SaLEM (v1.0) - the Soil and Landscape Evolution Model (SaLEM) for simulation of regolith depth in periglacial environments

NASA Astrophysics Data System (ADS)

Bock, Michael; Conrad, Olaf; Günther, Andreas; Gehrt, Ernst; Baritz, Rainer; Böhner, Jürgen

2018-04-01

We propose the implementation of the Soil and Landscape Evolution Model (SaLEM) for the spatiotemporal investigation of soil parent material evolution following a lithologically differentiated approach. Relevant parts of the established Geomorphic/Orogenic Landscape Evolution Model (GOLEM) have been adapted for an operational Geographical Information System (GIS) tool within the open-source software framework System for Automated Geoscientific Analyses (SAGA), thus taking advantage of SAGA's capabilities for geomorphometric analyses. The model is driven by palaeoclimatic data (temperature, precipitation) representative of periglacial areas in northern Germany over the last 50 000 years. The initial conditions have been determined for a test site by a digital terrain model and a geological model. Weathering, erosion and transport functions are calibrated using extrinsic (climatic) and intrinsic (lithologic) parameter data. First results indicate that our differentiated SaLEM approach shows some evidence for the spatiotemporal prediction of important soil parental material properties (particularly its depth). Future research will focus on the validation of the results against field data, and the influence of discrete events (mass movements, floods) on soil parent material formation has to be evaluated.

Aerial radiological survey of the Salem Nuclear Generating Station and surrounding area, Salem, New Jersey

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

Feimster, E.L.

An aerial radiological survey was performed from 24 July through 1 August 1980 over a 244-square-kilometer (95-square-mile) area centered on the Salem Nuclear Generating Station near Salem, New Jersey. All gamma ray data were collected by flying lines oriented north-south and spaced 152 meters (500 feet) apart, at an altitude of 91 meters (300 feet) above the ground. Processed data showed that all gamma rays detected within the survey area were those expected from naturally occurring terrestrial background emitters except directly over the site, where spectral analysis revealed the presence of /sup 60/Co. Count rates obtained from the aerial platformmore » were converted to exposure rates at 1 meter above the ground and are presented in the form of an exposure rate contour map. The resulting exposure rates varied between 5 and 55 microroentgens per hour (..mu..R/h). The river-shore areas ranged from 5 to 7 ..mu..R/h, inland areas showed 7 to 12 ..mu..R/h, and the site had a maximum exposure rate of 55 ..mu..R/h. These values include an estimated cosmic ray contribution of 4 ..mu..R/h. The exposure rates obtained from soil samples taken within the survey area displayed good agreement with the aerial data.« less

Supplement Analysis for the Transmission System Vegetation Management Program FEIS (DOE/EIS-0285/SA-137- Chemawa-Salem 1&2)

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

Sherer, Brett M.

Vegetation Management for the Chemawa-Salem #1 115 kV and #2 230 kV transmission lines from Chemawa Substation to Salem Substation. BPA proposes to remove unwanted vegetation along the right-of-way, access roads, switch platforms, and around tower structures of the subject transmission line corridor that may impede the operation and maintenance of the identified transmission lines. BPA plans to conduct vegetation control with the goal of removing tall growing vegetation that is currently or will soon be a hazard to the transmission line. BPA’s overall goal is to have lowgrowing plant communities along the rights-of-way to control the development of potentiallymore » threatening vegetation.« less

Witchcraft or mycotoxin? The Salem witch trials.

PubMed

Woolf, A

2000-01-01

The Salem witchcraft trials of 1692 have been studied by many historians looking for the complex social, political, and psychological determinants behind the community-wide hysteria that led to a travesty of justice and the deaths of 20 innocent Puritans. Recently, ergot poisoning has been put forth by some as a previously unsuspected cause of the bizarre behaviors of the young adolescent girls who accused the townsfolk of witchcraft. In this essay the circumstances behind the ergot poisoning theory for this historical event are described. When the evidence is weighed carefully both pro and con, it seems unlikely that ergotism explains much of what went on in colonial Salem.

76 FR 19148 - PSEG Nuclear, LLC, Hope Creek Generating Station and Salem Nuclear Generating Station, Units 1...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-04-06

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-272, 50-311, 50-354; NRC-2009-0390 and NRC-2009-0391] PSEG Nuclear, LLC, Hope Creek Generating Station and Salem Nuclear Generating Station, Units 1 and 2; Notice of Availability of the Final Supplement 45 to the Generic Environmental Impact Statement for License Renewal of Nuclear Plants Notice is...

