 2772 MW(t) Babcock and Wilcox plant and a 200% cold-leg-break LOCA calculation for a 3315 MW(t) Westinghouse plant. Detailed results are presented for several CCTF UPI tests and the Westinghouse plant analysis.« less

75 FR 29588 - Office of New Reactors: Proposed NUREG-0800; Standard Review Plan Section 13.6.6, Draft Revision...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-26

... NUCLEAR REGULATORY COMMISSION [NRC-2010-0184] Office of New Reactors: Proposed NUREG-0800; Standard Review Plan Section 13.6.6, Draft Revision 0 on Cyber Security Plan AGENCY: Nuclear Regulatory... Plants,'' on a proposed Standard Review Plan (SRP) Section 13.6.6 on ``Cyber Security Plan'' (Agencywide...

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

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0055] Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of Final Design Approval The U.S. Nuclear Regulatory Commission has issued a final design approval (FDA) to GE Hitachi Nuclear Energy (GEH) for the economic...

X-ray Analysis of Defects and Anomalies in AGR-5/6/7 TRISO Particles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Helmreich, Grant W.; Hunn, John D.; Skitt, Darren J.

2017-06-01

Coated particle fuel batches J52O-16-93164, 93165, 93166, 93168, 93169, 93170, and 93172 were produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), or may be used for other tests. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture of 15.4%-enriched uranium carbide and uranium oxide (UCO), with the exception of Batchmore » 93164, which used similar kernels from BWXT lot J52L-16-69316. The TRISO-coatings consisted of a ~50% dense carbon buffer layer with 100-μmnominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. Each coated particle batch was sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batch was designated by appending the letter A to the end of the batch number (e.g., 93164A). Secondary upgrading by sieving was performed on the upgraded batches to remove specific anomalies identified during analysis for Defective IPyC, and the upgraded batches were designated by appending the letter B to the end of the batch number (e.g., 93165B). Following this secondary upgrading, coated particle composite J52R-16-98005 was produced by BWXT as fuel for the AGR Program’s AGR-5/6/7 irradiation test in the INL ATR. This composite was comprised of coated particle fuel batches J52O-16-93165B, 93168B, 93169B, and 93170B.« less

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

... the proposed Revision 4 to Standard Review Plan Section 8.1 on ``Electric Power-- Introduction.'' The... NUCLEAR REGULATORY COMMISSION [NRC-2011-0119] Office of New Reactors; Proposed Revision 4 to Standard Review Plan Section 8.1 on Electric Power--Introduction, Correction AGENCY: Nuclear Regulatory...

Acceptance Test Data for the AGR-5/6/7 Irradiation Test Fuel Composite Defective IPyC Fraction and Pyrocarbon Anisotropy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Helmreich, Grant W.; Hunn, John D.; Skitt, Darren J.

Coated particle composite J52R-16-98005 was produced by Babcock and Wilcox Technologies (BWXT) as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR). This composite was comprised of four coated particle fuel batches J52O-16-93165B (26%), 93168B (26%), 93169B (24%), and 93170B (24%), chosen based on the Quality Control (QC) data acquired for each individual candidate AGR-5/6/7 batch. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT Lot J52R-16-69317more » containing a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93165A). Secondary upgrading by sieving was performed on the A-designated batches to remove particles with missing or very-thin buffer layers that were identified during previous analysis of the individual batches for defective IPyC, as reported in the acceptance test data report for the AGR-5/6/7 production batches [Hunn et al. 2017]. The additionally-upgraded batches were designated by appending the letter B to the end of the batch number (e.g., 93165B).« less

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

Interpretational confounding is due to misspecification, not to type of indicator: comment on Howell, Breivik, and Wilcox (2007).

PubMed

Bollen, Kenneth A

2007-06-01

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 (formative) indicators rests on several claims: (a) A latent variable exists apart from the model when there are effect (reflective) indicators but not when there are causal (formative) indicators, (b) causal (formative) indicators need not have the same consequences, (c) causal (formative) indicators are inherently subject to interpretational confounding, and (d) a researcher cannot detect interpretational confounding when using causal (formative) indicators. This article shows that each claim is false. Rather, interpretational confounding is more a problem of structural misspecification of a model combined with an underidentified model that leaves these misspecifications undetected. Interpretational confounding does not occur if the model is correctly specified whether a researcher has causal (formative) or effect (reflective) indicators. It is the validity of a model not the type of indicator that determines the potential for interpretational confounding. Copyright 2007 APA, all rights reserved.

Tank Vapor Sampling and Analysis Data Package for Tank 241-Z-361 Sampled 09/22/1999 and 09/271999 During Sludge Core Removal

DOE Office of Scientific and Technical Information (OSTI.GOV)

VISWANATH, R.S.

This data package presents sampling data and analytical results from the September 22 and 27, 1999, headspace vapor sampling of Hanford Site Tank 241-2-361 during sludge core removal. The Lockheed Martin Hanford Corporation (LMHC) sampling team collected the samples and Waste Management Laboratory (WML) analyzed the samples in accordance with the requirements specified in the 241-2361 Sludge Characterization Sampling and Analysis Plan, (SAP), HNF-4371, Rev. 1, (Babcock and Wilcox Hanford Corporation, 1999). Six SUMMA{trademark} canister samples were collected on each day (1 ambient field blank and 5 tank vapor samples collected when each core segment was removed). The samples weremore » radiologically released on September 28 and October 4, 1999, and received at the laboratory on September 29 and October 6, 1999. Target analytes were not detected at concentrations greater than their notification limits as specified in the SAP. Analytical results for the target analytes and tentatively identified compounds (TICs) are presented in Section 2.2.2 starting on page 2B-7. Three compounds identified for analysis in the SAP were analyzed as TICs. The discussion of this modification is presented in Section 2.2.1.2.« less

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.

Advancements in NDE for utilities and the petrochemical industry through electromagnetic acoustic transducers (EMATs)

NASA Astrophysics Data System (ADS)

Robertson, M. O.; Stevens, Donald M.; Schlader, Daniel M.; Tilley, Richard M.

1998-03-01

The ultrasonic testing (UT) method continues to broaden in its effectiveness and capabilities for nondestructive evaluation (NDE). Much of this expansion can be attributed to advancements in specific techniques of the method. The utilization of electromagnetic acoustic transducers (EMATs) in dedicated ultrasonic systems has provided McDermott Technology, Inc. (MTI), formerly Babcock & Wilcox, with significant advantages over conventional ultrasonics. In recent years, through significant R&D, MTI has been instrumental in bringing about considerable advancements in the maturing EMAT technology. Progress in electronic design, magnet configurations, and sensor concepts has greatly improved system capabilities while reducing cost and equipment size. These improvements, coupled with the inherent advantages of utilizing the non-contact EMAT technique, have combined to make this technology a viable option for many commercial system inspection applications. MTI has recently completed the development and commercialization of an EMAT-based UT scanner for boiler tube thickness measurements. MTI is currently developing an automated EMAT scanner, based on phased array technology, for complete volumetric inspection of circumferential girth welds associated with pipelines (intended primarily for offshore applications). Additional benefits of phased array technology for providing materials characterization are currently being researched.

A Coupled 2 × 2D Babcock-Leighton Solar Dynamo Model. II. Reference Dynamo Solutions

NASA Astrophysics Data System (ADS)

Lemerle, Alexandre; Charbonneau, Paul

2017-01-01

In this paper we complete the presentation of a new hybrid 2 × 2D flux transport dynamo (FTD) model of the solar cycle based on the Babcock-Leighton mechanism of poloidal magnetic field regeneration via the surface decay of bipolar magnetic regions (BMRs). This hybrid model is constructed by allowing the surface flux transport (SFT) simulation described in Lemerle et al. to provide the poloidal source term to an axisymmetric FTD simulation defined in a meridional plane, which in turn generates the BMRs required by the SFT. A key aspect of this coupling is the definition of an emergence function describing the probability of BMR emergence as a function of the spatial distribution of the internal axisymmetric magnetic field. We use a genetic algorithm to calibrate this function, together with other model parameters, against observed cycle 21 emergence data. We present a reference dynamo solution reproducing many solar cycle characteristics, including good hemispheric coupling, phase relationship between the surface dipole and the BMR-generating internal field, and correlation between dipole strength at cycle maximum and peak amplitude of the next cycle. The saturation of the cycle amplitude takes place through the quenching of the BMR tilt as a function of the internal field. The observed statistical scatter about the mean BMR tilt, built into the model, acts as a source of stochasticity which dominates amplitude fluctuations. The model thus can produce Dalton-like epochs of strongly suppressed cycle amplitude lasting a few cycles and can even shut off entirely following an unfavorable sequence of emergence events.

The Simulator Development for RDE Reactor

NASA Astrophysics Data System (ADS)

Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.

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.

Reynolds number effects in combustion noise

NASA Technical Reports Server (NTRS)

Seshan, P. K.

1981-01-01

Acoustic emission spectra have been obtained for non-premixed turbulent combustion from two small diameter laboratory gas burners, two commercial gas burners and a large gas burner in the firebox of a Babcock-Wilcox Boiler (50,000 lb steam/hr). The changes in burner size and firing rate represent changes in Reynolds number and changes in air/fuel ratio represent departure from stoichiometric proportions. The combustion efficiency was measured independently through gas analysis. The acoustic spectra obtained from the various burners exhibit a persistent shape over the Reynolds number range of 8200-82,000. The spectra were analyzed for identification of a predictable frequency domain that is most responsive to, and readily correlated with, combustion efficiency. A simple parameter (consisting of the ratio of the average acoustic power output in the most responsive frequency bandwidth to the acoustic power level of the loudest frequency) is proposed whose value increases significantly and unmistakably as combustion efficiency approaches 100%. The dependence of the most responsive frequency domain on the various Reynolds numbers associated with turbulent jets is discussed.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marrocco, M.

The Ohio Power Company`s Tidd Pressurized Fluidized Bed Combined Cycle (PFBC) program continues to be the only operating PFBC demonstration program in the nation. The 70 MWe Tidd Demonstration Plant is a Round 1 Clean Coal Technology Project constructed to demonstrate the viability of PFBC combined cycle technology. The plant is now in Rs fourth year of operation. The technology has clearly demonstrated Rs ability to achieve sulfur capture of greater than 95%. The calcium to sulfur molar ratios have been demonstrated to exceed original projections. Unit availability has steadily increased and has been demonstrated to be competitive with othermore » technologies. The operating experience of the first forty-four months of testing has moved the PFBC process from a {open_quotes}promising technology{close_quotes} to available, proven option for efficient, environmentally acceptable base load generation. Funding for the $210 million program is provided by Ohio Power Company, The U.S. Department of Energy, The Ohio Coal Development Office, and the PFBC process vendors - Asea Brown Boveri Carbon (ABBC) and Babcock and Wilcox (B&W).« less

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

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

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Additional standards and provisions affecting class 103... Regulatory Approvals § 50.43 Additional standards and provisions affecting class 103 licenses and... propose nuclear reactor designs which differ significantly from light-water reactor designs that were...

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

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 1 2014-01-01 2014-01-01 false Additional standards and provisions affecting class 103... Regulatory Approvals § 50.43 Additional standards and provisions affecting class 103 licenses and... propose nuclear reactor designs which differ significantly from light-water reactor designs that were...

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

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Additional standards and provisions affecting class 103... Regulatory Approvals § 50.43 Additional standards and provisions affecting class 103 licenses and... propose nuclear reactor designs which differ significantly from light-water reactor designs that were...

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

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Additional standards and provisions affecting class 103... Regulatory Approvals § 50.43 Additional standards and provisions affecting class 103 licenses and... propose nuclear reactor designs which differ significantly from light-water reactor designs that were...

40 CFR 63.1406 - Reactor batch process vent provisions.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 40 Protection of Environment 11 2011-07-01 2011-07-01 false Reactor batch process vent provisions... § 63.1406 Reactor batch process vent provisions. (a) Emission standards. Owners or operators of reactor... reactor batch process vent located at a new affected source shall control organic HAP emissions by...

40 CFR 63.1406 - Reactor batch process vent provisions.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 11 2010-07-01 2010-07-01 true Reactor batch process vent provisions... § 63.1406 Reactor batch process vent provisions. (a) Emission standards. Owners or operators of reactor... reactor batch process vent located at a new affected source shall control organic HAP emissions by...

THE EXPERIENCE IN THE UNITED STATES WITH REACTOR OPERATION AND REACTOR SAFEGUARDS

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCullough, C.R.

1958-10-31

Reactors are operating or planned at locations in the United States in cities, near cities, and at remote locations. There is a general pattern that the higher power reactors are not in, but fairly uear cities, and the testing reactors for more hazardous experiments are at remote locations. A great deal has been done on the theoretical and experimental study of importunt features of reactor design. The metal-water reaction is still a theoretical possibility but tests of fuel element burnout under conditions approaching reactor operation gave no reaction. It appears that nucleate boiling does not necessarily result in steam blanketingmore » and fuel melting. Much attention is being given to the calculation of core kinetics but it is being found that temperature, power, and void coefficients cannot be calculated with accuracy and experiments are required. Some surprises are found giving positive localized void coefficients. Possible oscillatory behavior of reactors is being given careful study. No dangerous oscillations have been found in operating reactors but osciliations hare appeared in experimeats. The design of control and safety systems varies wvith different constructors. The relation of control to the kinetic behavior of the reactor is being studied. The importance of sensing element locations in order to know actual local reactor power level is being recognized. The time constants of instrumentation as related to reactor kinetics are being studied. Pressure vessels for reactors are being designed and manufactured. Many of these are beyond any previous experience. The stress problem is being given careful study. The effect of radiation is being studied experimentally. The stress problems of piping and pressure vessels is a difficult design problem being met successfully in reactor plants. The proper organization and procedure for operation of reactors is being evolved for resourch, testing, and power reactors. The importance of written standards and

40 CFR 60.702 - Standards.

Code of Federal Regulations, 2014 CFR

2014-07-01

... PERFORMANCE FOR NEW STATIONARY SOURCES Standards of Performance for Volatile Organic Compound Emissions From Synthetic Organic Chemical Manufacturing Industry (SOCMI) Reactor Processes § 60.702 Standards. Each owner...

40 CFR 60.702 - Standards.

Code of Federal Regulations, 2013 CFR

2013-07-01

... PERFORMANCE FOR NEW STATIONARY SOURCES Standards of Performance for Volatile Organic Compound Emissions From Synthetic Organic Chemical Manufacturing Industry (SOCMI) Reactor Processes § 60.702 Standards. Each owner...

78 FR 76054 - Establishment of Class E Airspace; Chatom, AL

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-16

... Instrument Approach Procedures at Roy Wilcox Airport. This action enhances the safety and management of... new RNAV (GPS) Standard Instrument Approach Procedures developed at the airport. The geographic...

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stimpson, Shane; Collins, Benjamin; Kochunas, Brendan

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

Generation III reactors safety requirements and the design solutions

NASA Astrophysics Data System (ADS)

Felten, P.

2009-03-01

Nuclear energy's public acceptance, and hence its development, depends on its safety. As a reactor designer, we will first briefly remind the basic safety principles of nuclear reactors' design. We will then show how the industry, and in particular Areva with its EPR, made design evolution in the wake of the Three Miles Island accident in 1979. In particular, for this new generation of reactors, severe accidents are taken into account beyond the standard design basis accidents. Today, Areva's EPR meets all so-called "generation III" safety requirements and was licensed by several nuclear safety authorities in the world. Many innovative solutions are integrated in the EPR, some of which will be introduced here.

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.

40 CFR 63.1407 - Non-reactor batch process vent provisions.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 11 2010-07-01 2010-07-01 true Non-reactor batch process vent... § 63.1407 Non-reactor batch process vent provisions. (a) Emission standards. (1) Owners or operators of non-reactor batch process vents located at new or existing affected sources with 0.25 tons per year (0...

40 CFR 63.1407 - Non-reactor batch process vent provisions.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 40 Protection of Environment 11 2011-07-01 2011-07-01 false Non-reactor batch process vent... § 63.1407 Non-reactor batch process vent provisions. (a) Emission standards. (1) Owners or operators of non-reactor batch process vents located at new or existing affected sources with 0.25 tons per year (0...

Coal resources for the Chemard Lake (Naborton No. 2) coal zone of the lower Wilcox group (Paleocene), northwestern Louisiana

USGS Publications Warehouse

Warwick, Peter D.; Podwysocki, Steven M.; Schultz, Adam C.; Warwick, Peter D.; Karlsen, Alexander K.; Merrill, Matthew D.; Valentine, Brett J.

2011-01-01

The lower part of the Wilcox Group of northwest Louisiana contains shallow (less than 500 ft) coal deposits that are mined for use in mine-mouth electric power-generating plants. The coal deposits, which are lignite A in apparent rank (Pierce et al., 2011), occur on the eastern part of the Sabine uplift (Figure 1). The coal zones and associated strata in the assessment area generally dip away from the axis of the Red River-Bull Bayou dome that is located in the north-central part of the Louisiana Sabine assessment area (Figure 1). This assessment area includes parts of four parishes: De Soto, Red River, Natchitoches, and Sabine (Figure 2). The assessment area was selected because of its proximity to current mining areas and the availability of stratigraphic data in the area. The assessment area is roughly 60 miles long and 15 miles wide and generally extends across the central-eastern part of the Sabine uplift in northwest Louisiana (Figure 2). More than 950 stratigraphic records from rotary and core drill holes were used to assess the coal resources of the Louisiana Sabine area. Of these, 210 are public data points and are located in or near the areas that have been permitted or proposed for surface mining (Figure 2; Appendix 1). Most of the stratigraphic data used for this assessment were provided to the U.S. Geological Survey (USGS) on a confidential basis by various coal companies for use in regional studies.

Design and analysis of a nuclear reactor core for innovative small light water reactors

NASA Astrophysics Data System (ADS)

Soldatov, Alexey I.

In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment (8.0%) and specific MASLWR operational conditions (such as lower operational pressure and temperature, and increased leakage due to small core). Neutron physics calculations are performed with the commercial nuclear industry tools CASMO-4 and SIMULATE-3, developed by Studsvik Scandpower Inc. The first set of results are generated from infinite lattice level calculations with CASMO-4, and focus on evaluation of the principal differences between standard PWR fuel and MASLWR fuel. Chapter 4-1 covers aspects of fuel isotopic composition changes with burnup, evaluation of kinetic parameters and reactivity coefficients. Chapter 4-2 discusses gadolinium self-shielding and shadowing effects, and subsequent impacts on power generation peaking and Reactor Control System shadowing. The second aspect of the research is dedicated to core design issues, such as reflector design (chapter 4-3), burnable absorber distribution and programmed fuel burnup and fuel use strategy (chapter 4-4). This section also includes discussion of the parameters important for safety and evaluation of Reactor Control System options for the proposed core design. An evaluation of the sensitivity of the proposed design to uncertainty in calculated parameters is presented in chapter 4-5. The results presented in this dissertation cover a new area of reactor design and operational parameters, and may be applicable to other small and large pressurized water reactor designs.

On Heat Loading, Novel Divertors, and Fusion Reactors

NASA Astrophysics Data System (ADS)

Kotschenreuther, Mike

2006-10-01

A new magnetic divertor geometry has been proposed to solve reactor heat exhaust problems, which are far more severe for a reactor than for ITER. Using reactor-compatible coils to generate an extra X-point downstream from the main X-point, the new X-divertor (XD) is shown to greatly expand magnetic flux at the divertor plates. As a result, the heat is distributed over a larger area and the line length is greatly increased. The heat-flux limitations of a standard divertor (SD) force a high core radiation fraction (fRad) in most reactor designs that necessarily have a several times higher ratio of heating power to radius (P/R) than ITER. It is argued that such high values of fRad will probably have serious deleterious consequences on the core confinement and stability of a burning plasma. Operation with internal transport barriers (ITBs) does not appear to overcome this problem. By reducing the core fRad within an acceptable range, the X-divertor is shown to substantially lower the core confinement requirement for a fusion reactor. As a bonus, the XD also enables the use of liquid metals by reducing the MHD drag. A possible series of experiments for an efficient and attractive path to practical fusion power is suggested.

Science & Technology Review October/November 2016

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vogt, R. L.; Meissner, C. N.; Kotta, P. R.

At Lawrence Livermore National Laboratory, we focus on science and technology research to ensure our nation’s security. We also apply that expertise to solve other important national problems in energy, bioscience, and the environment. Science & Technology Review is published eight times a year to communicate, to a broad audience, the Laboratory’s scientific and technological accomplishments in fulfilling its primary missions. The publication’s goal is to help readers understand these accomplishments and appreciate their value to the individual citizen, the nation, and the world. The Laboratory is operated by Lawrence Livermore National Security, LLC (LLNS), for the Department of Energy’smore » National Nuclear Security Administration. LLNS is a partnership involving Bechtel National, University of California, Babcock & Wilcox, Washington Division of URS Corporation, and Battelle in affiliation with Texas A&M University. More information about LLNS is available online at www.llnsllc.com. Please address any correspondence (including name and address changes) to S&TR, Mail Stop L-664, Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94551, or telephone (925) 423-3893. Our e-mail address is str-mail@llnl.gov. S&TR is available on the Web at str.llnl.gov.« less

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.

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

2016-01-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

2016-03-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

Report detailing comparative analysis of results from high flux isotope reactor and national institute of standards technology small-angle neutron scattering experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sokolov, Mikhail A.; Littrell, Ken; Wells, Peter

, see Ref. [5] and [6] for details. UCSB has performed a large number of SANS experiments in the past at the National Institute of Standards and Technology (NIST) Center for Neutron Research (NCNR). These data are taken from RPV steels irradiated in a wide range of flux-fluence space and will be very useful in comparing to the upcoming UCSB ATR-2 irradiation characterization since most of the SANS experiments with ATR-2 materials will be performed at ORNL High Flux Isotope Reactor (HFIR). However in the previous report [7], some discrepancies were observed between HFIR and NCNR generated data. One of the hypotheses was that there was some kind of extra scattering occurring off the sample holders that results in the HFIR curves falling above the NCNR curves. To test this hypothesis, UCSB provided thermally aged samples that have been previously run at NCNR to ORNL for testing at HFIR while ORNL performed some improvements to experimental set up at HFIR. This report provides the status for the Level 3 Milestone (M3LW-15OR0402013), Complete report detailing comparative analysis of results from High Flux Isotope Reactor and National Institute of Standards and Technology small-angle neutron scattering experiments. This milestone is associated with small-angle neutron scattering characterization at the High Flux Isotope Reactor of various model alloys that had been previously characterized at NCNR by UCSB.« less

Biogenic origin of coalbed gas in the northern Gulf of Mexico Coastal Plain, U.S.A

USGS Publications Warehouse

Warwick, Peter D.; Breland, F. Clayton; Hackley, Paul C.

2008-01-01

New coal-gas exploration and production in northern Louisiana and south-central Mississippi, Gulf of Mexico Basin, is focused on the Wilcox Group (Paleocene–Eocene), where the depth to targeted subbituminous C to high volatile C bituminous coal beds ranges from 300 to 1680 m, and individual coal beds have a maximum thickness of about 6 m. Total gas content (generally excluding residual gas) of the coal beds ranges from less than 0.37 cm3/g (as-analyzed or raw basis; 1.2 cm3/g, dry, ash free basis, daf) at depths less than 400 m, to greater than 7.3 cm3/g (as-analyzed basis; 8.76 cm3/g, daf) in deeper (> 1,500 m) parts of the basin. About 20 Wilcox coal-gas wells in northern Louisiana produce from 200 to 6485 m3 of gas/day and cumulative gas production from these wells is approximately 25 million m3 (as of December, 2006). U.S. Geological Survey assessment of undiscovered, technically recoverable gas resources in the Gulf of Mexico Coastal Plain, including northern and south-central Mississippi, indicates that coal beds of the Wilcox Group contain an estimated mean total 109.3 million m3 (3.86 trillion ft3) of producible natural gas.To determine the origin of the Wilcox Group coal gases in northern Louisiana, samples of gas, water, and oil were collected from Wilcox coal and sandstone reservoirs and from under- and overlying Late Cretaceous and Eocene carbonate and sandstone reservoirs. Isotopic data from Wilcox coal-gas samples have an average δ13CCH4 value of − 62.6‰ VPDB (relative to Vienna Peedee Belemnite) and an average δDCH4 value of − 199.9‰ VSMOW (relative to Vienna Standard Mean Ocean Water). Values of δ13CCO2 range from − 25.4 to 3.42‰ VPDB. Produced Wilcox saline water collected from oil, conventional gas, and coalbed gas wells have δDH2O values that range from − 27.3 to − 18.0‰ VSMOW. These data suggest that the coal gases primarily are generated in saline formation water by bacterial reduction of CO2

Support vector machines for nuclear reactor state estimation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zavaljevski, N.; Gross, K. C.

2000-02-14

Validation of nuclear power reactor signals is often performed by comparing signal prototypes with the actual reactor signals. The signal prototypes are often computed based on empirical data. The implementation of an estimation algorithm which can make predictions on limited data is an important issue. A new machine learning algorithm called support vector machines (SVMS) recently developed by Vladimir Vapnik and his coworkers enables a high level of generalization with finite high-dimensional data. The improved generalization in comparison with standard methods like neural networks is due mainly to the following characteristics of the method. The input data space is transformedmore » into a high-dimensional feature space using a kernel function, and the learning problem is formulated as a convex quadratic programming problem with a unique solution. In this paper the authors have applied the SVM method for data-based state estimation in nuclear power reactors. In particular, they implemented and tested kernels developed at Argonne National Laboratory for the Multivariate State Estimation Technique (MSET), a nonlinear, nonparametric estimation technique with a wide range of applications in nuclear reactors. The methodology has been applied to three data sets from experimental and commercial nuclear power reactor applications. The results are promising. The combination of MSET kernels with the SVM method has better noise reduction and generalization properties than the standard MSET algorithm.« less

High-Temperature Gas-Cooled Test Reactor Point Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

2016-04-01

A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

78 FR 69139 - Physical Security-Design Certification and Operating Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-18

... scheduled to close on October 30, 2013. The Nuclear Energy Institute (NEI) submitted a letter on October 9... NUCLEAR REGULATORY COMMISSION [NRC-2013-0225] Physical Security--Design Certification and Operating Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard review plan--draft section...

Future Scenarios for Fission Based Reactors

NASA Astrophysics Data System (ADS)

David, S.

2005-04-01

The coming century will see the exhaustion of standard fossil fuels, coal, gas and oil, which today represent 75% of the world energy production. Moreover, their use will have caused large-scale emission of greenhouse gases (GEG), and induced global climate change. This problem is exacerbated by a growing world energy demand. In this context, nuclear power is the only GEG-free energy source available today capable of responding significantly to this demand. Some scenarios consider a nuclear energy production of around 5 Gtoe in 2050, wich would represent a 20% share of the world energy supply. Present reactors generate energy from the fission of U-235 and require around 200 tons of natural Uranium to produce 1GWe.y of energy, equivalent to the fission of one ton of fissile material. In a scenario of a significant increase in nuclear energy generation, these standard reactors will consume the whole of the world's estimated Uranium reserves in a few decades. However, natural Uranium or Thorium ore, wich are not themselves fissile, can produce a fissile material after a neutron capture ( 239Pu and 233U respectively). In a breeder reactor, the mass of fissile material remains constant, and the fertile ore is the only material to be consumed. In this case, only 1 ton of natural ore is needed to produce 1GWe.y. Thus, the breeding concept allows optimal use of fertile ore and development of sustainable nuclear energy production for several thousand years into the future. Different sustainable nuclear reactor concepts are studied in the international forum "generation IV". Different types of coolant (Na, Pb and He) are studied for fast breeder reactors based on the Uranium cycle. The thermal Thorium cycle requires the use of a liquid fuel, which can be reprocessed online in order to extract the neutron poisons. This paper presents these different sustainable reactors, based on the Uranium or Thorium fuel cycles and will compare the different options in term of fissile

On heat loading, novel divertors, and fusion reactors

NASA Astrophysics Data System (ADS)

Kotschenreuther, M.; Valanju, P. M.; Mahajan, S. M.; Wiley, J. C.

2007-07-01

The limited thermal power handling capacity of the standard divertors (used in current as well as projected tokamaks) is likely to force extremely high (˜90%) radiation fractions frad in tokamak fusion reactors that have heating powers considerably larger than ITER [D. J. Campbell, Phys. Plasmas 8, 2041 (2001)]. Such enormous values of necessary frad could have serious and debilitating consequences on the core confinement, stability, and dependability for a fusion power reactor, especially in reactors with Internal Transport Barriers. A new class of divertors, called X-divertors (XD), which considerably enhance the divertor thermal capacity through a flaring of the field lines only near the divertor plates, may be necessary and sufficient to overcome these problems and lead to a dependable fusion power reactor with acceptable economics. X-divertors will lower the bar on the necessary confinement to bring it in the range of the present experimental results. Its ability to reduce the radiative burden imparts the X-divertor with a key advantage. Lower radiation demands allow sharply peaked density profiles that enhance the bootstrap fraction creating the possibility for a highly increased beta for the same beta normal discharges. The X-divertor emerges as a beta-enhancer capable of raising it by up to roughly a factor of 2.

Neck Torque Study Induced by Head-Borne Visual Augmentation Systems (VAS) in Ground-Based Applications. Version V1.2

DTIC Science & Technology

2010-04-01

Standard Issue M, L and  XL  ACH  AN/PVS‐14  Norotos, Standard Issue M, L and  XL  ACH  AN/PVS‐15  Norotos Low Profile, (USASOC) Wilcox Low Profile...WARCOM)  M, L and  XL  ACH  AN/PVS‐23  ITT F5050 (USASOC) M, L and  XL  ACH   Fusion Goggles  ITT F5050 (USASOC)  M, L and  XL  ACH and SOCOM  LWH – Large  AN...PEQ‐20  AN/PEQ‐20 ENVG Mount M, L and  XL  ACH  ADM‐NVG  ADM  Wilcox Mount  M, L and  XL  ACH and SOCOM  Lightweight Helmet (LWH) ‐ Large  FGS‐PI  ITT F5050

Washington State water quality temperature standards as related to reactor operation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ballowe, J.W.

1968-08-14

The purpose of this report is to provide a basic working tool for determining the relationship between the allowable temperature increase within the Columbia River reach at the Hanford Site and the actual temperature increase as associated with various reactor operating modes. This basic tool can be utilized for day-to-day operating purposes or for the achievement of historical information.

Analysis of standard reference materials by absolute INAA

NASA Astrophysics Data System (ADS)

Heft, R. E.; Koszykowski, R. F.

1981-07-01

Three standard reference materials: flyash, soil, and ASI 4340 steel, are analyzed by a method of absolute instrumental neutron activation analysis. Two different light water pool-type reactors were used to produce equivalent analytical results even though the epithermal to thermal flux ratio in one reactor was higher than that in the other by a factor of two.

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 1 2014-01-01 2014-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

Development and industrial application of oil-reburning for NO{sub x} emission control in utility boilers

DOE Office of Scientific and Technical Information (OSTI.GOV)

De Michele, G.; Pasini, S.; Tarli, R.

1996-01-01

ENEL is conducting a comprehensive modification program within its generating system, in order to comply with the new Italian air quality standards for fossil fuel-fired power plants, which set a limit for NO{sub x} of 200 mg/Nm{sup 3} corrected to 3% O{sub 2} for oil and gas and to 6% O{sub 2} for coal. Among all combustion modification technologies reburning has proven particularly attractive, since it has been demonstrated that it generally permits to satisfy the regulatory requirements in gas and oil fired units, thus avoiding the use of SCR. The ENEL generating system essentially employs tangentially fired (TF) andmore » front/opposed wall fired boilers belonging, respectively, to the CEI and Babcock & Wilcox (B&W) technology, designed and constructed by Ansaldo and Franco Tosi (now bought by Ansaldo). In this framework ENEL has signed respectively with Ansaldo Energia and Combustion Engineering Inc. (CEI), and with Ansaldo Energia two separate agreements to apply Reburn Technology in oil and gas, tangentially-fired (TF) utility boilers, the first, and oil and gas wall-fired (WF) utility boilers, the second, both in Italy and abroad. This paper outlines the technical knowledge available for the design of reburn systems for a retrofit application and describes the main results obtained, after retrofit, at Torvaldaliga No. 2 power station, 320 MW{sub e} (TF), firing both oil and gas as reburn fuels, and at Cassano unit No. 1, 75 MW{sub e} (WF), firing gas as main and reburn fuels. Reference is also made to the development of the projects for the application of the technology at Monfalcone, 320 MW{sub e} (WF), in the oil over oil configuration, whose demonstration is planned for the Autumn of 1995, and at Porto Tolle unit No. 1, 660 MW{sub e} (TF), that is planned to start in January 1996.« less

How much does the MSW effect contribute to the reactor antineutrino anomaly?

DOE Office of Scientific and Technical Information (OSTI.GOV)

Valdiviesso, G. A.

2015-05-15

It has been pointed out that there is a 5.7 ± 2.3 discrepancy between the predicted and the observed reactor antineutrino flux in very short baseline experiments. Several causes for this anomaly have been discussed, including a possible non-standard forth sterile neutrino. In order to quantify how much non-standard this anomaly really is, the standard MSW effect is reviewed. Knowing that reactor antineutrinos are produced in a dense medium (the nuclear fuel) and is usually detected in a less dense one (water, or scintillator), non-adiabatic effects are expected to happen, creating a difference between the creation and detection mixing angles.

10 CFR 2.110 - Filing and administrative action on submittals for standard design approval or early review of...

Code of Federal Regulations, 2012 CFR

2012-01-01

..., Office of New Reactors, or Director, Office of Nuclear Reactor Regulation, as appropriate shall publish... Nuclear Reactor Regulation, as appropriate shall publish in the Federal Register a determination as to... standard design approval or early review of site suitability issues. 2.110 Section 2.110 Energy NUCLEAR...

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.

10 CFR 2.110 - Filing and administrative action on submittals for standard design approval or early review of...

Code of Federal Regulations, 2013 CFR

2013-01-01

..., Office of Nuclear Reactor Regulation, as appropriate shall publish in the Federal Register a notice of... Director, Office of New Reactors or Director, Office of Nuclear Reactor Regulation, as appropriate shall... standard design approval or early review of site suitability issues. 2.110 Section 2.110 Energy NUCLEAR...

10 CFR 2.110 - Filing and administrative action on submittals for standard design approval or early review of...

Code of Federal Regulations, 2014 CFR

2014-01-01

..., Office of Nuclear Reactor Regulation, as appropriate shall publish in the Federal Register a notice of... Director, Office of New Reactors or Director, Office of Nuclear Reactor Regulation, as appropriate shall... standard design approval or early review of site suitability issues. 2.110 Section 2.110 Energy NUCLEAR...