Salem Program Demonstrates Five Keys to a Successful Business--School Partnership.

ERIC Educational Resources Information Center

Jensen, Mary Cihak

1985-01-01

Partnerships between schools and businesses are becoming more frequent and more reciprocal as the two find common interests; the Salem-Keizer (Oregon) Public Schools provide a nationally recognized example of such a partnership. Salem's program began in 1981 with a joint school board and Chamber of Commerce policy statement. Despite economic…

Low-temperature overpressurization protection system setpoint analysis using RETRAN-02/MOD5 for Salem

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

Dodson, R.J.; Feltus, M.A.

The low-temperature overpressurization protection system (LTOPS) is designed to protect the reactor pressure vessel (RPV) from brittle failure during startup and cooldown maneuvers in Westinghouse pressurized water reactors. For the Salem power plants, the power-operated relief valves (PORVs) mitigate pressure increases above a setpoint where an operational startup transient may put the RPV in the embrittlement fracture zone. The Title 10, Part 50, Code of Federal Regulations Appendix G limit, given by plant technical specifications, conservatively bounds the maximum pressure allowed during those transients where the RPV can suffer brittle fracture (usually below 350{degrees}F). The Appendix G limit is amore » pressure versus temperature curve that is more restrictive at lower RPV temperatures and allows for higher pressures as the temperature approaches the upper bounding fracture temperature.« less

Copper in the intake and discharge zones of the Surry and Salem Nuclear Power Stations

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

Harrison, F.L.; Bishop, D.J.; Rice, D.W. Jr.

Copper concentrations were measured in the soluble and particulate fractions of water samples and bedload sediments collected in intake and discharge areas of the Surry and Salem Nuclear Power Stations during normal operations. Additional samples of water and suspended particles were collected during startup of Unit 2 at the Salem Power Station. In water samples collected from Surry, total copper ranged from 6.5 to 24.7 and labile copper from 0.5 to 2.9 ..mu..g/L; in those from Salem, total copper ranged from 6.7 to 10.6 and labile from 0.9 to 3.8 ..mu..g/L. At both sites the highest total copper concentration wasmore » measured in January 1979 during a period of high runoff. In general, differences between influent and effluent waters were small; the maximum was 4.2 ..mu..g Cu/L. Copper concentration in the water during startup of Unit 2 of Salem was high initially (>2500 ..mu..g Cu/L) but was almost entirely in the particulate fraction; labile copper was only 0.6 ..mu..g/L. The apparent complexing capacity (ACC) of the waters from Surry ranged from 6 to 40 and those from Salem from 5 to 60 ..mu..g Cu/L. Ranges in dissolved organic carbon were smaller, 2.9 to 5.1 and 2.2 to 5.0 mg C/L for Surry and Salem, respectively, and showed no relationship with ACC. Ultrafiltration of discharge waters indicated that, in most samples, the largest fraction of copper in the untreated water was in the >10,000 <100,000 molecular weight fraction; in waters treated to destroy dissolved organic carbon, it was generally in the >100,000 molecular weight fraction.Copper concentrations in intact bedload sediments from the intake area of Surry ranged from 2.3 to 26 and of Salem from 36 to 74 ..mu..g/g dry weight; those in the discharge area of Surry ranged from 13 to 30 and of Salem from 3 to 67. We noted considerable spatial heterogeneity both at the intake and discharge areas, and higher copper concentrations in the <62-..mu..m fraction than in intact sediments.« less

75 FR 6223 - PSEG Nuclear LLC; Hope Creek Generating Station and Salem Nuclear Generating Station, Unit Nos. 1...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-08

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-272, 50-311 and 50-354; NRC-2010-0043] PSEG Nuclear LLC; Hope Creek Generating Station and Salem Nuclear Generating Station, Unit Nos. 1 and 2; Environmental Assessment and Finding of No Significant Impact The U.S. Nuclear Regulatory Commission (NRC) is considering issuance of an Exemption, pursuant...