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

..., review of standard referenced design approvals, special projects, inspections and import and export... standard referenced design approvals, special projects, inspections and import and export licenses..., approvals of facility standard reference designs, re-qualification and replacement examinations for reactor...

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

Code of Federal Regulations, 2014 CFR

2014-01-01

..., review of standard referenced design approvals, special projects, inspections and import and export... standard referenced design approvals, special projects, inspections and import and export licenses..., approvals of facility standard reference designs, re-qualification and replacement examinations for reactor...

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

Code of Federal Regulations, 2013 CFR

2013-01-01

..., review of standard referenced design approvals, special projects, inspections and import and export... standard referenced design approvals, special projects, inspections and import and export licenses..., approvals of facility standard reference designs, re-qualification and replacement examinations for reactor...

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

..., review of standard referenced design approvals, special projects, inspections and import and export... standard referenced design approvals, special projects, inspections and import and export licenses..., approvals of facility standard reference designs, re-qualification and replacement examinations for reactor...

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. .

COMO: a numerical model for predicting furnace performance in axisymmetric geometries. Volume 1. Technical summary. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fiveland, W.A.; Oberjohn, W.J.; Cornelius, D.K.

1985-12-01

This report summarizes the work conducted during a 30-month contract with the United States Department of Energy (DOE) Pittsburgh Energy Technology Center (PETC). The general objective is to develop and verify a computer code capable of modeling the major aspects of pulverized coal combustion. Achieving this objective will lead to design methods applicable to industrial and utility furnaces. The combustion model (COMO) is based mainly on an existing Babcock and Wilcox (B and W) computer program. The model consists of a number of relatively independent modules that represent the major processes involved in pulverized coal combustion: flow, heterogeneous and homogeneousmore » chemical reaction, and heat transfer. As models are improved or as new ones are developed, this modular structure allows portions of the COMO model to be updated with minimal impact on the remainder of the program. The report consists of two volumes. This volume (Volume 1) contains a technical summary of the COMO model, results of predictions for gas phase combustion, pulverized coal combustion, and a detailed description of the COMO model. Volume 2 is the Users Guide for COMO and contains detailed instructions for preparing the input data and a description of the program output. Several example cases have been included to aid the user in usage of the computer program for pulverized coal applications. 66 refs., 41 figs., 21 tabs.« less

Simulation of 0.3 MWt AFBC test rig burning Turkish lignites

DOE Office of Scientific and Technical Information (OSTI.GOV)

Selcuk, N.; Degirmenci, E.; Oymak, O.

1997-12-31

A system model coupling bed and freeboard models for continuous combustion of lignite particles of wide size distribution burning in their own ash in a fluidized bed combustor was modified to incorporate: (1) a procedure for faster computation of particle size distributions (PSDs) without any sacrifice in accuracy; (2) energy balance on char particles for the determination of variation of temperature with particle size, (3) plug flow assumption for the interstitial gas. An efficient and accurate computer code developed for the solution of the conservation equations for energy and chemical species was applied to the prediction of the behavior ofmore » a 0.3 MWt AFBC test rig burning low quality Turkish lignites. The construction and operation of the test rig was carried out within the scope of a cooperation agreement between Middle East Technical University (METU) and Babcock and Wilcox GAMA (BWG) under the auspices of Canadian International Development Agency (CIDA). Predicted concentration and temperature profiles and particle size distributions of solid streams were compared with measured data and found to be in reasonable agreement. The computer code replaces the conventional numerical integration of the analytical solution of population balance with direct integration in ODE form by using a powerful integrator LSODE (Livermore Solver for Ordinary Differential Equations) resulting in two orders of magnitude decrease in CPU (Central Processing Unit) time.« less

TRAC-PF1/MOD1 pretest predictions of MIST experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

Los Alamos National Laboratory is a participant in the Integral System Test (IST) program initiated in June 1983 to provide integral system test data on specific issues and phenomena relevant to post small-break loss-of-coolant accidents (SBLOCAs) in Babcock and Wilcox plant designs. The Multi-Loop Integral System Test (MIST) facility is the largest single component in the IST program. During Fiscal Year 1986, Los Alamos performed five MIST pretest analyses. The five experiments were chosen on the basis of their potential either to approach the facility limits or to challenge the predictive capability of the TRAC-PF1/MOD1 code. Three SBLOCA tests weremore » examined which included nominal test conditions, throttled auxiliary feedwater and asymmetric steam-generator cooldown, and reduced high-pressure-injection (HPI) capacity, respectively. Also analyzed were two ''feed-and-bleed'' cooling tests with reduced HPI and delayed HPI initiation. Results of the tests showed that the MIST facility limits would not be approached in the five tests considered. Early comparisons with preliminary test data indicate that the TRAC-PF1/MOD1 code is correctly calculating the dominant phenomena occurring in the MIST facility during the tests. Posttest analyses are planned to provide a quantitative assessment of the code's ability to predict MIST transients.« less

Demonstration of coal reburning for cyclone boiler NO{sub x} control. Appendix, Book 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

Based on the industry need for a pilot-scale cyclone boiler simulator, Babcock Wilcox (B&W) designed, fabricated, and installed such a facility at its Alliance Research Center (ARC) in 1985. The project involved conversion of an existing pulverized coal-fired facility to be cyclone-firing capable. Additionally, convective section tube banks were installed in the upper furnace in order to simulate a typical boiler convection pass. The small boiler simulator (SBS) is designed to simulate most fireside aspects of full-size utility boilers such as combustion and flue gas emissions characteristics, fireside deposition, etc. Prior to the design of the pilot-scale cyclone boiler simulator,more » the various cyclone boiler types were reviewed in order to identify the inherent cyclone boiler design characteristics which are applicable to the majority of these boilers. The cyclone boiler characteristics that were reviewed include NO{sub x} emissions, furnace exit gas temperature (FEGT) carbon loss, and total furnace residence time. Previous pilot-scale cyclone-fired furnace experience identified the following concerns: (1) Operability of a small cyclone furnace (e.g., continuous slag tapping capability). (2) The optimum cyclone(s) configuration for the pilot-scale unit. (3) Compatibility of NO{sub x} levels, carbon burnout, cyclone ash carryover to the convection pass, cyclone temperature, furnace residence time, and FEGT.« less

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, Sharon S.; Warwick, Peter D.; Ruppert, Leslie F.; Pontolillo, James

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

77 FR 62270 - Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-12

... for Passive Advanced Light Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard... Passive Advanced Light Water Reactors.'' The current SRP does not contain guidance on the proposed RTNSS for Passive Advance Light Water Reactors. DATES: Submit comments by November 13, 2012. Comments...

Projected Standard on neutron skyshine. [Skyshine

DOE Office of Scientific and Technical Information (OSTI.GOV)

Westfall, R.M.; Williams, D.S.

1987-07-01

Current interest in neutron skyshine arises from the application of dry fuel handling and storage techniques at reactor sites, at the proposed monitored retrievable storage facility and at other facilities being considered as part of the civilian radioactive waste management programs. The chairman of Standards Subcommittee ANS-6, Radiation Protection and Shielding, has requested that a work group be formed to characterize the neutron skyshine problem and, if necessary, prepare a draft Standard. The work group is comprised of representatives of storage cask vendors, architect engineering firms, nuclear utilities, the academic community and staff members of national laboratories and government agencies.more » The purpose of this presentation summary is to describe the activities of the work group and the scope and contents of the projected Standard, ANS-6.6.2, ''Calculation and Measurement of Direct and Scattered Neutron Radiation from Nuclear Power Operations.'' The specific source under consideration by the work group is an array of dry fuel casks located at a reactor site. However, it is recognized that the scope of the standard should be broad enough to encompass other neutron sources. The Standard will define appropriate methodology for properly characterizing the neutron dose due to skyshine. This dose characterization is necessary, for example, in demonstrating compliance with pertinent regulatory criteria.« less

Standards in nuclear science and technology. A bibliography

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1973-09-01

Abstracts of 1803 U. S. and non-U. S. publications concerning a broad range of standards used in nuclear science and technology are included. The publication dates span the period 1962 through 1972, inclusive. Abstracts are arranged chronologically within four categories entitled Reactors and Engineering, Instruments and Calibration, Radiation and Radiation Protection, and Miscellaneous. A subject index is also included. (auth)

Nuclear reactor construction with bottom supported reactor vessel

DOEpatents

Sharbaugh, John E.

1987-01-01

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

Implementation of New Reactivity Measurement System and New Reactor Noise Analysis Equipment in a VVER-440 Nuclear Power Plant

NASA Astrophysics Data System (ADS)

Vegh, János; Kiss, Sándor; Lipcsei, Sándor; Horvath, Csaba; Pos, István; Kiss, Gábor

2010-10-01

The paper deals with two recently developed, high-precision nuclear measurement systems installed at the VVER-440 units of the Hungarian Paks NPP. Both developments were motivated by the reactor power increase to 108%, and by the planned plant service time extension. The first part describes the RMR start-up reactivity measurement system with advanced services. High-precision picoampere meters were installed at each reactor unit and measured ionization chamber current signals are handled by a portable computer providing data acquisition and online reactivity calculation service. Detailed offline evaluation and analysis of reactor start-up measurements can be performed on the portable unit, too. The second part of the paper describes a new reactor noise diagnostics system using state-of-the-art data acquisition hardware and signal processing methods. Details of the new reactor noise measurement evaluation software are also outlined. Noise diagnostics at Paks NPP is a standard tool for core anomaly detection and for long-term noise trend monitoring. Regular application of these systems is illustrated by real plant data, e.g., results of standard reactivity measurements during a reactor startup session are given. Noise applications are also illustrated by real plant measurements; results of core anomaly detection are presented.

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 %).

Globalization of ASME Nuclear Codes and Standards

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 countriesmore » 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)« less

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay

DOE PAGES

An, F. P.; Balantekin, A. B.; Band, H. R.; ...

2017-06-19

Here, the Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 GW th reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective 239Pu fission fractions F 239 from 0.25 to 0.35, Daya Bay measures an average IBD yield ¯σf of (5.90±0.13)×10 –43 cm 2/fission and a fuel-dependent variation in the IBDmore » yield, dσ f/dF 239, of (–1.86±0.18)×10 –43 cm 2/fission. This observation rejects the hypothesis of a constant antineutrino flux as a function of the 239Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1σ. This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes 235U, 239Pu, 238U, and 241Pu. Based on measured IBD yield variations, yields of (6.17±0.17) and (4.27±0.26)×10 –43 cm 2/fission have been determined for the two dominant fission parent isotopes 235U and 239Pu. A 7.8% discrepancy between the observed and predicted 235U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.« less

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay.

PubMed

An, F P; Balantekin, A B; Band, H R; Bishai, M; Blyth, S; Cao, D; Cao, G F; Cao, J; Chan, Y L; Chang, J F; Chang, Y; Chen, H S; Chen, Q Y; Chen, S M; Chen, Y X; Chen, Y; Cheng, J; Cheng, Z K; Cherwinka, J J; Chu, M C; Chukanov, A; Cummings, J P; Ding, Y Y; Diwan, M V; Dolgareva, M; Dove, J; Dwyer, D A; Edwards, W R; Gill, R; Gonchar, M; Gong, G H; Gong, H; Grassi, M; Gu, W Q; Guo, L; Guo, X H; Guo, Y H; Guo, Z; Hackenburg, R W; Hans, S; He, M; Heeger, K M; Heng, Y K; Higuera, A; Hsiung, Y B; Hu, B Z; Hu, T; Huang, E C; Huang, H X; Huang, X T; Huang, Y B; Huber, P; Huo, W; Hussain, G; Jaffe, D E; Jen, K L; Ji, X P; Ji, X L; Jiao, J B; Johnson, R A; Jones, D; Kang, L; Kettell, S H; Khan, A; Kohn, S; Kramer, M; Kwan, K K; Kwok, M W; Langford, T J; Lau, K; Lebanowski, L; Lee, J; Lee, J H C; Lei, R T; Leitner, R; Leung, J K C; Li, C; Li, D J; Li, F; Li, G S; Li, Q J; Li, S; Li, S C; Li, W D; Li, X N; Li, X Q; Li, Y F; Li, Z B; Liang, H; Lin, C J; Lin, G L; Lin, S; Lin, S K; Lin, Y-C; Ling, J J; Link, J M; Littenberg, L; Littlejohn, B R; Liu, J L; Liu, J C; Loh, C W; Lu, C; Lu, H Q; Lu, J S; Luk, K B; Ma, X Y; Ma, X B; Ma, Y Q; Malyshkin, Y; Martinez Caicedo, D A; McDonald, K T; McKeown, R D; Mitchell, I; Nakajima, Y; Napolitano, J; Naumov, D; Naumova, E; Ngai, H Y; Ochoa-Ricoux, J P; Olshevskiy, A; Pan, H-R; Park, J; Patton, S; Pec, V; Peng, J C; Pinsky, L; Pun, C S J; Qi, F Z; Qi, M; Qian, X; Qiu, R M; Raper, N; Ren, J; Rosero, R; Roskovec, B; Ruan, X C; Steiner, H; Stoler, P; Sun, J L; Tang, W; Taychenachev, D; Treskov, K; Tsang, K V; Tull, C E; Viaux, N; Viren, B; Vorobel, V; Wang, C H; Wang, M; Wang, N Y; Wang, R G; Wang, W; Wang, X; Wang, Y F; Wang, Z; Wang, Z; Wang, Z M; Wei, H Y; Wen, L J; Whisnant, K; White, C G; Whitehead, L; Wise, T; Wong, H L H; Wong, S C F; Worcester, E; Wu, C-H; Wu, Q; Wu, W J; Xia, D M; Xia, J K; Xing, Z Z; Xu, J L; Xu, Y; Xue, T; Yang, C G; Yang, H; Yang, L; Yang, M S; Yang, M T; Yang, Y Z; Ye, M; Ye, Z; Yeh, M; Young, B L; Yu, Z Y; Zeng, S; Zhan, L; Zhang, C; Zhang, C C; Zhang, H H; Zhang, J W; Zhang, Q M; Zhang, R; Zhang, X T; Zhang, Y M; Zhang, Y X; Zhang, Y M; Zhang, Z J; Zhang, Z Y; Zhang, Z P; Zhao, J; Zhou, L; Zhuang, H L; Zou, J H

2017-06-23

The Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 GW_{th} reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective ^{239}Pu fission fractions F_{239} from 0.25 to 0.35, Daya Bay measures an average IBD yield σ[over ¯]_{f} of (5.90±0.13)×10^{-43}  cm^{2}/fission and a fuel-dependent variation in the IBD yield, dσ_{f}/dF_{239}, of (-1.86±0.18)×10^{-43}  cm^{2}/fission. This observation rejects the hypothesis of a constant antineutrino flux as a function of the ^{239}Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1σ. This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes ^{235}U, ^{239}Pu, ^{238}U, and ^{241}Pu. Based on measured IBD yield variations, yields of (6.17±0.17) and (4.27±0.26)×10^{-43}  cm^{2}/fission have been determined for the two dominant fission parent isotopes ^{235}U and ^{239}Pu. A 7.8% discrepancy between the observed and predicted ^{235}U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay

NASA Astrophysics Data System (ADS)

An, F. P.; Balantekin, A. B.; Band, H. R.; Bishai, M.; Blyth, S.; Cao, D.; Cao, G. F.; Cao, J.; Chan, Y. L.; Chang, J. F.; Chang, Y.; Chen, H. S.; Chen, Q. Y.; Chen, S. M.; Chen, Y. X.; Chen, Y.; Cheng, J.; Cheng, Z. K.; Cherwinka, J. J.; Chu, M. C.; Chukanov, A.; Cummings, J. P.; Ding, Y. Y.; Diwan, M. V.; Dolgareva, M.; Dove, J.; Dwyer, D. A.; Edwards, W. R.; Gill, R.; Gonchar, M.; Gong, G. H.; Gong, H.; Grassi, M.; Gu, W. Q.; Guo, L.; Guo, X. H.; Guo, Y. H.; Guo, Z.; Hackenburg, R. W.; Hans, S.; He, M.; Heeger, K. M.; Heng, Y. K.; Higuera, A.; Hsiung, Y. B.; Hu, B. Z.; Hu, T.; Huang, E. C.; Huang, H. X.; Huang, X. T.; Huang, Y. B.; Huber, P.; Huo, W.; Hussain, G.; Jaffe, D. E.; Jen, K. L.; Ji, X. P.; Ji, X. L.; Jiao, J. B.; Johnson, R. A.; Jones, D.; Kang, L.; Kettell, S. H.; Khan, A.; Kohn, S.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Langford, T. J.; Lau, K.; Lebanowski, L.; Lee, J.; Lee, J. H. C.; Lei, R. T.; Leitner, R.; Leung, J. K. C.; Li, C.; Li, D. J.; Li, F.; Li, G. S.; Li, Q. J.; Li, S.; Li, S. C.; Li, W. D.; Li, X. N.; Li, X. Q.; Li, Y. F.; Li, Z. B.; Liang, H.; Lin, C. J.; Lin, G. L.; Lin, S.; Lin, S. K.; Lin, Y.-C.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, J. L.; Liu, J. C.; Loh, C. W.; Lu, C.; Lu, H. Q.; Lu, J. S.; Luk, K. B.; Ma, X. Y.; Ma, X. B.; Ma, Y. Q.; Malyshkin, Y.; Martinez Caicedo, D. A.; McDonald, K. T.; McKeown, R. D.; Mitchell, I.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Ngai, H. Y.; Ochoa-Ricoux, J. P.; Olshevskiy, A.; Pan, H.-R.; Park, J.; Patton, S.; Pec, V.; Peng, J. C.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Qiu, R. M.; Raper, N.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Steiner, H.; Stoler, P.; Sun, J. L.; Tang, W.; Taychenachev, D.; Treskov, K.; Tsang, K. V.; Tull, C. E.; Viaux, N.; Viren, B.; Vorobel, V.; Wang, C. H.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, W.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Wei, H. Y.; Wen, L. J.; Whisnant, K.; White, C. G.; Whitehead, L.; Wise, T.; Wong, H. L. H.; Wong, S. C. F.; Worcester, E.; Wu, C.-H.; Wu, Q.; Wu, W. J.; Xia, D. M.; Xia, J. K.; Xing, Z. Z.; Xu, J. L.; Xu, Y.; Xue, T.; Yang, C. G.; Yang, H.; Yang, L.; Yang, M. S.; Yang, M. T.; Yang, Y. Z.; Ye, M.; Ye, Z.; Yeh, M.; Young, B. L.; Yu, Z. Y.; Zeng, S.; Zhan, L.; Zhang, C.; Zhang, C. C.; Zhang, H. H.; Zhang, J. W.; Zhang, Q. M.; Zhang, R.; Zhang, X. T.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Y. M.; Zhang, Z. J.; Zhang, Z. Y.; Zhang, Z. P.; Zhao, J.; Zhou, L.; Zhuang, H. L.; Zou, J. H.; Daya Bay Collaboration

2017-06-01

The Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 G Wth reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective 239Pu fission fractions F239 from 0.25 to 0.35, Daya Bay measures an average IBD yield σ¯f of (5.90 ±0.13 )×10-43 cm2/fission and a fuel-dependent variation in the IBD yield, d σf/d F239, of (-1.86 ±0.18 )×10-43 cm2/fission . This observation rejects the hypothesis of a constant antineutrino flux as a function of the 239Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1 σ . This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes 235U, 239Pu, 238U, and 241Pu. Based on measured IBD yield variations, yields of (6.17 ±0.17 ) and (4.27 ±0.26 )×10-43 cm2 /fission have been determined for the two dominant fission parent isotopes 235U and 239Pu. A 7.8% discrepancy between the observed and predicted 235U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay

DOE Office of Scientific and Technical Information (OSTI.GOV)

An, F. P.; Balantekin, A. B.; Band, H. R.

Here, the Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 GW th reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective 239Pu fission fractions F 239 from 0.25 to 0.35, Daya Bay measures an average IBD yield ¯σf of (5.90±0.13)×10 –43 cm 2/fission and a fuel-dependent variation in the IBDmore » yield, dσ f/dF 239, of (–1.86±0.18)×10 –43 cm 2/fission. This observation rejects the hypothesis of a constant antineutrino flux as a function of the 239Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1σ. This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes 235U, 239Pu, 238U, and 241Pu. Based on measured IBD yield variations, yields of (6.17±0.17) and (4.27±0.26)×10 –43 cm 2/fission have been determined for the two dominant fission parent isotopes 235U and 239Pu. A 7.8% discrepancy between the observed and predicted 235U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.« less

An RF-Powered Micro-Reactor for Efficient Extraction and Hydrolysis

NASA Astrophysics Data System (ADS)

Scott, V.

2014-12-01

An RF sample-processing micro-reactor that was developed as part of potential in situ Exploration Missions to inner- and outer-planetary bodies was designed to utilize aqueous solutions subjected to 60 GHz radiation at 730 mW of input power to extract target organic compounds and molecular and inorganic ions as well as to hydrolyze complex polymeric materials. Successful identification and characterization of these molecules relies on the sample-processing techniques utilized alongside state-of-the-art detection and analysis. For mass and power restrictions put on space exploration missions, smaller and more efficient instruments are highly desirable. The RF micro-reactor potentially offers a simplified alternative to the typical gold-standard extractions that often use solvents, chemicals, and conditions that can vary wildly and depend on the targeted molecules. Instead, this instrument uses a single solvent ­— water — that can be "tuned" under the different experimental conditions, leveraging the operating principles of the Sub-Critical Water Extractor. Proof-of-concept experiments examining the hydrolysis of glycosidic and peptide bonds were successful in demonstrating the RF micro-reactor's capabilities. Progress toward coupling the reactor with a micro-scale sample-handling system enabling slurry delivery has been made and preliminary results on heterogeneous reactions and extractions will be presented.

Thorium and Molten Salt Reactors: Essential Questions for Classroom Discussions

NASA Astrophysics Data System (ADS)

DiLisi, Gregory A.; Hirsch, Allison; Murray, Meredith; Rarick, Richard

2018-04-01

A little-known type of nuclear reactor called the "molten salt reactor" (MSR), in which nuclear fuel is dissolved in a liquid carrier salt, was proposed in the 1940s and developed at the Oak Ridge National Laboratory in the 1960s. Recently, the MSR has generated renewed interest as a remedy for the drawbacks associated with conventional uranium-fueled light-water reactors (LWRs) in use today. Particular attention has been given to the "thorium molten salt reactor" (TMSR), an MSR engineered specifically to use thorium as its fuel. The purpose of this article is to encourage the TPT community to incorporate discussions of MSRs and the thorium fuel cycle into courses such as "Physics and Society" or "Frontiers of Physics." With this in mind, we piloted a pedagogical approach with 27 teachers in which we described the underlying physics of the TMSR and posed five essential questions for classroom discussions. We assumed teachers had some preexisting knowledge of nuclear reactions, but such prior knowledge was not necessary for inclusion in the classroom discussions. Overall, our material was perceived as a real-world example of physics, fit into a standards-based curriculum, and filled a need in the teaching community for providing unbiased references of alternative energy technologies.

Plant maintenance and advanced reactors issue, 2004

DOE Office of Scientific and Technical Information (OSTI.GOV)

Agnihotri, Newal

2004-09-15

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Optimism about the future of nuclear power, by Ruth G. Shaw, Duke Power Company; Licensed in three countries, by GE Energy; Enhancing public acceptance, by Westinghouse Electric Company; Standardized MOV program, by Ted Neckowicz, Exelon; Inservice testing, by Steven Unikewicz, U.S. Nuclear Regulatory Commission; Asian network for education, Fatimah Mohd Amin, Malaysian Institute for Nuclear Technology Research; and, Cooling water intake optimization, by Jeffrey M. Jones and Bert Mayer, P.E., Framatome ANP.

Modeling residence-time distribution in horizontal screw hydrolysis reactors

DOE PAGES

Sievers, David A.; Stickel, Jonathan J.

2017-10-12

The dilute-acid thermochemical hydrolysis step used in the production of liquid fuels from lignocellulosic biomass requires precise residence-time control to achieve high monomeric sugar yields. Difficulty has been encountered reproducing residence times and yields when small batch reaction conditions are scaled up to larger pilot-scale horizontal auger-tube type continuous reactors. A commonly used naive model estimated residence times of 6.2-16.7 min, but measured mean times were actually 1.4-2.2 the estimates. Here, this study investigated how reactor residence-time distribution (RTD) is affected by reactor characteristics and operational conditions, and developed a method to accurately predict the RTD based on key parameters.more » Screw speed, reactor physical dimensions, throughput rate, and process material density were identified as major factors affecting both the mean and standard deviation of RTDs. The general shape of RTDs was consistent with a constant value determined for skewness. The Peclet number quantified reactor plug-flow performance, which ranged between 20 and 357.« less

Modeling residence-time distribution in horizontal screw hydrolysis reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sievers, David A.; Stickel, Jonathan J.

The dilute-acid thermochemical hydrolysis step used in the production of liquid fuels from lignocellulosic biomass requires precise residence-time control to achieve high monomeric sugar yields. Difficulty has been encountered reproducing residence times and yields when small batch reaction conditions are scaled up to larger pilot-scale horizontal auger-tube type continuous reactors. A commonly used naive model estimated residence times of 6.2-16.7 min, but measured mean times were actually 1.4-2.2 the estimates. Here, this study investigated how reactor residence-time distribution (RTD) is affected by reactor characteristics and operational conditions, and developed a method to accurately predict the RTD based on key parameters.more » Screw speed, reactor physical dimensions, throughput rate, and process material density were identified as major factors affecting both the mean and standard deviation of RTDs. The general shape of RTDs was consistent with a constant value determined for skewness. The Peclet number quantified reactor plug-flow performance, which ranged between 20 and 357.« less

151. ARAIII Reactor building (ARA608) Details of reactor pit and ...

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

151. ARA-III Reactor building (ARA-608) Details of reactor pit and instrument plan. Aerojet-general 880-area/GCRE-608-T-19. Date: November 1958. Ineel index code no. 063-0608-25-013-102678. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

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.

NASA Trapezoidal Wing Simulation Using Stress-w and One- and Two-Equation Turbulence Models

NASA Technical Reports Server (NTRS)

Rodio, J. J.; Xiao, X; Hassan, H. A.; Rumsey, C. L.

2014-01-01

The Wilcox 2006 stress-omega model (also referred to as WilcoxRSM-w2006) has been implemented in the NASA Langley code CFL3D and used to study a variety of 2-D and 3-D configurations. It predicted a variety of basic cases reasonably well, including secondary flow in a supersonic rectangular duct. One- and two-equation turbulence models that employ the Boussinesq constitutive relation were unable to predict this secondary flow accurately because it is driven by normal turbulent stress differences. For the NASA trapezoidal wing at high angles of attack, the WilcoxRSM-w2006 model predicted lower maximum lift than experiment, similar to results of a two-equation model.

Uranium in NIMROC standard igneous rock samples

NASA Technical Reports Server (NTRS)

Rowe, M. W.; Herndon, J. M.

1976-01-01

Results are reported for analysis of the uranium in multiple samples of each of six igneous-rock standards (dunite, granite, lujavrite, norite, pyroxenite, and syenite) prepared as geochemical reference standards for elemental and isotopic compositions. Powdered rock samples were examined by measuring delayed neutron emission after irradiation with a flux of the order of 10 to the 13th power neutrons/sq cm per sec in a nuclear reactor. The measurements are shown to compare quite favorably with previous uranium determinations for other standard rock samples.

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.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 philosophymore » 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.« less

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.

Multi-reactor power system configurations for multimegawatt nuclear electric propulsion

NASA Technical Reports Server (NTRS)

George, Jeffrey A.

1991-01-01

A modular, multi-reactor power system and vehicle configuration for piloted nuclear electric propulsion (NEP) missions to Mars is presented. Such a design could provide enhanced system and mission reliability, allowing a comfortable safety margin for early manned flights, and would allow a range of piloted and cargo missions to be performed with a single power system design. Early use of common power modules for cargo missions would also provide progressive flight experience and validation of standardized systems for use in later piloted applications. System and mission analysis are presented to compare single and multi-reactor configurations for piloted Mars missions. A conceptual design for the Hydra modular multi-reactor NEP vehicle is presented.

High Temperature Gas Reactors: Assessment of Applicable Codes and Standards

DOE Office of Scientific and Technical Information (OSTI.GOV)

McDowell, Bruce K.; Nickolaus, James R.; Mitchell, Mark R.

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, simplemore » 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

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

Regional geohydrology of the northern Louisiana salt-dome basin; Part II, Geohydrologic maps of the Tertiary aquifers and related confining layers

USGS Publications Warehouse

Ryals, G.N.

1984-01-01

Regional geohydrologic maps show the altitude of the base and the thickness of the aquifers of Tertiary age and related confining layers in the northern Louisiana salt-dome basin. The limit of freshwater in aquifers is also shown. The basin has an area of about 3,000 square miles, and four geologic units of Tertiary age contain regional aquifers. From oldest (deepest) to youngest, the aquifers are in the Wilcox Group, Carrizo Sand, Sparta Sand, and Cockfield Formation. As the Wilcox is hydraulically interconnected with the overlying Carrizo, they are treated as one hydrologic unit, the Wilcox-Carrizo aquifer. The aquifers are separated by confining layers that retard water movement. In the northwestern part of the area, the Wilcox-Carrizo aquifer is separated from the underlying sand facies of the Nacatoch Sand (Cretaceous age) by a confining layer composed of the Midway Group (Tertiary age) and the underlying Arkadelphia Marl and an upper clay and marl facies of the Nacatoch Sand (both of Cretaceous age). In the remainder of the area, the Wilcox-Carrizo aquifer is separated from an underlying Cretaceous aquifer comprised of the Tokio Formation and Brownstown Marl by the Midway Group and several underlying Cretaceous units which in order of increasing age are the Arkadelphia Maril, Nacatoch Sand, Saratoga Chalk, Marlbrook Marl , and Annona Chalk. The Wilcox-Carrizo aquifer is separated from the Sparta aquifer by the overyling Cane River Formation. The Sparta aquifer is separated from the Cockfield aquifer by the overlying Cook Mountain Formation. (USGS)

Update on reactors and reactor instruments in Asia

NASA Astrophysics Data System (ADS)

Rao, K. R.

1991-10-01

The 1980s have seen the commissioning of several medium flux (∼10 14 neutrons/cm 2s) research reactors in Asia. The reactors are based on indigenous design and development in India and China. At Dhruva reactor (India), a variety of neutron spectrometers have been established that have provided useful data related to the structure of high- Tc materials, phonon density of states, magnetic moment distributions and micellar aggregation during the last couple of years. Polarised neutron analysis, neutron interferometry and neutron spin echo methods are some of the new techniques under development. The spectrometers and associated automaton, detectors and neutron guides have all been indigenously developed. This paper summarises the developments and on-going activities in Bangladesh, China, India, Indonesia, Korea, Malaysia, the Philippines and Thailand.

10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

Code of Federal Regulations, 2010 CFR

2010-01-01

... Water Reactor A Appendix A to Part 52 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES... Rule for the U.S. Advanced Boiling Water Reactor I. Introduction Appendix A constitutes the standard design certification for the U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance with 10 CFR...

10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

Code of Federal Regulations, 2011 CFR

2011-01-01

... Water Reactor A Appendix A to Part 52 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES... Rule for the U.S. Advanced Boiling Water Reactor I. Introduction Appendix A constitutes the standard design certification for the U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance with 10 CFR...

Standard-less analysis of Zircaloy clad samples by an instrumental neutron activation method

NASA Astrophysics Data System (ADS)

Acharya, R.; Nair, A. G. C.; Reddy, A. V. R.; Goswami, A.

2004-03-01

A non-destructive method for analysis of irregular shape and size samples of Zircaloy has been developed using the recently standardized k0-based internal mono standard instrumental neutron activation analysis (INAA). The samples of Zircaloy-2 and -4 tubes, used as fuel cladding in Indian boiling water reactors (BWR) and pressurized heavy water reactors (PHWR), respectively, have been analyzed. Samples weighing in the range of a few tens of grams were irradiated in the thermal column of Apsara reactor to minimize neutron flux perturbations and high radiation dose. The method utilizes in situ relative detection efficiency using the γ-rays of selected activation products in the sample for overcoming γ-ray self-attenuation. Since the major and minor constituents (Zr, Sn, Fe, Cr and/or Ni) in these samples were amenable to NAA, the absolute concentrations of all the elements were determined using mass balance instead of using the concentration of the internal mono standard. Concentrations were also determined in a smaller size Zircaloy-4 sample by irradiating in the core position of the reactor to validate the present methodology. The results were compared with literature specifications and were found to be satisfactory. Values of sensitivities and detection limits have been evaluated for the elements analyzed.

PR-EDB: Power Reactor Embrittlement Database - Version 3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Subramani, Ranjit

2008-03-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industrymore » standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access

Metal fires and their implications for advanced reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nowlen, Steven Patrick; Figueroa, Victor G.; Olivier, Tara Jean

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 thesemore » 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

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.

Background studies for the MINER Coherent Neutrino Scattering reactor experiment

NASA Astrophysics Data System (ADS)

Agnolet, G.; Baker, W.; Barker, D.; Beck, R.; Carroll, T. J.; Cesar, J.; Cushman, P.; Dent, J. B.; De Rijck, S.; Dutta, B.; Flanagan, W.; Fritts, M.; Gao, Y.; Harris, H. R.; Hays, C. C.; Iyer, V.; Jastram, A.; Kadribasic, F.; Kennedy, A.; Kubik, A.; Lang, K.; Mahapatra, R.; Mandic, V.; Marianno, C.; Martin, R. D.; Mast, N.; McDeavitt, S.; Mirabolfathi, N.; Mohanty, B.; Nakajima, K.; Newhouse, J.; Newstead, J. L.; Ogawa, I.; Phan, D.; Proga, M.; Rajput, A.; Roberts, A.; Rogachev, G.; Salazar, R.; Sander, J.; Senapati, K.; Shimada, M.; Soubasis, B.; Strigari, L.; Tamagawa, Y.; Teizer, W.; Vermaak, J. I. C.; Villano, A. N.; Walker, J.; Webb, B.; Wetzel, Z.; Yadavalli, S. A.

2017-05-01

The proposed Mitchell Institute Neutrino Experiment at Reactor (MINER) experiment at the Nuclear Science Center at Texas A&M University will search for coherent elastic neutrino-nucleus scattering within close proximity (about 2 m) of a 1 MW TRIGA nuclear reactor core using low threshold, cryogenic germanium and silicon detectors. Given the Standard Model cross section of the scattering process and the proposed experimental proximity to the reactor, as many as 5-20 events/kg/day are expected. We discuss the status of preliminary measurements to characterize the main backgrounds for the proposed experiment. Both in situ measurements at the experimental site and simulations using the MCNP and GEANT4 codes are described. A strategy for monitoring backgrounds during data taking is briefly discussed.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. J. Diamond

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 reloadmore » 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.« less

Technical assumption for Mo-99 production in the MARIA reactor. Feasibility study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jaroszewicz, J.; Pytel, K.; Dabkowski, L.