Sediment oxygen demand in the Saddle River and Salem River watersheds, New Jersey, July-August 2008

USGS Publications Warehouse

Heckathorn, Heather A.; Gibs, Jacob

2010-01-01

Many factors, such as river depth and velocity, biochemical oxygen demand, and algal productivity, as well as sediment oxygen demand, can affect the concentration of dissolved oxygen in the water column. Measurements of sediment oxygen demand, in conjunction with those of other water-column water-quality constituents, are useful for quantifying the mechanisms that affect in-stream dissolved-oxygen concentrations. Sediment-oxygen-demand rates are also needed to develop and calibrate a water-quality model being developed for the Saddle River and Salem River Basins in New Jersey to predict dissolved-oxygen concentrations. This report documents the methods used to measure sediment oxygen demand in the Saddle River and Salem River watersheds along with the rates of sediment oxygen demand that were obtained during this investigation. In July and August 2008, sediment oxygen demand was measured in situ in the Saddle River and Salem River watersheds. In the Saddle River Basin, sediment oxygen demand was measured twice at two sites and once at a third location; in the Salem River Basin, sediment oxygen demand was measured three times at two sites and once at a third location. In situ measurements of sediment oxygen demand in the Saddle River and Salem River watersheds ranged from 0.8 to 1.4 g/m2d (grams per square meter per day) and from 0.6 to 7.1 g/m2d at 20 degrees Celsius, respectively. Except at one site in this study, rates of sediment oxygen demand generally were low. The highest rate of sediment oxygen demand measured during this investigation, 7.1 g/m2d, which occurred at Courses Landing in the Salem River Basin, may be attributable to the consumption of oxygen by a large amount of organic matter (54 grams per kilogram as organic carbon) in the streambed sediments or to potential error during data collection. In general, sediment oxygen demand increased with the concentration of organic carbon in the streambed sediments. Repeated measurements made 6 to 7 days apart

Salem witchcraft and lessons for contemporary forensic psychiatry.

PubMed

Friedman, Susan Hatters; Howie, Andrew

2013-01-01

In 1692 and 1693, in Salem, Massachusetts, more than 150 colonists were accused of witchcraft, resulting in 19 being hanged and one man being crushed to death. Contributions to these events included: historical, religious and cultural belief systems; social and community concerns; economic, gender, and political factors; and local family grievances. Child witnessing, certainty of physician diagnosis, use of special evidence in the absence of scholarly and legal scrutiny, and tautological reasoning were important factors, as well. For forensic psychiatry, the events at Salem in 1692 still hold contemporary implications. These events of three centuries ago call to mind more recent daycare sexual abuse scandals.

Mineralogical characterization of the Shelburne marble and the Salem limestone

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

McGee, E.S.

Samples of Shelburne marble and Salem limestone were selected to represent marbles and limestones used in buildings and monuments. The Royal variety of Shelburne marble is a white marble predominantly composed of calcite but has heterogeneously distributed gray inclusions. The select buff Salem limestone is a beige, homogeneous, fossiliferous limestone, predominantly composed of fragments of echinoderms and bryozoans. The author reports that both samples are appropriate test stones for the National Acid Precipitation Assessment Program stone exposure studies.

Salem 98: A post-plume phase, federal participation exercise

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

NONE

Salem 98 was the largest nuclear power plant post-plume phase exercise since the 1993 FRMAC-93 exercise at the Fort Calhoun Nuclear Power Plant in Nebraska. Salem 98 was a 3 Day exercise, held on May 5--7, 1998, involving participation by the States of New Jersey and Delaware and associated State and county agencies. Public Service Electric and Gas was the host utility and Salem County the host county. Federal participation included the Nuclear Regulatory Commission, Federal Emergency Management Agency, Department of Energy, Environmental Protection Agency, US department of Agriculture and Department of Health and Human Services. In addition, the Americanmore » Nuclear Insurers participated, adding a dimension to the exercise not experienced often enough. This was a stand-alone post-plume phase exercise, which took place 2 months after the evaluated plume phase exercise held on March 3, 1998, also including participation by various Federal agencies. This exercise demonstrated the positive working relationship among utility, State, county, and Federal responders in response to a postulated major nuclear power plant emergency with significant offsite consequences.« less

Internet Witch-Hunt: Digitizing Salem Village.

ERIC Educational Resources Information Center

Ferris, William R.; Ray, Benjamin

2000-01-01

Presents an interview with Benjamin Ray, a professor of religious studies at the University of Virginia (Charlottesville), focusing on Ray's work with the Salem Witchcraft Project. Addresses how Ray became interested in the topic and the use of the Internet within the discipline of history. (CMK)

Pathogen Transport and Fate Modeling in the Upper Salem River Watershed using SWAT Model

EPA Science Inventory

SWAT (Soil and Water Assessment Tool) is a dynamic watershed model that is applied to simulate the impact of land management practices on water quality over a continuous period. The Upper Salem River, located in Salem County New Jersey, is listed by the New Jersey Department of ...