2008-07-15

The main objective of U-235 irradiation is to obtain the Tc-99m isotope which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short life time, is a reaction of radioactive decay of Mo-99 into Tc- 99m. One of the possible sources of molybdenum can be achieved in course of the U-235 fission reaction. The paper presents activities and the calculations results obtained upon the feasibility study on irradiation of U-235 targets for production of molybdenum in the MARIA reactor. The activities including technical assumption were focused on performing calculation for modelling ofmore » the target and irradiation device as well as adequate equipment and tools for processing in reactor. It has been assumed that the basic component of fuel charge is an aluminium cladded plate with dimensions of 40x230x1.45 containing 4.7 g U-235. The presumed mode of the heat removal generated in the fuel charge of the reactor primary cooling circuit influences the construction of installation to be used for irradiation and the technological instrumentation. The outer channel construction for irradiation has to be identical as the standard fuel channel construction of the MARIA reactor. It enables to use the existing slab and reactor mounting sockets for the fastening of the molybdenum channel as well as the cooling water delivery system. The measurement of water temperature cooling a fuel charge and control of water flow rate in the channel can also be carried out be means of the standard instrumentation of the reactor. (author)« less

KSC-02pd1410

NASA Image and Video Library

2002-09-30

KENNEDY SPACE CENTER, FLA. - Jack Wilcox (at the microphone) answers questions from the media during a reunion with his rescuers at the Hangar AF docks, Cape Canaveral Air Force Station. At right is his wife, Patty. Wilcox reunited with the men aboard KSC's Freedom Star SRB retrieval ship that was in the vicinity when Wilcox suffered decompression sickness on a diving expedition 20 miles off shore in the Atlantic Ocean. When the Freedom Star team heard the call for help, they asked the Coast Guard if they could assist. The ship was out on a crane certification exercise and coincidentally had a diver medical technician and other divers training on the crane. The ship's divers were trained for the hyperbaric chamber on board. Upon reaching the Army dock, KSC Occupational Health physician Skip Beeler entered the chamber and continued the process of helping to stabilize Wilcox. After several hours in the chamber, Wilcox, who lives in Orlando, was airlifted to Florida Hospital Orlando.

KSC-02pd1409

NASA Image and Video Library

2002-09-30

KENNEDY SPACE CENTER, FLA. - KENNEDY SPACE CENTER, FLA. - Jack Wilcox answers questions from the media during a reunion at the Hangar AF docks, Cape Canaveral Air Force Station, with his rescuers. At right is his wife, Patty. Wilcox reunited with the men aboard KSC's Freedom Star SRB retrieval ship that was in the vicinity when Wilcox suffered decompression sickness on a diving expedition 20 miles off shore in the Atlantic Ocean. When the Freedom Star team heard the call for help, they asked the Coast Guard if they could assist. The ship was out on a crane certification exercise and coincidentally had a diver medical technician and other divers training on the crane. The ship's divers were trained for the hyperbaric chamber on board. Upon reaching the Army dock, KSC Occupational Health physician Skip Beeler entered the chamber and continued the process of helping to stabilize Wilcox. After several hours in the chamber, Wilcox, who lives in Orlando, was airlifted to Florida Hospital Orlando.

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.

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.

Control of reactor coolant flow path during reactor decay heat removal

DOEpatents

Hunsbedt, Anstein N.

1988-01-01

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

Assessing pretreatment reactor scaling through empirical analysis

DOE PAGES

Lischeske, James J.; Crawford, Nathan C.; Kuhn, Erik; ...

2016-10-10

Pretreatment is a critical step in the biochemical conversion of lignocellulosic biomass to fuels and chemicals. Due to the complexity of the physicochemical transformations involved, predictively scaling up technology from bench- to pilot-scale is difficult. This study examines how pretreatment effectiveness under nominally similar reaction conditions is influenced by pretreatment reactor design and scale using four different pretreatment reaction systems ranging from a 3 g batch reactor to a 10 dry-ton/d continuous reactor. The reactor systems examined were an Automated Solvent Extractor (ASE), Steam Explosion Reactor (SER), ZipperClave(R) reactor (ZCR), and Large Continuous Horizontal-Screw Reactor (LHR). To our knowledge, thismore » is the first such study performed on pretreatment reactors across a range of reaction conditions (time and temperature) and at different reactor scales. The comparative pretreatment performance results obtained for each reactor system were used to develop response surface models for total xylose yield after pretreatment and total sugar yield after pretreatment followed by enzymatic hydrolysis. Near- and very-near-optimal regions were defined as the set of conditions that the model identified as producing yields within one and two standard deviations of the optimum yield. Optimal conditions identified in the smallest-scale system (the ASE) were within the near-optimal region of the largest scale reactor system evaluated. A reaction severity factor modeling approach was shown to inadequately describe the optimal conditions in the ASE, incorrectly identifying a large set of sub-optimal conditions (as defined by the RSM) as optimal. The maximum total sugar yields for the ASE and LHR were 95%, while 89% was the optimum observed in the ZipperClave. The optimum condition identified using the automated and less costly to operate ASE system was within the very-near-optimal space for the total xylose yield of both the ZCR and the LHR, and

Assessing pretreatment reactor scaling through empirical analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lischeske, James J.; Crawford, Nathan C.; Kuhn, Erik

Pretreatment is a critical step in the biochemical conversion of lignocellulosic biomass to fuels and chemicals. Due to the complexity of the physicochemical transformations involved, predictively scaling up technology from bench- to pilot-scale is difficult. This study examines how pretreatment effectiveness under nominally similar reaction conditions is influenced by pretreatment reactor design and scale using four different pretreatment reaction systems ranging from a 3 g batch reactor to a 10 dry-ton/d continuous reactor. The reactor systems examined were an Automated Solvent Extractor (ASE), Steam Explosion Reactor (SER), ZipperClave(R) reactor (ZCR), and Large Continuous Horizontal-Screw Reactor (LHR). To our knowledge, thismore » is the first such study performed on pretreatment reactors across a range of reaction conditions (time and temperature) and at different reactor scales. The comparative pretreatment performance results obtained for each reactor system were used to develop response surface models for total xylose yield after pretreatment and total sugar yield after pretreatment followed by enzymatic hydrolysis. Near- and very-near-optimal regions were defined as the set of conditions that the model identified as producing yields within one and two standard deviations of the optimum yield. Optimal conditions identified in the smallest-scale system (the ASE) were within the near-optimal region of the largest scale reactor system evaluated. A reaction severity factor modeling approach was shown to inadequately describe the optimal conditions in the ASE, incorrectly identifying a large set of sub-optimal conditions (as defined by the RSM) as optimal. The maximum total sugar yields for the ASE and LHR were 95%, while 89% was the optimum observed in the ZipperClave. The optimum condition identified using the automated and less costly to operate ASE system was within the very-near-optimal space for the total xylose yield of both the ZCR and the LHR, and

A Reload and Startup Plan for and #8233;Conversion of the NIST Research Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Diamond, D. J.; Varuttamaseni, A.

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 portionmore » 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.« less

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.

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…

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.

Development of ASTM Standard for SiC-SiC Joint Testing Final Scientific/Technical Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jacobsen, George; Back, Christina

2015-10-30

As the nuclear industry moves to advanced ceramic based materials for cladding and core structural materials for a variety of advanced reactors, new standards and test methods are required for material development and licensing purposes. For example, General Atomics (GA) is actively developing silicon carbide (SiC) based composite cladding (SiC-SiC) for its Energy Multiplier Module (EM2), a high efficiency gas cooled fast reactor. Through DOE funding via the advanced reactor concept program, GA developed a new test method for the nominal joint strength of an endplug sealed to advanced ceramic tubes, Fig. 1-1, at ambient and elevated temperatures called themore » endplug pushout (EPPO) test. This test utilizes widely available universal mechanical testers coupled with clam shell heaters, and specimen size is relatively small, making it a viable post irradiation test method. The culmination of this effort was a draft of an ASTM test standard that will be submitted for approval to the ASTM C28 ceramic committee. Once the standard has been vetted by the ceramics test community, an industry wide standard methodology to test joined tubular ceramic components will be available for the entire nuclear materials community.« less

Performance and emissions of a catalytic reactor with propane, diesel, and Jet A fuels

NASA Technical Reports Server (NTRS)

Anderson, D. N.

1977-01-01

Tests were made to determine the performance and emissions of a catalytic reactor operated with propane, No. 2 diesel, and Jet A fuels. A 12-cm diameter and 16-cm long catalytic reactor using a proprietary noble metal catalyst was operated at an inlet temperature of 800 K, a pressure of 300,000 Pa and reference velocities of 10 to 15 m/s. No significant differences between the performance of the three fuels were observed when 98.5 percent purity propane was used. The combustion efficiency for 99.8-percent purity propane tested later was significantly lower, however. The diesel fuel contained 135 ppm of bound nitrogen and consequently produced the highest NOx emissions of the three fuels. As much as 85 percent of the bound nitrogen was converted to NOx. Steady-state emissions goals based on half the most stringent proposed automotive standards were met when the reactor was operated at an adiabatic combustion temperature higher than 1350 K with all fuels except the 99.8-percent purity propane. With that fuel, a minimum temperature of 1480 K was required.

Nuclear Reactor Physics

NASA Astrophysics Data System (ADS)

Stacey, Weston M.

2001-02-01

An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

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)

KSC-02pd1405

NASA Image and Video Library

2002-09-30

KENNEDY SPACE CENTER, FLA. -- Jack Wilcox, right, returns to the ship that helped rescue him at sea Sept. 11. With him is his wife, Patty. Wilcox reunited with the men aboard KSC's Freedom Star SRB retrieval ship that was in the vicinity when Wilcox suffered decompression sickness on a diving expedition 20 miles off shore in the Atlantic Ocean. When the Freedom Star team heard the call for help, they asked the Coast Guard if they could assist. The ship was out on a crane certification exercise and coincidentally had a diver medical technician and other divers training on the crane. The ship's divers were trained for the hyperbaric chamber on board. Upon reaching the Army dock, KSC Occupational Health physician Skip Beeler entered the chamber and continued the process of helping to stabilize Wilcox. After several hours in the chamber, Wilcox, who lives in Orlando, was airlifted to Florida Hospital Orlando. The reunion was held at the Hangar AF docks on the Cape Canaveral Air Force Station.

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

2014-01-01

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).

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

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0010] Knowledge and Abilities Catalog for Nuclear Power... comment a draft NUREG, NUREG-2104, Revision 0, ``Knowledge and Abilities Catalog for Nuclear Power Plant... developed using this Catalog along with the Operator Licensing Examination Standards for Power Reactors...

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.

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.

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.

Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Diamond, D. J.; Baek, J. S.; Hanson, A. L.

The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in anmore » aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.« less

Esther Wilcox | NREL

Science.gov Websites

Research Associate, University of Colorado, 2004-2007 Patents "Methods for improving syngas-to-alcohol catalyst activity and selectivity," U.S. Patent No. 8,318,986 (2012) "Methods for improving

Nuclear Reactors and Technology

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 inmore » 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.« less

A neutronics feasibility study for the LEU conversion of Poland's Maria research reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bretscher, M. M.

1998-10-14

The MARIA reactor is a high-flux multipurpose research reactor which is water-cooled and moderated with both beryllium and water. Standard HEU (80% {sup 235}U)fuel assemblies consist of six concentric fuel tubes of a U-Al alloy clad in aluminum. Although the inventory of HEU (80%) fuel is nearly exhausted, a supply of highly-loaded 36%-enriched fuel assemblies is available at the reactor site. Neutronic equilibrium studies have been made to determine the relative performance of fuels with enrichments of 80%, 36% and 19.7%. These studies indicate that LEU (19.7%) densities of about 2.5 gU/cm{sup 3} and 3.8 gU/cm{sup 3} are required tomore » match the performance of the MARIA reactor with 80%-enriched and with 36%-enriched fuels, respectively.« less

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-30

... the Standard Review Plan (SRP), concerning the physical security reviews of design certification... NRC staff with the physical security review of applications for design certifications, incorporate... NUCLEAR REGULATORY COMMISSION [NRC-2013-0225] Proposed Revision to Physical Security--Standard...

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.

Metal halides vapor lasers with inner reactor and small active volume.

NASA Astrophysics Data System (ADS)

Shiyanov, D. V.; Sukhanov, V. B.; Evtushenko, G. S.

2018-04-01

Investigation of the energy characteristics of copper, manganese, lead halide vapor lasers with inner reactor and small active volume 90 cm3 was made. The optimal operating pulse repetition rates, temperatures, and buffer gas pressure for gas discharge tubes with internal and external electrodes are determined. Under identical pump conditions, such systems are not inferior in their characteristics to standard metal halide vapor lasers. It is shown that the use of a zeolite halogen generator provides lifetime laser operation.

Spinning fluids reactor

DOEpatents

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

2012-11-20

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

NEUTRONIC REACTOR SYSTEM

DOEpatents

Treshow, M.

1959-02-10

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

Nuclear reactor neutron shielding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Speaker, Daniel P; Neeley, Gary W; Inman, James B

A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactormore » cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.« less

Extrapolating Solar Dynamo Models Throughout the Heliosphere

NASA Astrophysics Data System (ADS)

Cox, B. T.; Miesch, M. S.; Augustson, K.; Featherstone, N. A.

2014-12-01

There are multiple theories that aim to explain the behavior of the solar dynamo, and their associated models have been fiercely contested. The two prevailing theories investigated in this project are the Convective Dynamo model that arises from the pure solving of the magnetohydrodynamic equations, as well as the Babcock-Leighton model that relies on sunspot dissipation and reconnection. Recently, the supercomputer simulations CASH and BASH have formed models of the behavior of the Convective and Babcock-Leighton models, respectively, in the convective zone of the sun. These models show the behavior of the models within the sun, while much less is known about the effects these models may have further away from the solar surface. The goal of this work is to investigate any fundamental differences between the Convective and Babcock-Leighton models of the solar dynamo outside of the sun and extending into the solar system via the use of potential field source surface extrapolations implemented via python code that operates on data from CASH and BASH. The use of real solar data to visualize supergranular flow data in the BASH model is also used to learn more about the behavior of the Babcock-Leighton Dynamo. From the process of these extrapolations it has been determined that the Babcock-Leighton model, as represented by BASH, maintains complex magnetic fields much further into the heliosphere before reverting into a basic dipole field, providing 3D visualisations of the models distant from the sun.

Analysis of supercritical CO{sub 2} cycle control strategies and dynamic response for Generation IV Reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moisseytsev, A.; Sienicki, J. J.

2011-04-12

The analysis of specific control strategies and dynamic behavior of the supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle has been extended to the two reactor types selected for continued development under the Generation IV Nuclear Energy Systems Initiative; namely, the Very High Temperature Reactor (VHTR) and the Sodium-Cooled Fast Reactor (SFR). Direct application of the standard S-CO{sub 2} recompression cycle to the VHTR was found to be challenging because of the mismatch in the temperature drop of the He gaseous reactor coolant through the He-to-CO{sub 2} reactor heat exchanger (RHX) versus the temperature rise of the CO{sub 2} through themore » RHX. The reference VHTR features a large temperature drop of 450 C between the assumed core outlet and inlet temperatures of 850 and 400 C, respectively. This large temperature difference is an essential feature of the VHTR enabling a lower He flow rate reducing the required core velocities and pressure drop. In contrast, the standard recompression S-CO{sub 2} cycle wants to operate with a temperature rise through the RHX of about 150 C reflecting the temperature drop as the CO{sub 2} expands from 20 MPa to 7.4 MPa in the turbine and the fact that the cycle is highly recuperated such that the CO{sub 2} entering the RHX is effectively preheated. Because of this mismatch, direct application of the standard recompression cycle results in a relatively poor cycle efficiency of 44.9%. However, two approaches have been identified by which the S-CO{sub 2} cycle can be successfully adapted to the VHTR and the benefits of the S-CO{sub 2} cycle, especially a significant gain in cycle efficiency, can be realized. The first approach involves the use of three separate cascaded S-CO{sub 2} cycles. Each S-CO{sub 2} cycle is coupled to the VHTR through its own He-to-CO{sub 2} RHX in which the He temperature is reduced by 150 C. The three respective cycles have efficiencies of 54, 50, and 44%, respectively, resulting in a net

Thorium fueled reactor

NASA Astrophysics Data System (ADS)

Sipaun, S.

2017-01-01

Current development in thorium fueled reactors shows that they can be designed to operate in the fast or thermal spectrum. The thorium/uranium fuel cycle converts fertile thorium-232 into fissile uranium-233, which fissions and releases energy. This paper analyses the characteristics of thorium fueled reactors and discusses the thermal reactor option. It is found that thorium fuel can be utilized in molten salt reactors through many configurations and designs. A balanced assessment on the feasibility of adopting one reactor technology versus another could lead to optimized benefits of having thorium resource.

Assessment and mitigation of power quality problems for PUSPATI TRIGA Reactor (RTP)

NASA Astrophysics Data System (ADS)

Zakaria, Mohd Fazli; Ramachandaramurthy, Vigna K.

2017-01-01

An electrical power systems are exposed to different types of power quality disturbances. Investigation and monitoring of power quality are necessary to maintain accurate operation of sensitive equipment especially for nuclear installations. This paper will discuss the power quality problems observed at the electrical sources of PUSPATI TRIGA Reactor (RTP). Assessment of power quality requires the identification of any anomalous behavior on a power system, which adversely affects the normal operation of electrical or electronic equipment. A power quality assessment involves gathering data resources; analyzing the data (with reference to power quality standards) then, if problems exist, recommendation of mitigation techniques must be considered. Field power quality data is collected by power quality recorder and analyzed with reference to power quality standards. Normally the electrical power is supplied to the RTP via two sources in order to keep a good reliability where each of them is designed to carry the full load. The assessment of power quality during reactor operation was performed for both electrical sources. There were several disturbances such as voltage harmonics and flicker that exceeded the thresholds. To reduce these disturbances, mitigation techniques have been proposed, such as to install passive harmonic filters to reduce harmonic distortion, dynamic voltage restorer (DVR) to reduce voltage disturbances and isolate all sensitive and critical loads.

An approach to model reactor core nodalization for deterministic safety analysis

NASA Astrophysics Data System (ADS)

Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

2016-01-01

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

An approach to model reactor core nodalization for deterministic safety analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my; Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to bemore » employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.« less

Removal of natural organic matter and arsenic from water by electrocoagulation/flotation continuous flow reactor.

PubMed

Mohora, Emilijan; Rončević, Srdjan; Dalmacija, Božo; Agbaba, Jasmina; Watson, Malcolm; Karlović, Elvira; Dalmacija, Milena

2012-10-15

The performance of the laboratory scale electrocoagulation/flotation (ECF) reactor in removing high concentrations of natural organic matter (NOM) and arsenic from groundwater was analyzed in this study. An ECF reactor with bipolar plate aluminum electrodes was operated in the horizontal continuous flow mode. Electrochemical and flow variables were optimized to examine ECF reactor contaminants removal efficiency. The optimum conditions for the process were identified as groundwater initial pH 5, flow rate=4.3 l/h, inter electrode distance=2.8 cm, current density=5.78 mA/cm(2), A/V ratio=0.248 cm(-1). The NOM removal according to UV(254) absorbance and dissolved organic matter (DOC) reached highest values of 77% and 71% respectively, relative to the raw groundwater. Arsenic removal was 85% (6.2 μg As/l) relative to raw groundwater, satisfying the drinking water standards. The specific reactor electrical energy consumption was 17.5 kWh/kg Al. The specific aluminum electrode consumption was 66 g Al/m(3). According to the obtained results, ECF in horizontal continuous flow mode is an energy efficient process to remove NOM and arsenic from groundwater. Copyright © 2012 Elsevier B.V. All rights reserved.

Thermos reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Labrousse, M.; Lerouge, B.; Dupuy, G.

1978-04-01

THERMOS is a water reactor designed to provide hot water up to 120/sup 0/C for district heating or for desalination applications. It is a 100-MW reactor based on proven technology: oxide fuel plate elements, integrated primary circuit, and reactor vessel located in the bottom of a pool. As in swimming pool reactors, the pool is used for biological shielding, emergency core cooling, and fission product filtering (in case of an accident). Before economics, safety is the main characteristic of the concept: no fuel failure admitted, core under water in any accidental configuration, inspection of every ''nuclear'' component, and double-wall containment.

COMMERCIALIZATION OF AN ATMOSPHERIC IRON-BASED CDCL PROCESS FOR POWER PRODUCTION. PHASE I: TECHNOECONOMIC ANALYSIS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vargas, Luis

Coal Direct Chemical Looping (CDCL) is an advanced oxy-combustion technology that has potential to enable substantial reductions in the cost and energy penalty associated with carbon dioxide (CO2) capture from coal-fired power plants. Through collaborative efforts, the Babcock & Wilcox Power Generation Group (B&W) and The Ohio State University (OSU) developed a conceptual design for a 550 MWe (net) supercritical CDCL power plant with greater than 90% CO2 capture and compression. Process simulations were completed to enable an initial assessment of its technical performance. A cost estimate was developed following DOE’s guidelines as outlined in NETL’s report “Quality Guidelines formore » Energy System Studies: Cost Estimation Methodology for NETL Assessments of Power Plant Performance”, (2011/1455). The cost of electricity for the CDCL plant without CO2 Transportation and Storage cost resulted in $ $102.67 per MWh, which corresponds to a 26.8 % increase in cost of electricity (COE) when compared to an air-fired pulverized-coal supercritical power plant. The cost of electricity is strongly depending on the total plant cost and cost of the oxygen carrier particles. The CDCL process could capture further potential savings by increasing the performance of the particles and reducing the plant size. During the techno-economic analysis, the team identified technology and engineering gaps that need to be closed to bring the technology to commercialization. The technology gaps were focused in five critical areas: (i) moving bed reducer reactor, (ii) fluidized bed combustor, (iii) particle riser, (iv) oxygen-carrier particle properties, and (v) process operation. The key technology gaps are related to particle performance, particle manufacturing cost, and the operation of the reducer reactor. These technology gaps are to be addressed during Phase II of project. The project team is proposing additional lab testing to be completed on the particle and a 3MWth pilot

Fuel Fabrication and Nuclear Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Karpius, Peter Joseph

2017-02-02

The uranium from the enrichment plant is still in the form of UF 6. UF 6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF 6 is converted into UO 2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

Thermochemical reactor systems and methods

DOEpatents

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.

CADMIUM-RARE EARTH BORATE GLASS AS REACTOR CONTROL MATERIAL

DOEpatents

Ploetz, G.L.; Ray, W.E.

1958-11-01

A reactor control rod fabricated from a cadmiumrare earth-borate glass is presented. The rare earth component of this glass is selected from among those rare earths having large neutron capture cross sections, such as samarium, gadolinium or europium. Partlcles of this glass are then dispersed in a metal matrix by standard powder metallurgy techniques.

FORCE2: A state-of-the-art two-phase code for hydrodynamic calculations

NASA Astrophysics Data System (ADS)

Ding, Jianmin; Lyczkowski, R. W.; Burge, S. W.

1993-02-01

A three-dimensional computer code for two-phase flow named FORCE2 has been developed by Babcock and Wilcox (B & W) in close collaboration with Argonne National Laboratory (ANL). FORCE2 is capable of both transient as well as steady-state simulations. This Cartesian coordinates computer program is a finite control volume, industrial grade and quality embodiment of the pilot-scale FLUFIX/MOD2 code and contains features such as three-dimensional blockages, volume and surface porosities to account for various obstructions in the flow field, and distributed resistance modeling to account for pressure drops caused by baffles, distributor plates and large tube banks. Recently computed results demonstrated the significance of and necessity for three-dimensional models of hydrodynamics and erosion. This paper describes the process whereby ANL's pilot-scale FLUFIX/MOD2 models and numerics were implemented into FORCE2. A description of the quality control to assess the accuracy of the new code and the validation using some of the measured data from Illinois Institute of Technology (UT) and the University of Illinois at Urbana-Champaign (UIUC) are given. It is envisioned that one day, FORCE2 with additional modules such as radiation heat transfer, combustion kinetics and multi-solids together with user-friendly pre- and post-processor software and tailored for massively parallel multiprocessor shared memory computational platforms will be used by industry and researchers to assist in reducing and/or eliminating the environmental and economic barriers which limit full consideration of coal, shale and biomass as energy sources, to retain energy security, and to remediate waste and ecological problems.

Low NO{sub x} burner modifications to front-fired pulverized coal boilers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Broderick, R.G.; Wagner, M.

1998-07-01

Madison Gas and Electric Blount Street Station Units 8 and 9 are Babcock and Wilcox pulverized coal fired and natural gas fired boilers. These boilers were build in the late 1950's and early 1960's with each boiler rated at 425,000 lb./hr of steam producing 50 MW of electricity. The boilers are rated at 9,500 F at 1,350 psig. Each unit is equipped with one Ljungstroem air heater and two B and W EL pulverizers. These units burn subbituminous coal with higher heating value of 10,950 Btu/LB on an as-received basis. The nitrogen content is approximately 1.23% with 15% moisture. Inmore » order to comply with the new Clean Air Act Madison Gas and Electric needs to reduce NO{sub x} on these units to less than .5 LB/mmBtu. Baseline NO{sub x} emissions on these units range between .8--.9 lb./mmBtu. LOIs average approximately 8%. Madison Gas and Electric contracted with RJM Corporation to modify the existing burners to achieve this objective. These modifications consisted of adding patented circumferentially and radially staged flame stabilizers, modifying the coal pipe, and replacing the coal impeller with a circumferentially staged coal spreader. RJM Corporation utilized computational fluid dynamics modeling in order to design the equipment to modify these burners. The equipment was installed during the March 1997 outage and start-up and optimization was conducted in April 1997. Final performance results and economic data will be included in the final paper.« less

Long Distance Reactor Antineutrino Flux Monitoring

NASA Astrophysics Data System (ADS)

Dazeley, Steven; Bergevin, Marc; Bernstein, Adam

2015-10-01

The feasibility of antineutrino detection as an unambiguous and unshieldable way to detect the presence of distant nuclear reactors has been studied. While KamLAND provided a proof of concept for long distance antineutrino detection, the feasibility of detecting single reactors at distances greater than 100 km has not yet been established. Even larger detectors than KamLAND would be required for such a project. Considerations such as light attenuation, environmental impact and cost, which favor water as a detection medium, become more important as detectors get larger. We have studied both the sensitivity of water based detection media as a monitoring tool, and the scientific impact such detectors might provide. A next generation water based detector may be able to contribute to important questions in neutrino physics, such as supernova neutrinos, sterile neutrino oscillations, and non standard electroweak interactions (using a nearby compact accelerator source), while also providing a highly sensitive, and inherently unshieldable reactor monitoring tool to the non proliferation community. In this talk I will present the predicted performance of an experimental non proliferation and high-energy physics program. Lawrence Livermore National Laboratory is operated by Lawrence Livermore National Security, LLC, for the U.S. Department of Energy, National Nuclear Security Administration under Contract DE-AC52-07NA27344. Release number LLNL-ABS-674192.

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. Copyright © 2015 Elsevier Ltd. All rights reserved.

REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

2011-07-18

Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Manymore » research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and

Once-through integral system (OTIS): Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gloudemans, J R

1986-09-01

A scaled experimental facility, designated the once-through integral system (OTIS), was used to acquire post-small break loss-of-coolant accident (SBLOCA) data for benchmarking system codes. OTIS was also used to investigate the application of the Abnormal Transient Operating Guidelines (ATOG) used in the Babcock and Wilcox (B and W) designed nuclear steam supply system (NSSS) during the course of an SBLOCA. OTIS was a single-loop facility with a plant to model power scale factor of 1686. OTIS maintained the key elevations, approximate component volumes, and loop flow resistances, and simulated the major component phenomena of a B and W raised-loop nuclearmore » plant. A test matrix consisting of 15 tests divided into four categories was performed. The largest group contained 10 tests and was defined to parametrically obtain an extensive set of plant-typical experimental data for code benchmarking. Parameters such as leak size, leak location, and high-pressure injection (HPI) shut-off head were individually varied. The remaining categories were specified to study the impact of the ATOGs (2 tests), to note the effect of guard heater operation on observed phenomena (2 tests), and to provide a data set for comparison with previous test experience (1 test). A summary of the test results and a detailed discussion of Test 220100 is presented. Test 220100 was the nominal or reference test for the parametric studies. This test was performed with a scaled 10-cm/sup 2/ leak located in the cold leg suction piping.« less

TREAT Reactor Control and Protection System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.

1985-01-01

The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS).more » The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab.« less

NUCLEAR REACTOR

DOEpatents

Grebe, J.J.

1959-07-14

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

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 D.; 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

Treatment of low-strength soluble wastewater using an anaerobic baffled reactor (ABR).

PubMed

Gopala Krishna, G V T; Kumar, Pramod; Kumar, Pradeep

2009-01-01

Treatment of low-strength soluble wastewater (COD approximately 500 mg/L) was studied using an eight chambered anaerobic baffled reactor (ABR). At pseudo steady-state (PSS), the average total and soluble COD values (COD(T) and COD(S)) at 8h hydraulic retention time (HRT) were found to be around 50 and 40 mg/L, respectively, while at 10h HRT average COD(T) and COD(S) values were of the order of 47 and 37 mg/L, respectively. COD and BOD (3 day, 27 degrees C) removal averaged more than 90%. Effluent conformed to Indian standards laid down for BOD (less than 30 mg/L). Reactor effluent characteristics exhibited very low values of standard deviation indicating excellent reactor stability at PSS in terms of effluent characteristics. Based on mass balance calculations, more than 60% of raw wastewater COD was estimated to be recovered as CH(4) in the gas phase. Compartment-wise profiles indicated that most of the BOD and COD got reduced in the initial compartments only. Sudden drop in pH (7.8-6.7) and formation of volatile fatty acids (VFA) (53-85 mg/L) were observed in the first compartment due to acidogenesis and acetogenesis. The pH increased and VFA concentration decreased longitudinally down the reactor. Residence time distribution (RTD) studies revealed that the flow pattern in the ABR was neither completely plug-flow nor perfectly mixed. Observations from scanning electron micrographs (SEM) suggest that distinct phase separation takes place in an ABR.

Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

NASA Astrophysics Data System (ADS)

Aji, Indarta Kuncoro; Waris, Abdul; Permana, Sidik

2015-09-01

Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF2-ThF4-233UF4 respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.

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.

Study of dietary supplements compositions by neutron activation analysis at the VR-1 training reactor

NASA Astrophysics Data System (ADS)

Stefanik, Milan; Rataj, Jan; Huml, Ondrej; Sklenka, Lubomir

2017-11-01

The VR-1 training reactor operated by the Czech Technical University in Prague is utilized mainly for education of students and training of various reactor staff; however, R&D is also carried out at the reactor. The experimental instrumentation of the reactor can be used for the irradiation experiments and neutron activation analysis. In this paper, the neutron activation analysis (NAA) is used for a study of dietary supplements containing the zinc (one of the essential trace elements for the human body). This analysis includes the dietary supplement pills of different brands; each brand is represented by several different batches of pills. All pills were irradiated together with the standard activation etalons in the vertical channel of the VR-1 reactor at the nominal power (80 W). Activated samples were investigated by the nuclear gamma-ray spectrometry technique employing the semiconductor HPGe detector. From resulting saturated activities, the amount of mineral element (Zn) in the pills was determined using the comparative NAA method. The results show clearly that the VR-1 training reactor is utilizable for neutron activation analysis experiments.

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)

Characteristics and Dose Levels for Spent Reactor Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coates, Cameron W

2007-01-01

Current guidance considers highly radioactive special nuclear materials to be those materials that, unshielded, emit a radiation dose [rate] measured at 1 m which exceeds 100 rem/h. Smaller, less massive fuel assemblies from research reactors can present a challenge from the point of view of self protection because of their size (lower dose, easier to handle) and the desirability of higher enrichments; however, a follow-on study to cross-compare dose trends of research reactors and power reactors was deemed useful to confirm/verify these trends. This paper summarizes the characteristics and dose levels of spent reactor fuels for both research reactors andmore » power reactors and extends previous studies aimed at quantifying expected dose rates from research reactor fuels worldwide.« less

Reactor safety method

DOEpatents

Vachon, Lawrence J.

1980-03-11

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

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.

Secondary Organic Aerosol Production from Gasoline Vehicle Exhaust: Effects of Engine Technology, Cold Start, and Emission Certification Standard.

PubMed

Zhao, Yunliang; Lambe, Andrew T; Saleh, Rawad; Saliba, Georges; Robinson, Allen L

2018-02-06

Secondary organic aerosol (SOA) formation from dilute exhaust from 16 gasoline vehicles was investigated using a potential aerosol mass (PAM) oxidation flow reactor during chassis dynamometer testing using the cold-start unified cycle (UC). Ten vehicles were equipped with gasoline direct injection engines (GDI vehicles) and six with port fuel injection engines (PFI vehicles) certified to a wide range of emissions standards. We measured similar SOA production from GDI and PFI vehicles certified to the same emissions standard; less SOA production from vehicles certified to stricter emissions standards; and, after accounting for differences in gas-particle partitioning, similar effective SOA yields across different engine technologies and certification standards. Therefore the ongoing, dramatic shift from PFI to GDI vehicles in the United States should not alter the contribution of gasoline vehicles to ambient SOA and the natural replacement of older vehicles with newer ones certified to stricter emissions standards should reduce atmospheric SOA levels. Compared to hot operations, cold-start exhaust had lower effective SOA yields, but still contributed more SOA overall because of substantially higher organic gas emissions. We demonstrate that the PAM reactor can be used as a screening tool for vehicle SOA production by carefully accounting for the effects of the large variations in emission rates.

Reactor water cleanup system

DOEpatents

Gluntz, Douglas M.; Taft, William E.

1994-01-01

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

Reactor water cleanup system

DOEpatents

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

1994-12-20

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

NUCLEAR REACTOR

DOEpatents

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

1958-01-21

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

The S-Matrix and Acoustic Signal Structure in Simple and Compound Waveguides.