Multi-Century Record of Anthropogenic Impacts on an Urbanized Mesotidal Estuary: Salem Sound, MA

EPA Science Inventory

Salem, MA, located north of Boston, has a rich, well-documented history dating back to settlement in 1626 CE, but the associated anthropogenic impacts on Salem Sound are poorly constrained. This project utilized dated sediment cores from the sound to assess the proxy record of an...

Pathogen Transport and Fate Modeling in the Upper Salem River Watershed Using SWAT Model

EPA Science Inventory

SWAT (Soil and Water Assessment Tool) is a dynamic watershed model that is applied to simulate the impact of land management practices on water quality over a continuous period. The Upper Salem River, located in Salem County New Jersey, is listed by the New Jersey Department of ...

Anthropogenic Influences on Estuarine Sedimentation in Salem Sound, MA

NASA Astrophysics Data System (ADS)

Kristiansen, E. R.; Hubeny, J. B.; Zhu, J.; Olsen, C. R.; Warren, B.

2010-12-01

The Salem Sound watershed (MA) historically has been a region of significant industrial activity. Two specific point sources for pollution in the region are the South Essex Sewerage District (SESD) wastewater treatment facility, and the Salem Harbor Power Station, a coal-burning power plant. This study tests the hypothesis that human impact on Salem Sound is preserved in the sediment record. A sediment core was taken near the location of the SESD outfall. This core was analyzed for content of organic matter via loss on ignition (LOI), as well as magnetic susceptibility. An age model was constructed using 137Cs and 210Pb. Below 31 cm (mid-nineteenth century), the core contains mean background values of 2.7% LOI and values increase above this depth. At 21cm, a rapid increase in organic matter concentration from 6.6% to 11.8% is observed. This depth corresponds to ~1905 which is contemporaneous with construction of the outfall pipe discharging raw wastewater. At a depth of 7 cm (mid 1970s), LOI values decrease from 11.7% to 9.3%. This shift is likely attributed to SESD beginning primary treatment in 1977. LOI values continue to drop at 2cm (late 1990s), from 7.8% to 6.3%, and remain at 6.1% to the modern surface, likely a result of SESD upgrading to secondary treatment in 1998. Magnetic susceptibility also shows variability down core that is likely attributed to human impact. At a depth of approximately 20cm susceptibility values start increasing from 4.2 SI units until they reach a peak at 15cm (8.8 SI units). This increase can be attributed to the industrial revolution and increased industrial activity in the area. A decrease in susceptibility is observed at 15cm to 11cm (5.6 SI units) that may be attributed to the Great Depression and less fossil fuels being burned due to the economic situation. At approximately 10cm and 8.0 SI units, an increasing trend is first observed. This trend continues up to the modern surface where it eventually reaches 19.9 SI units. This

The Specter of Salem in American Culture.

ERIC Educational Resources Information Center

Adams, Gretchen A.

2003-01-01

Focuses on the impact of the Salem (Massachusetts) witchcraft trials on U.S. society over the centuries, particularly in relation to the trials as a metaphor for persecution and witch-hunting. Explores different events in history, such as the anti-communism evident in the 1950's and abolitionism during the U.S. Civil War. (CMK)

A "clean cigarette" for a clean nation: a case study of Salem Pianissimo in Japan.

PubMed

Assunta, M; Chapman, S

2004-12-01

To illustrate, through internal industry documents, how RJ Reynolds exploited the concerns of the Japanese society about cleanliness to market the concept of cleaner, implicitly healthier cigarettes in Japan. Systematic keyword and opportunistic website searches of formerly private internal industry documents. Industry documents show that RJ Reynolds developed marketing plans based upon their cultural assumptions of Japanese people as fastidious about hygiene and manners, and with relatively high penchants to try new products. RJ Reynolds found there was also a growing concern for health, the environment, and smokers were conscious about annoying others. Deodorised consumer products were one of Japan's biggest trends. These characteristics presented RJ Reynolds with a profitable formula for marketing Salem Pianissimo, a clean cigarette with less smell and smoke. Salem Pianissimo, a 100 mm cigarette claiming to contain 1 mg tar and 0.1 mg nicotine, targeted women since menthol cigarettes were popular among 18-24 year old female smokers, although Japan's law prohibited those below 20 years to smoke and the tobacco industry had a voluntary code disallowing advertising to women and youth. RJ Reynolds successfully launched its clean cigarette, Salem Pianissimo, in Japan aiming to exploit perceived cultural characteristics such as a penchant for cleanliness, an eagerness to try new products, and social harmony.

A "clean cigarette" for a clean nation: a case study of Salem Pianissimo in Japan

PubMed Central

Assunta, M; Chapman, S

2004-01-01

Objective: To illustrate, through internal industry documents, how RJ Reynolds exploited the concerns of the Japanese society about cleanliness to market the concept of cleaner, implicitly healthier cigarettes in Japan. Design: Systematic keyword and opportunistic website searches of formerly private internal industry documents. Results: Industry documents show that RJ Reynolds developed marketing plans based upon their cultural assumptions of Japanese people as fastidious about hygiene and manners, and with relatively high penchants to try new products. RJ Reynolds found there was also a growing concern for health, the environment, and smokers were conscious about annoying others. Deodorised consumer products were one of Japan's biggest trends. These characteristics presented RJ Reynolds with a profitable formula for marketing Salem Pianissimo, a clean cigarette with less smell and smoke. Salem Pianissimo, a 100 mm cigarette claiming to contain 1 mg tar and 0.1 mg nicotine, targeted women since menthol cigarettes were popular among 18–24 year old female smokers, although Japan's law prohibited those below 20 years to smoke and the tobacco industry had a voluntary code disallowing advertising to women and youth. Conclusion: RJ Reynolds successfully launched its clean cigarette, Salem Pianissimo, in Japan aiming to exploit perceived cultural characteristics such as a penchant for cleanliness, an eagerness to try new products, and social harmony. PMID:15564222

75 FR 76040 - Dell Products LP, Winston-Salem (WS-1) Division, Including On-Site Leased Workers From Adecco...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-07

..., Winston-Salem, North Carolina. The notice was published in the Federal Register on April 23, 2010 (75 FR... from Staffing Solutions, South East, and Omni Resources and Recovery. The notices were published in the... firm. The workers are engaged in employment related to the production of desktop computers. New...

76 FR 2710 - Dell Products LP, Winston-Salem (WS-1) Division, Including On-Site Leased Workers From Adecco...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-14

..., APN and ICONMA, Winston-Salem, North Carolina. The notice was published in the Federal Register on... Seaton Corporation. The notices were published on the Federal Register on April 19, 2010 (75 FR 20385... workers are engaged in employment related to the production of desktop computers. New information shows...

Selected elements and organic chemicals in streambed sediment in the Salem area, Oregon, 1999

USGS Publications Warehouse

Tanner, Dwight Q.

2002-09-13

The high levels of contaminants in some Salem area streams indicates the need for further study to assess the biological effects of these contaminants. Future monitoring in the Salem area could include bioassays using benthic invertebrates and the measurement of organochlorine compounds, including DDT, DDE, DDD, and dieldrin in fish tissue. Because resident fish may be consumed by humans and wildlife, fish tissue analyses would be helpful to determine the health risk associated with fish consumption.

Selection, procurement and description of Salem Limestone samples used to study the effects of acid rain

NASA Astrophysics Data System (ADS)

Ross, M.; Knab, L.

1984-07-01

The selection, procurement, and description of the Salem Limestone to be used in field exposure tests to assess the effects of acid rain on building stone are described. The rationale for choosing Salem Limestone is given and a brief geological description of the stone is provided. Preparation of the stone samples for field exposure, including cutting, surface finishing and labelling is presented.

Hazardous Waste Cleanup: Mannington Mills Incorporated in Salem, New Jersey

EPA Pesticide Factsheets

Mannington Mills, Inc. manufactures vinyl based flooring for residential and commercial use. The facility is located on Mannington Mills Road in Salem, New Jersey and has been in operation since 1924. The historic and current manufacturing facility acreage

PATHOGEN TRANSPORT AND FATE MODELING IN THE UPPER SALEM RIVER WATERSHED USING SWAT MODEL - PEER-REVIEWED JOURNAL ARTICLE

EPA Science Inventory

Simulation of the fate and transport of pathogen contamination was conducted with SWAT for the Upper Salem River Watershed, located in Salem County, New Jersey. This watershed is 37 km² and land uses are predominantly agricultural. The watershed drains to a 32 km str...

Circulating water traveling screen modifications to improve impinged fish survival and debris handling at Salem Generating Station

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

Ronafalvy, J.P.; Cheesman, R.R.; Matejek, W.M.