DTIC Science & Technology

1982-12-01

RD-A125 583 THE S-MATRIX AND ACOUSTIC SIGNAL STRUCTURE IN SIMPLE- L/1 AND COMPOUND WAVEGUIDES(U) UTAH UNIV SALT LAKE CITY DEPT OF MATHEMATICS C H...WILCOX DEC 82 TSR-45 UNCLASSIFIED N6@8i4-76-C-8276 F/G 12/1 NL IEINEIIIIIIEIhllhlllllllIflllllflflflflflEN L-- U5-12 III,2,0 III.J --IL.,5 MICROCOP ...RESLUIO TETCHRNATIONA BUREA OF 20NADS16 THE S-MATRIX AND ACOUSTIC SIGNAL STRUCTURE IN SIMPLE AND COMPOUND WAVEGUIDES C. H. Wilcox Technical Simmary Report

Pressurized fluidized bed reactor

DOEpatents

Isaksson, J.

1996-03-19

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

Pressurized fluidized bed reactor

DOEpatents

Isaksson, Juhani

1996-01-01

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

NEUTRONIC REACTOR

DOEpatents

Daniels, F.

1959-10-27

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

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.

NEUTRONIC REACTOR SHIELDING

DOEpatents

Borst, L.B.

1961-07-11

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

Coupled hydrology and biogeochemistry of Paleocene–Eocene coal beds, northern Gulf of Mexico

USGS Publications Warehouse

McIntosh, Jennifer C.; Warwick, Peter D.; Martini, Anna M.; Osborn, Stephen G.

2010-01-01

Thirty-six formation waters, gas, and microbial samples were collected and analyzed from natural gas and oil wells producing from the Paleocene to Eocene Wilcox Group coal beds and adjacent sandstones in north-central Louisiana, USA, to investigate the role hydrology plays on the generation and distribution of microbial methane. Major ion chemistry and Cl−Br relations of Wilcox Group formation waters suggest mixing of freshwater with halite-derived brines. High alkalinities (up to 47.8 meq/L), no detectable SO4, and elevated δ13C values of dissolved inorganic carbon (up to 20.5‰ Vienna Peedee belemnite [VPDB]) and CO2 (up to 17.67‰ VPDB) in the Wilcox Group coals and adjacent sandstones indicate the dominance of microbial methanogenesis. The δ13C and δD values of CH4, and carbon isotope fractionation of CO2 and CH4, suggest CO2 reduction is the major methanogenic pathway. Geochemical indicators for methanogenesis drop off significantly at chloride concentrations above ∼1.7 mol/L, suggesting that high salinities inhibit microbial activity at depths greater than ∼1.6 km. Formation waters in the Wilcox Group contain up to 1.6% modern carbon (A14C) to at least 1690 m depth; the covariance of δD values of co-produced H2O and CH4 indicate that the microbial methane was generated in situ with these Late Pleistocene or younger waters. The most enriched carbon isotope values for dissolved inorganic carbon (DIC) and CO2, and highest alkalinities, were detected in Wilcox Group sandstone reservoirs that were CO2 flooded in the 1980s for enhanced oil recovery, leading to the intriguing hypothesis that CO2 sequestration may actually enhance methanogenesis in organic-rich formations.

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.

Neutrino scattering and the reactor antineutrino anomaly

NASA Astrophysics Data System (ADS)

Garcés, Estela; Cañas, Blanca; Miranda, Omar; Parada, Alexander

2017-12-01

Low energy threshold reactor experiments have the potential to give insight into the light sterile neutrino signal provided by the reactor antineutrino anomaly and the gallium anomaly. In this work we analyze short baseline reactor experiments that detect by elastic neutrino electron scattering in the context of a light sterile neutrino signal. We also analyze the sensitivity of experimental proposals of coherent elastic neutrino nucleus scattering (CENNS) detectors in order to exclude or confirm the sterile neutrino signal with reactor antineutrinos.

97. ARAIII. ML1 reactor has been moved into GCRE reactor ...

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

97. ARA-III. ML-1 reactor has been moved into GCRE reactor building (ARA-608) for examination of corrosion on its underside and repair. May 24, 1963. Ineel photo no. 63-3485. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Neutron fluxes in test reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Youinou, Gilles Jean-Michel

Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Clarke, Kester Diederik

The intent of this report is to document a procedure used at LANL for HIP bonding aluminum cladding to U-10Mo fuel foils using a formed HIP can for the Domestic Reactor Conversion program in the NNSA Office of Material, Management and Minimization, and provide some details that may not have been published elsewhere. The HIP process is based on the procedures that have been used to develop the formed HIP can process, including the baseline process developed at Idaho National Laboratory (INL). The HIP bonding cladding process development is summarized in the listed references. Further iterations with Babcock & Wilcoxmore » (B&W) to refine the process to meet production and facility requirements is expected.« less

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

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.

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.

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.

NEUTRONIC REACTOR MANIPULATING DEVICE

DOEpatents

Ohlinger, L.A.

1962-08-01

A cable connecting a control rod in a reactor with a motor outside the reactor for moving the rod, and a helical conduit in the reactor wall, through which the cable passes are described. The helical shape of the conduit prevents the escape of certain harmful radiations from the reactor. (AEC)

Summary of Thermocouple Performance During Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor and Out-of-Pile Thermocouple Testing in Support of Such Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

A. J. Palmer; DC Haggard; J. W. Herter

High temperature gas reactor experiments create unique challenges for thermocouple based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B); and tungsten-rhenium thermocouples (Types C and W). For lower temperature applications, previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type Nmore » thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of these Nickel based thermocouples is limited when the temperature exceeds 1000°C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past ten years, three long-term Advanced Gas Reactor (AGR) experiments have been conducted with measured temperatures ranging from 700oC – 1200oC. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out of pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150oC and 1200oC for 2000 hours at each temperature, followed by 200 hours at 1250oC, and 200 hours at 1300oC. The standard Type N design utilizes high purity crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including Haynes 214 alloy sheath, spinel (MgAl2O4) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard fired

ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

E. Blanford; E. Keldrauk; M. Laufer

2010-09-20

Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement,more » and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction

Summary of thermocouple performance during advanced gas reactor fuel irradiation experiments in the advanced test reactor and out-of-pile thermocouple testing in support of such experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Palmer, A. J.; Haggard, DC; Herter, J. W.

High temperature gas reactor experiments create unique challenges for thermocouple-based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time-dependent change in composition and, as a consequence, a time-dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C). For lower temperature applications, previous experiences with Type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly, Type N thermocouples are expected to bemore » only slightly affected by neutron fluence. Currently, the use of these nickel-based thermocouples is limited when the temperature exceeds 1000 deg. C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past 10 years, three long-term Advanced Gas Reactor experiments have been conducted with measured temperatures ranging from 700 deg. C - 1200 deg. C. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out-of-pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150 deg. C and 1200 deg. C for 2,000 hours at each temperature, followed by 200 hours at 1250 deg. C and 200 hours at 1300 deg. C. The standard Type N design utilizes high purity, crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including a Haynes 214 alloy sheath, spinel (MgAl{sub 2}O{sub 4}) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose

CANDU in-reactor quantitative visual-based inspection techniques

NASA Astrophysics Data System (ADS)

Rochefort, P. A.

2009-02-01

sealed by rolling its ends into the rolled joint area. During reactor refurbishment, the original FC calandria tubes are removed, potentially scratching the rolled joint area and, thereby, compromising the seal with the new FC calandria tube. The procedure involves delivering an inspection module having a radiation-resistant camera, standard lighting, and a structured lighting projector. The surface is inspected by rotating the module within the rolled joint area. If a flaw is detected, its depth and width are gauged from the profile variation of the structured lighting in a captured image. As well, the diameter profile of the area is measured from the analysis of a series of captured circumferential images of the structured lighting profiles on the surface.

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.

2016-08-31

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF 2) 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. Themore » 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.« less

F Reactor Inspection

ScienceCinema

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2018-01-16

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

F Reactor Inspection

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-10-29

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

NEUTRONIC REACTOR

DOEpatents

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

1959-03-24

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

Current drive for stability of thermonuclear plasma reactor

NASA Astrophysics Data System (ADS)

Amicucci, L.; Cardinali, A.; Castaldo, C.; Cesario, R.; Galli, A.; Panaccione, L.; Paoletti, F.; Schettini, G.; Spigler, R.; Tuccillo, A.

2016-01-01

To produce in a thermonuclear fusion reactor based on the tokamak concept a sufficiently high fusion gain together stability necessary for operations represent a major challenge, which depends on the capability of driving non-inductive current in the hydrogen plasma. This request should be satisfied by radio-frequency (RF) power suitable for producing the lower hybrid current drive (LHCD) effect, recently demonstrated successfully occurring also at reactor-graded high plasma densities. An LHCD-based tool should be in principle capable of tailoring the plasma current density in the outer radial half of plasma column, where other methods are much less effective, in order to ensure operations in the presence of unpredictably changes of the plasma pressure profiles. In the presence of too high electron temperatures even at the periphery of the plasma column, as envisaged in DEMO reactor, the penetration of the coupled RF power into the plasma core was believed for long time problematic and, only recently, numerical modelling results based on standard plasma wave theory, have shown that this problem should be solved by using suitable parameter of the antenna power spectrum. We show here further information on the new understanding of the RF power deposition profile dependence on antenna parameters, which supports the conclusion that current can be actively driven over a broad layer of the outer radial half of plasma column, thus enabling current profile control necessary for the stability of a reactor.

155. ARAIII Reactor building (ARA608) Details of reactor pit showing ...

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

155. ARA-III Reactor building (ARA-608) Details of reactor pit showing tray supports and fuel element storage rack. Aerojet-general 880-area/GCRE-608-MS-2. Date: November 1958. Ineel index code no. 063-0608-40-013-102625. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

FAST NEUTRON REACTOR

DOEpatents

Soodak, H.; Wigner, E.P.

1961-07-25

A reactor comprising fissionable material in concentration sufficiently high so that the average neutron enengy within the reactor is at least 25,000 ev is described. A natural uranium blanket surrounds the reactor, and a moderating reflector surrounds the blanket. The blanket is thick enough to substantially eliminate flow of neutrons from the reflector.

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.

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)

Removal of anaerobic soluble microbial products in a biological activated carbon reactor.

PubMed

Dong, Xiaojing; Zhou, Weili; He, Shengbing

2013-09-01

The soluble microbial products (SMP) in the biological treatment effluent are generally of great amount and are poorly biodegradable. Focusing on the biodegradation of anaerobic SMP, the biological activated carbon (BAC) was introduced into the anaerobic system. The experiments were conducted in two identical lab-scale up-flow anaerobic sludge blanket (UASB) reactors. The high strength organics were degraded in the first UASB reactor (UASB1) and the second UASB (UASB2, i.e., BAC) functioned as a polishing step to remove SMP produced in UASB1. The results showed that 90% of the SMP could be removed before granular activated carbon was saturated. After the saturation, the SMP removal decreased to 60% on the average. Analysis of granular activated carbon adsorption revealed that the main role of SMP removal in BAC reactor was biodegradation. A strain of SMP-degrading bacteria, which was found highly similar to Klebsiella sp., was isolated, enriched and inoculated back to the BAC reactor. When the influent chemical oxygen demand (COD) was 10,000 mg/L and the organic loading rate achieved 10 kg COD/(m3 x day), the effluent from the BAC reactor could meet the discharge standard without further treatment. Anaerobic BAC reactor inoculated with the isolated Klebsiella was proved to be an effective, cheap and easy technical treatment approach for the removal of SMP in the treatment of easily-degradable wastewater with COD lower than 10,000 mg/L.

Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aji, Indarta Kuncoro, E-mail: indartaaji@s.itb.ac.id; Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Permana, Sidik

Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 datamore » library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.« less

Reactor Operations Monitoring System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hart, M.M.

1989-01-01

The Reactor Operations Monitoring System (ROMS) is a VME based, parallel processor data acquisition and safety action system designed by the Equipment Engineering Section and Reactor Engineering Department of the Savannah River Site. The ROMS will be analyzing over 8 million signal samples per minute. Sixty-eight microprocessors are used in the ROMS in order to achieve a real-time data analysis. The ROMS is composed of multiple computer subsystems. Four redundant computer subsystems monitor 600 temperatures with 2400 thermocouples. Two computer subsystems share the monitoring of 600 reactor coolant flows. Additional computer subsystems are dedicated to monitoring 400 signals from assortedmore » process sensors. Data from these computer subsystems are transferred to two redundant process display computer subsystems which present process information to reactor operators and to reactor control computers. The ROMS is also designed to carry out safety functions based on its analysis of process data. The safety functions include initiating a reactor scram (shutdown), the injection of neutron poison, and the loadshed of selected equipment. A complete development Reactor Operations Monitoring System has been built. It is located in the Program Development Center at the Savannah River Site and is currently being used by the Reactor Engineering Department in software development. The Equipment Engineering Section is designing and fabricating the process interface hardware. Upon proof of hardware and design concept, orders will be placed for the final five systems located in the three reactor areas, the reactor training simulator, and the hardware maintenance center.« less

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. Copyright © 2016 Elsevier Ltd. All rights reserved.

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.

Hybrid adsorptive membrane reactor

DOEpatents

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

2011-03-01

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.

HORIZONTAL BOILING REACTOR SYSTEM

DOEpatents

Treshow, M.

1958-11-18

Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

75 FR 29656 - Amendment of Class E Airspace; Mountain View, AR

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-27

...-1181; Airspace Docket No. 09-ASW-36] Amendment of Class E Airspace; Mountain View, AR AGENCY: Federal... Mountain View, AR. Decommissioning of the Wilcox non-directional beacon (NDB) at Mountain View Wilcox Memorial Field Airport, Mountain View, AR, has made this action necessary to enhance the safety and...

76 FR 2688 - Change in Bank Control Notices; Acquisitions of Shares of a Bank or Bank Holding Company

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-14

... 442, St. Louis, Missouri 63166-2034: 1. The Coffey Family Control Group, which consists of Avery..., Minnesota, and both as members of The Wilcox Family Control Group; to acquire control of Wilcox Bancshares... FEDERAL RESERVE SYSTEM Change in Bank Control Notices; Acquisitions of Shares of a Bank or Bank...

Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bokhari, Ishtiaq H.

2004-12-15

The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried outmore » at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined.« less

Impact of Fission Neutron Energies on Reactor Antineutrino Spectra

NASA Astrophysics Data System (ADS)

Hermanek, Keith; Littlejohn, Bryce; Gustafson, Ian

2017-09-01

Recent measurements of the reactor antineutrino spectra (Double Chooz, Reno, and Daya Bay) have shown a discrepancy in the 5-7 MeV region when compared to current theoretical models (Vogel and Huber-Mueller). There are numerous theories pertaining to this antineutrino anomaly, including theories that point to new physics beyond the standard model. In the paper ``Possible Origins and Implications of the Shoulder in Reactor Neutrino Spectra'' by A. Hayes et al., explanations for this anomaly are suggested. One theory is that there are interactions from fast and epithermal incident neutrons which are significant enough to create more events in the 5-7 MeV by a noticeable amount. In our research, we used the Oklo software network created by Dan Dwyer. This generates ab initio antineutrino and beta decay spectra based on standard fission yield databases ENDF, JENDL, JEFF, and the beta decay transition database ENSDF-6. Utilizing these databases as inputs, we show with reasonable assumptions one can prove contributions of fast and epithermal neutrons is less than 3% in the 5-7 MeV region. We also discovered rare isotopes are present in beta decay chains but not well measured and have no corresponding database information, and studied its effect onto the spectrum.

High-rate treatment of molasses wastewater by combination of an acidification reactor and a USSB reactor.

PubMed

Onodera, Takashi; Sase, Shinya; Choeisai, Pairaya; Yoochatchaval, Wilasinee; Sumino, Haruhiko; Yamaguchi, Takashi; Ebie, Yoshitaka; Xu, Kaiqin; Tomioka, Noriko; Syutsubo, Kazuaki

2011-01-01

A combination of an acidification reactor and an up-flow staged sludge bed (USSB) reactor was applied for treatment of molasses wastewater containing a large amount of organic compounds and sulfate. The USSB reactor had three gas-solid separators (GSS) along the height of the reactor. The combined system was continuously operated at mesophilic temperature over 400 days. In the acidification reactor, acid formation and sulfate reduction were effectively carried out. The sugars contained in the influent wastewater were mostly acidified into acetate, propionate, and n-butyrate. In addition, 10-30% of influent sulfur was removed from the acidification reactor by means of sulfate reduction followed by stripping of hydrogen sulfide. The USSB achieved a high organic loading rate (OLR) of 30 kgCOD m(-3) day(-1) with 82% COD removal. Vigorous biogas production was observed at a rate of 15 Nm(3) biogas m(-3) reactor day(-1). The produced biogas, including hydrogen sulfide, was removed from the wastewater mostly via the GSS. The GSS provided a moderate superficial biogas flux and low sulfide concentration in the sludge bed, resulting in the prevention of sludge washout and sulfide inhibition of methanogens. By advantages of this feature, the USSB may have been responsible for achieving sufficient retention (approximately 60 gVSS L(-1)) of the granular sludge with high methanogenic activity (0.88 gCOD gVSS(-1) day(-1) for acetate and as high as 2.6 gCOD gVSS(-1) day(-1) for H(2)/CO(2)). Analysis of the microbial community revealed that sugar-degrading acid-forming bacteria proliferated in the sludge of the USSB as well as the acidification reactor at high OLR conditions.

Scanning tunneling microscope assembly, reactor, and system

DOEpatents

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.

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.

Weld monitor and failure detector for nuclear reactor system

DOEpatents

Sutton, Jr., Harry G.