This paper summarizes modifications made to the Salem Generating Station`s (Salem) Circulating Water Traveling Screens (CWTS) as required by Salem`s New Jersey Pollutant Discharge Elimination System Permit (NJPDES Permit). The modifications incorporated newly designed CWTS baskets with hydrodynamically improved fish buckets (greatly reduced turbulence in the bucket); smooth woven mesh screens with 0.250 x 0.500 inch rectangular mesh openings (formerly 0.375 inch square mesh openings); lighter composite basket frame material allowing increased screen rotation speed; improved low and high pressure spray wash patterns; improved screen to sluice trough flap seal design and miscellaneous reliability improvements. In order to address themore » overall effects of the CWTS modifications on fish losses, the effect of both entrainment and impingement by fish size was addressed in a comparison study between modified and unmodified units. The results of the assessment indicate a 51% reduction in overall weakfish (Cynoscion regales) losses (expressed as equivalent adults, larger than 187 mm total length individuals). These modifications also enhanced debris removal capability of the CWTSs. 5 refs., 5 figs.« less

33 CFR 165.510 - Delaware Bay and River, Salem River, Christina River and Schuylkill River-Regulated Navigation Area.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 33 Navigation and Navigable Waters 2 2014-07-01 2014-07-01 false Delaware Bay and River, Salem River, Christina River and Schuylkill River-Regulated Navigation Area. 165.510 Section 165.510... Limited Access Areas Fifth Coast Guard District § 165.510 Delaware Bay and River, Salem River, Christina...

33 CFR 165.510 - Delaware Bay and River, Salem River, Christina River and Schuylkill River-Regulated Navigation Area.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 33 Navigation and Navigable Waters 2 2012-07-01 2012-07-01 false Delaware Bay and River, Salem River, Christina River and Schuylkill River-Regulated Navigation Area. 165.510 Section 165.510... Limited Access Areas Fifth Coast Guard District § 165.510 Delaware Bay and River, Salem River, Christina...

Radio Wave Propagation over Salem

NASA Astrophysics Data System (ADS)

Jaiswal, R. S.; Uma, S.; Raj, M. V. A.

2007-07-01

In this paper study of rainfall has been carried out over Salem, a place in Southern India. Rainfall rate values have been recorded using a fast response rain gauge installed at Sona College of Technology. The derived rainfall rates have been used to estimate attenuation in the 10-100 GHz frequency range. Using the estimated co-polar attenuation cross polar discriminations (XPD) have been computed using ITU-R(2002) model in the 10-35 GHz range. The study shows that attenuation and cross polarization vary with frequency, elevation angle and rainfall rate. The study also depicts the cumulative distribution of rainfall rate, attenuation and XPD.

Mineralogical characterization of the Shelburne Marble and the Salem limestone: Test stones used to study the effects of acid rain

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

McGee, E.S.

The Salem Limestone and the Shelburne Marble are representative of limestones and marbles commonly used in buildings and monuments. Both stones are composed predominantly of calcite. The Salem Limestone is homogeneous in composition and mineralogic characteristics throughout the test block. The Shelburne Marble has compositionally homogeneous mineral phases, but the distribution of those phases within the test block is random. The mineralogy and physical characteristics of the Shelburne Marble and Salem Limestone test blocks described in the study provide a baseline for future studies of the weathering behavior of these stones. Because the Shelburne Marble and the Salem Limestone aremore » representative of typical commercial marbles and limestones, they are likely to be useful in a consortium study of the effects of acid precipitation on these two types of building stones.« less

Vision for the Year 2002: A Working Document. Salem County Board for Vocational Education Strategic Plan for the Salem County Vocational Technical Schools.

ERIC Educational Resources Information Center

Salem County Board for Vocational Education, Woodstown, NJ.

A strategic plan for vocational-technical education (VTE) in Salem County, New Jersey, in the year 2002 was developed by the county board of VTE and county advisory board with input from students, parents, employers, elected officials, and the community in general. Six strategic planning subcommittees were formed to review the district's VTE…

Factor analysis of rock, soil and water geochemical data from Salem magnesite mines and surrounding area, Salem, southern India

NASA Astrophysics Data System (ADS)

Satyanarayanan, M.; Eswaramoorthi, S.; Subramanian, S.; Periakali, P.