1987-01-01

Critical but inaccessible welds in a nuclear reactor system are monitored throughout the life of the reactor by providing small aperture means projecting completely through the reactor vessel wall and also through the weld or welds to be monitored. The aperture means is normally sealed from the atmosphere within the reactor. Any incipient failure or cracking of the weld will cause the environment contained within the reactor to pass into the aperture means and thence to the outer surface of the reactor vessel where its presence is readily detected.

NEUTRONIC REACTOR

DOEpatents

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

1958-09-01

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

D.M. McEligot; K. G. Condie; G. E. McCreery

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 Generationmore » IV program.« less

Validation of IRDFF in 252Cf standard and IRDF-2002 reference neutron fields

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simakov, Stanislav; Capote Noy, Roberto; Greenwood, Lawrence R.

The results of validation of the latest release of International Reactor Dosimetry and Fusion File, IRDFF-1.03, in the standard 252Cf(s.f.) and reference 235U(nth,f) neutron benchmark fields are presented. The spectrum-averaged cross sections were shown to confirm the recommended spectrum for 252Cf spontaneous fission source; that was not the case for the current recommended spectra for 235U(nth,f). IRDFF was also validated in the spectra of the research reactor facilities ISNF, Sigma-Sigma and YAYOI, which are available in the IRDF- 2002 collection. Before this analysis, the ISFN spectrum was resimulated to remove unphysical oscillations in spectrum. IRDFF-1.03 was shown to reasonably reproducemore » the spectrum-averaged data measured in these fields except for the case of YAYOI.« less

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)

Advanced Demonstration and Test Reactor Options Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Petti, David Andrew; Hill, R.; Gehin, J.

Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power’s share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercializationmore » of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy’s (DOE’s) broader commitment to pursuing an “all of the above” clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate “advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy�

Optimally moderated nuclear fission reactor and fuel source therefor

DOEpatents

Ougouag, Abderrafi M [Idaho Falls, ID; Terry, William K [Shelley, ID; Gougar, Hans D [Idaho Falls, ID

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.

Top shield temperatures, C and K Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Agar, J.D.

1964-12-28

A modification program is now in progress at the C and K Reactors consisting of an extensive renovation of the graphite channels in the vertical safety rod ststems. The present VSR channels are being enlarged by a graphite coring operation and channel sleeves will be installed in the larger channels. One problem associated with the coring operation is the danger of damaging top thermal shield cooling tubes located close to the VSR channels to such an extent that these tubes will have to be removed from service. If such a condition should exist at one or a number of locationsmore » in the top shield of the reactors after reactor startup, the question remains -- what would the resulting temperatures be of the various components of the top shields? This study was initiated to determine temperature distributions in the top shield complex at the C and K Reactors for various top thermal shield coolant system conditions. Since the top thermal shield cooling system at C Reactor is different than those at the K Reactors, the study was conducted separately for the two different systems.« less

REACTOR SHIELD

DOEpatents

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

1959-02-17

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

Advanced Test Reactor Tour

ScienceCinema

Miley, Don

2017-12-21

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored.

Methanation assembly using multiple reactors

DOEpatents

Jahnke, Fred C.; Parab, Sanjay C.

2007-07-24

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

Nuclear reactor control column

DOEpatents

Bachovchin, Dennis M.

1982-01-01

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

NEUTRONIC REACTOR SYSTEM

DOEpatents

Goett, J.J.

1961-01-24

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

Nuclear reactor overflow line

DOEpatents

Severson, Wayne J.

1976-01-01

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

Pressure-accelerated azide-alkyne cycloaddition: micro capillary versus autoclave reactor performance.

PubMed

Borukhova, Svetlana; Seeger, Andreas D; Noël, Timothy; Wang, Qi; Busch, Markus; Hessel, Volker

2015-02-01

Pressure effects on regioselectivity and yield of cycloaddition reactions have been shown to exist. Nevertheless, high pressure synthetic applications with subsequent benefits in the production of natural products are limited by the general availability of the equipment. In addition, the virtues and limitations of microflow equipment under standard conditions are well established. Herein, we apply novel-process-window (NPWs) principles, such as intensification of intrinsic kinetics of a reaction using high temperature, pressure, and concentration, on azide-alkyne cycloaddition towards synthesis of Rufinamide precursor. We applied three main activation methods (i.e., uncatalyzed batch, uncatalyzed flow, and catalyzed flow) on uncatalyzed and catalyzed azide-alkyne cycloaddition. We compare the performance of two reactors, a specialized autoclave batch reactor for high-pressure operation up to 1800 bar and a capillary flow reactor (up to 400 bar). A differentiated and comprehensive picture is given for the two reactors and the three methods of activation. Reaction speedup and consequent increases in space-time yields is achieved, while the process window for favorable operation to selectively produce Rufinamide precursor in good yields is widened. The best conditions thus determined are applied to several azide-alkyne cycloadditions to widen the scope of the presented methodology. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

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.

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.

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.

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.

NEUTRONIC REACTOR POWER PLANT

DOEpatents

Metcalf, H.E.

1962-12-25

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

REACTOR

DOEpatents

Christy, R.F.

1961-07-25

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

Phenomenology and modeling of particulate corrosion product behavior in Hanford N Reactor primary coolant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bechtold, D.B.

1983-12-31

The levels and composition of filterable corrosion products in the Hanford N Reactor Primary Loop are measurable by filtration. The suspended crud level has ranged from 0.0005 ppM to 6.482 ppM with a median 0.050 ppM. The composition approximates magnetite. The particle size distribution has been found in 31 cases to be uniformly a log normal distribution with a count median ranging from 1.10 to 2.31 microns with a median of 1.81 microns, and the geometric standard deviation ranging from 1.60 to 2.34 with a median of 1.84. An auto-correcting inline turbidimeter was found to respond to linearly to suspendedmore » crud levels over a range 0.05 to at least 6.5 ppM by direct comparison with filter sample weights. Cause of crud bursts in the primary loop were found to be power decreases. The crud transients associated with a reactor power drop, several reactor shutdowns, and several reactor startups could be modeled consistently with each other using a simple stirred-tank, first order exchange model of particulate between makeup, coolant, letdown, and loosely adherent crud on pipe walls. Over 3/10 of the average steady running particulate crud level could be accounted for by magnetically filterable particulate in the makeup feed. A simulation model of particulate transport has been coded in FORTRAN.« less

Reactor power system deployment and startup

NASA Technical Reports Server (NTRS)

Wetch, J. R.; Nelin, C. J.; Britt, E. J.; Klein, G.

1985-01-01

This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems.

Dynamic Response Testing in an Electrically Heated Reactor Test Facility

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.; Morton, T. J.

2006-01-01

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

Comparing the new generation accelerator driven subcritical reactor system (ADS) to traditional critical reactors

NASA Astrophysics Data System (ADS)

Kemah, Elif; Akkaya, Recep; Tokgöz, Seyit Rıza

2017-02-01

In recent years, the accelerator driven subcritical reactors have taken great interest worldwide. The Accelerator Driven System (ADS) has been used to produce neutron in subcritical state by the external proton beam source. These reactors, which are hybrid systems, are important in production of clean and safe energy and conversion of radioactive waste. The ADS with the selection of reliability and robust target materials have been the new generation of fission reactors. In addition, in the ADS Reactors the problems of long-lived radioactive fission products and waste actinides encountered in the fission process of the reactor during incineration can be solved, and ADS has come to the forefront of thorium as fuel for the reactors.

Small and medium power reactors 1987

NASA Astrophysics Data System (ADS)

1987-12-01

This TECDOC follows the publication of TECDOC-347: Small and Medium Power Reactors (SMPR) Project Initiation Study, Phase 1, published in 1985 and TECDOC-376: Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power program. It consists of two parts: (1) guidelines for the introduction of small and medium power reactors in developing countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of small and medium power reactors in developing countries; (2) up-dated information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex 1 of the above mentioned TECDOC-347.

Advantages of liquid fluoride thorium reactor in comparison with light water reactor

NASA Astrophysics Data System (ADS)

Bahri, Che Nor Aniza Che Zainul; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

2015-04-01

Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

Altering recharge dynamics through woody vegetation removal: a study on the Carrizo-Wilcox aquifer of south Texas

NASA Astrophysics Data System (ADS)

Mattox, A. M.

2011-12-01

Grasslands in many semi-arid regions of the world have seen an expansion of woody vegetation over the past century and many now exist largely as woodlands or shrublands. This "woody encroachment" results in numerous changes to ecosystem function, including alteration of element and water cycles. As in many parts of the world, these shrublands in south Texas have been subjected to a variety of management practices intended to reduce woody vegetation and increase the dominance of herbaceous vegetation. In addition to the intended change in vegetation structure, this activity has the potential to affect hydrologic fluxes and potentially increase deep drainage through reduced transpiration and rooting depths. However, there is significant uncertainty about the hydrologic response of vegetation to woody vegetation removal. We report here the results of a large manipulative experiment designed to assess the effects of woody vegetation removal on soil moisture movement in the vadose zone in an area that serves as a recharge zone for an unconsolidated sediment aquifer (Carrizo-Wilcox). In this study woody vegetation has been removed using a mechanical method (roller chopping) as well as a mechanical and chemical method (chainsaw removal + stump herbicide). The treated plots are located on three different soil types that represent the range of soils typical in this area. A water balance approach is used to assess soil moisture fluxes and potential deep drainage. In this first year of the study we quantified ecological and edaphic components that have the greatest effect on deep drainage, namely rooting depth, soil texture and antecedent soil water conditions. Exceptionally dry conditions this year have provided a unique opportunity to understand plant soil water interactions in the critical zone given the strong soil moisture limitations observed in the surface soil horizons. Understanding these interactions across different plant communities and soil textures are the

Radiation hardness of Efratom M-100 rubidium frequency standard

NASA Technical Reports Server (NTRS)

English, T. C.; Vorwerk, H.; Rudie, N. J.

1983-01-01

The effects of nuclear radiation on rubidium gas cell frequency standards and components are presented, including the results of recent tests where a continuously operating rubidium frequency standard (Effratom, Model M-100) was subjected to simultaneous neutron/gamma radiation. At the highest neutron fluence 7.5 10 to the 12th power n/sq cm and total dose 11 krad(Si) tested, the unit operated satisfactorily; the total frequency change over the 2 1/2 hour test period due to all causes, including repeated retraction from and insertion into the reactor, was less than 1 x 10 to the -10th power. The effects of combined neutron/gamma radiation on rubidium frequency standard physics package components were also studied, and the results are presented.

The behaviour of transuranic mixed oxide fuel in a Candu-900 reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morreale, A. C.; Ball, M. R.; Novog, D. R.

2012-07-01

The production of transuranic actinide fuels for use in current thermal reactors provides a useful intermediary step in closing the nuclear fuel cycle. Extraction of actinides reduces the longevity, radiation and heat loads of spent material. The burning of transuranic fuels in current reactors for a limited amount of cycles reduces the infrastructure demand for fast reactors and provides an effective synergy that can result in a reduction of as much as 95% of spent fuel waste while reducing the fast reactor infrastructure needed by a factor of almost 13.5 [1]. This paper examines the features of actinide mixed oxidemore » fuel, TRUMOX, in a CANDU{sup R}* nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 3.1 wt% actinide MOX fuel. Full lattice cell modeling was performed using the WIMS-AECL code, super-cell calculations were analyzed in DRAGON and full core analysis was executed in the RFSP 2-group diffusion code. A time-average full core model was produced and analyzed for reactor coefficients, reactivity device worth and online fuelling impacts. The standard CANDU operational limits were maintained throughout operations. The TRUMOX fuel design achieved a burnup of 27.36 MWd/kg HE. A full TRUMOX fuelled CANDU was shown to operate within acceptable limits and provided a viable intermediary step for burning actinides. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle. (authors)« less

A DOUBLE-RING ALGORITHM FOR MODELING SOLAR ACTIVE REGIONS: UNIFYING KINEMATIC DYNAMO MODELS AND SURFACE FLUX-TRANSPORT SIMULATIONS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Munoz-Jaramillo, Andres; Martens, Petrus C. H.; Nandy, Dibyendu

The emergence of tilted bipolar active regions (ARs) and the dispersal of their flux, mediated via processes such as diffusion, differential rotation, and meridional circulation, is believed to be responsible for the reversal of the Sun's polar field. This process (commonly known as the Babcock-Leighton mechanism) is usually modeled as a near-surface, spatially distributed {alpha}-effect in kinematic mean-field dynamo models. However, this formulation leads to a relationship between polar field strength and meridional flow speed which is opposite to that suggested by physical insight and predicted by surface flux-transport simulations. With this in mind, we present an improved double-ring algorithmmore » for modeling the Babcock-Leighton mechanism based on AR eruption, within the framework of an axisymmetric dynamo model. Using surface flux-transport simulations, we first show that an axisymmetric formulation-which is usually invoked in kinematic dynamo models-can reasonably approximate the surface flux dynamics. Finally, we demonstrate that our treatment of the Babcock-Leighton mechanism through double-ring eruption leads to an inverse relationship between polar field strength and meridional flow speed as expected, reconciling the discrepancy between surface flux-transport simulations and kinematic dynamo models.« less

Versatile Oxide Films Protect FeCrAl Alloys Under Normal Operation and Accident Conditions in Light Water Power Reactors

NASA Astrophysics Data System (ADS)

Rebak, Raul B.

2018-02-01

The US has currently a fleet of 99 nuclear power light water reactors which generate approximately 20% of the electricity consumed in the country. Near 90% of the reactors are at least 30 years old. There are incentives to make the existing reactors safer by using accident tolerant fuels (ATF). Compared to the standard UO2-zirconium-based system, ATF need to tolerate loss of active cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. Ferritic iron-chromium-aluminum (FeCrAl) alloys have been identified as an alternative to replace current zirconium alloys. They contain Fe (base) + 10-22 Cr + 4-6 Al and may contain smaller amounts of other elements such as molybdenum and traces of others. FeCrAl alloys offer outstanding resistance to attack by superheated steam by developing an alumina oxide on the surface in case of a loss of coolant accident like at Fukushima. FeCrAl alloys also perform well under normal operation conditions both in boiling water reactors and pressurized water reactors because they are protected by a thin oxide rich in chromium. Under normal operation condition, the key element is Cr and under accident conditions it is Al.

Reactor operation environmental information document

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haselow, J.S.; Price, V.; Stephenson, D.E.

1989-12-01

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

Plasma spark discharge reactor and durable electrode

DOEpatents

Cho, Young I.; Cho, Daniel J.; Fridman, Alexander; Kim, Hyoungsup

2017-01-10

A plasma spark discharge reactor for treating water. The plasma spark discharge reactor comprises a HV electrode with a head and ground electrode that surrounds at least a portion of the HV electrode. A passage for gas may pass through the reactor to a location proximate to the head to provide controlled formation of gas bubbles in order to facilitate the plasma spark discharge in a liquid environment.

Westinghouse Small Modular Reactor nuclear steam supply system design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Memmott, M. J.; Harkness, A. W.; Van Wyk, J.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development andmore » integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

Neutronic Reactor III

NASA Astrophysics Data System (ADS)

Fermi, Enrico; Zinn, Walter H.; Anderson, Herbert L.

An improvement of the reactors described in the previous Patents, aimed at increasing the reproduction factor, is reported here, such improvement being obtained by diminishing the neutron loss due to impurities within the reactor. This is achieved by encasing the reactor in a rubberized balloon cloth housing (or something like this) in order to eliminate the atmospheric air therefrom, thus eliminating both the effect of the danger coefficient of nitrogen (70% of the atmospheric air) and that of the argon present in the air, which can become radioactive. Since the removal of the air from the reactor may result in structural problems, caused by the forces brought into play by that evacuation, the reactor is then filled with a non-reactive (from a chemical and nuclear standpoint) gas such as helium or carbon dioxide. It is interesting to point out that the authors consider also the possibility to control (a little) the reproduction ratio of the reactor by varying the air content of it. Just a rapid mention of the main idea of the present Patent (i.e. the encasing of the pile in a balloon cloth) appeared in [Fermi (1942f)], but no detailed description of the system considered here is reported in any other published paper.

27 CFR 9.156 - Diablo Grande.

Code of Federal Regulations, 2011 CFR

2011-04-01

...”. (b) Approved maps. The appropriate maps for determining the boundary of the Diablo Grande...) Wilcox Ridge, California—Stanislaus Co., 1956 (Photorevised 1971); (4) Orestimba Peak, California...., R. 6E., on the Wilcox Ridge Quadrangle U.S.G.S. map. (3) Then proceed due west to Oristimba Creek in...

27 CFR 9.156 - Diablo Grande.

Code of Federal Regulations, 2010 CFR

2010-04-01

...”. (b) Approved maps. The appropriate maps for determining the boundary of the Diablo Grande...) Wilcox Ridge, California—Stanislaus Co., 1956 (Photorevised 1971); (4) Orestimba Peak, California...., R. 6E., on the Wilcox Ridge Quadrangle U.S.G.S. map. (3) Then proceed due west to Oristimba Creek in...

Reactor shutdown experience

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cletcher, J.W.

1995-10-01

This is a regular report of summary statistics relating to recent reactor shutdown experience. The information includes both number of events and rates of occurence. It was compiled from data about operating events that were entered into the SCSS data system by the Nuclear Operations Analysis Center at the Oak ridge National Laboratory and covers the six mont period of July 1 to December 31, 1994. Cumulative information, starting from May 1, 1994, is also reported. Updates on shutdown events included in earlier reports is excluded. Information on shutdowns as a function of reactor power at the time of themore » shutdown for both BWR and PWR reactors is given. Data is also discerned by shutdown type and reactor age.« less