2017-09-01

Geochemical analytical data of 15 representative rock samples, 34 soil samples and 55 groundwater samples collected from Salem magnesite mines and surrounding area in Salem, southern India, were subjected to R-mode factor analysis. A maximum of three factors account for 93.8 % variance in rock data, six factors for 84 % variance in soil data, five factors for 71.2 % in groundwater data during summer and six factors for 73.7 % during winter. Total dissolved solids are predominantly contributed by Mg, Na, Cl and SO4 ions in both seasons and are derived from the country rock and mining waste by dissolution of minerals like magnesite, gypsum, halite. The results also show that groundwater is enriched in considerable amount of minor and trace elements (Fe, Mn, Ni, Cr and Co). Nickel, chromium and cobalt in groundwater and soil are derived from leaching of huge mine dumps deposited by selective magnesite mining activity. The factor analysis on trivalent, hexavalent and total Cr in groundwater indicates that most of the Cr in summer is trivalent and in winter hexavalent. The gradational decrease in topographical elevation from northern mine area to the southern residential area, combined regional hydrogeological factors and distribution of ultramafic rocks in the northern part of the study area indicate that these toxic trace elements in water were derived from mine dumps.

Quantum Dynamical Applications of Salem's Theorem

NASA Astrophysics Data System (ADS)

Damanik, David; Del Rio, Rafael

2009-07-01

We consider the survival probability of a state that evolves according to the Schrödinger dynamics generated by a self-adjoint operator H. We deduce from a classical result of Salem that upper bounds for the Hausdorff dimension of a set supporting the spectral measure associated with the initial state imply lower bounds on a subsequence of time scales for the survival probability. This general phenomenon is illustrated with applications to the Fibonacci operator and the critical almost Mathieu operator. In particular, this gives the first quantitative dynamical bound for the critical almost Mathieu operator.

Salem Community College's 1999-2002 Strategic Plan Authoring & Implementation Strategy.

ERIC Educational Resources Information Center

Salem Community Coll., Penns Grove, NJ.

This document outlines the Strategic Planning Initiative (SPI) for New Jersey's Salem Community College. This is the first plan the college has authored in seven years. The report provides a theoretical framework for heterarchical planning, which allows for complexity and interrelations of structural analysis, and lateral decision making. The…

Ground water in the Springfield-Salem plateaus of southern Missouri and northern Arkansas

USGS Publications Warehouse

Harvey, Edward Joseph

1980-01-01

Average ground-water conditions have not changed significantly in the Springfield-Salem plateaus section of southern Missouri and northern Arkansas in the past 25 years except in the vicinity of well fields. The amount of ground water pumped is approximately 200 cubic feet per second, which is about 5 percent of the total discharge at the 80 percent point on flow-duration curves for major streams. Ground-water recharge is variable and occurs through sinkholes by infiltration in upland areas of good permeability, and through streambeds that lose flow. Main waterbearing zones lie in the Potosi Dolomite and the lower dolomite and sandstone of the Gasconade Dolomite. Cavernous connections from ground surface to depths as great as 1,500 feet occur in the West Plains area, Mo., and result in deep circulation of water. Municipal well-water in the area often becomes turbid after rainstorms, despite well depths of 1 ,500 feet and 950 to 1,000 feet of pressure-grouted casing. Ground-water movement is generaly north and south from the crest of the Springfield-Salem plateaus, which extend across southern Missouri from the St. Francois Mountains to the southwest. Interbasin diversion of surface- and ground-water flow is common. (USGS)

Stylolitization as source of cement in Mississippian Salem Limestone, west-central Indiana

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

Finkel, E.A.; Wilkinson, B.H.

The Mississippian Salem Limestone of west-central Indiana is a homogeneous cross-bedded grainstone containing numerous stylolites with amplitudes ranging up to 25 cm. Petrographic and geochemical analyses of closely spaced samples from four 1-m thick stylolite-bounded units document spatial trends in grainstone texture and composition, which correlate with proximity to bounding solution seams. Textural data indicate that stylolitization was locally preceded by grain compaction and that seam solution preferentially occurred within layers where grain packing was tightest. Amount of cement largely corresponds to volume of available pore space, and remaining porosity varies inversely to stylolite proximity. Trace-element compositions demonstrate that intergranularmore » spar is enriched in Mn and depleted in Mg relative to grains, and suggest a significant contribution of carbonate cement to grainstone pores from bounding solution seams. Data on grainstone and stylolite insoluble contents indicate that stylolite amplitude records 43% of actual section shortening. On average, seam solution within the Salem Limestone could have provided no less than 47% and no more than 90% of the CaCO{sub 3}, Fe, and Mn mass now in grainstone pores as intergranular spar cement. As such, stylolitization has played an important role during burial diagenesis, porosity occlusion, and permeability reduction within this Mississippian grainstone sequence. 17 figs., 1 tab.« less

Health Hazard Evaluation Report No. HETA-81-083-933, Salem Area Family Counseling Service, Salem, Oregon

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

Apol, A.G.

In November 1980, the National Institute for Occupational Safety and Health (NIOSH) received a request for a Health Hazard Evaluation from the Salem Area Family Counseling Service to determine the cause of the adverse health responses (headache, dizziness, irritation of the eyes and throat, dry and sore throat, fatigue, and lethargy) experienced by six of the seven employees in the office. NIOSH conducted an environmental survey on December 29-30, 1980. Due to the persistence of symptoms after the carpets and drapes were steamed cleaned, and the ventilation system cleaned, the firm relocated in another office building. Two months after relocation,more » the symptoms appeared to have disappeared.« less

Elusive or Illuminating: Using the Web To Explore the Salem Witchcraft Trials.

ERIC Educational Resources Information Center

Hurter, Stephanie R.

2003-01-01

Presents Web sites useful for teaching about the Salem (Massachusetts) witchcraft trials. Includes Web sites that offer primary source material, collections of Web sites, teaching material, and sites that are interactive, including features, such as QuickTime movies. (CMK)

Assessing Student Threats: A Handbook for Implementing the Salem-Keizer System

ERIC Educational Resources Information Center

Van Dreal, John

2011-01-01

"Assessing Student Threats: A Handbook for Implementing the Salem-Keizer System" is a manual for the implementation of a threat assessment system that follows the recommendations of the Safe Schools Initiative and the prescriptive outline provided by the FBI. Written from an educator's perspective with contributing authors from law…

Prevalence of early childhood caries and its risk factors in 18-72 month old children in Salem, Tamil Nadu.

PubMed

Stephen, Arokiaraj; Krishnan, Ramesh; Ramesh, Maya; Kumar, Vasaviah Suresh

2015-01-01

Early childhood caries (ECC) is one of the most common dental diseases among the preschoolers, leading to suboptimal health. A study was planned to evaluate the prevalence of ECC in preschool children in the age group between 18 and 72 months and its relationship with parent's education and socioeconomic status of the family. This cross-sectional study was conducted among 2771 children selected randomly, in the age group 18-72 months and attending playschools and primary schools in urban, semi-urban, and rural areas of Salem, Tamil Nadu. A modified Winter et al. questionnaire and a proforma were used for collecting information on each child. The completed questionnaire and proformas were statistically analyzed and evaluated. The statistical analysis was done using Student's t-test and analysis of variance (ANOVA) to obtain mean values. The prevalence of ECC in Salem was 16% with a mean dfs of 5.23 ± 1. Prevalence was high among the children of low socioeconomic status group and the children of working mothers, with a mean dfs of 10.47. The prevalence of ECC was 16% in Salem, Tamil Nadu. ECC was seen more in children of working mothers, lower parental education, and lower socioeconomic groups.

Faunas of Mississippian oolitic limestones: Evidence from Salem Limestone, southern Indiana

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

Feldman, H.R.

In the Salem Limestone of southern Indiana, a correlation exists between the faunal assemblage and abundance of grains with superficial oolitic coatings in grainstones. Coarse, poorly sorted fossiliferous grainstones are dominated by an echinoderm-bryozoan-brachiopod assemblage of fossils with few mollusks. The presence of large whole fossils, such as articulated crinoid calyces, suggest limited transport of skeletal components. Grainstones, dominated by grains with superficial coatings, and foraminifers tend to contain a diverse mollusk-dominated assemblage of gastropods, bivalves, rostroconchs, chitins, and scaphopods. These fossils are disarticulated, but generally are not fragmented even though many of them are thin and delicate. Echinoderms, brachiopods,more » and bryozoans are repsented in the mollusk-domdinated assemblage almost exclusively by well-rounded and coated fragments, suggesting that they are not in situ. The presence of similar molluscan assemblages in other Mississippian coated-grain grainstones from Alabama (the Monteagle Limestone) and Oklahoma (an unnamed limestone) indicates that the assemblage may have been wide-spread. Mississippian grainstones dominated by oolites (which are not prominent in the Salem) generally have very few fossils.« less

Appraisal of storm-water quality near Salem, Oregon

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

Miller, T.L.

Stormwater runoff for the period December 1979 to May 1981, at 13 sites in the vicinity of Salem, Oregon, was sampled and analyzed for water quality. Constituent concentrations for urban storm water were relatively small when compared to samples from Portland and Medford, Oregon and to samples from Denver, Colorado. The data indicated that levels of suspended sediment, ultimate CBOD (carbonaceous biochemical oxygen demand), and total lead increased with increased urbanization. Because of small chemical concentrations and winter high flow and low temperature conditions in the Willamette River, Salem storm water probably has little effect on biological or on mostmore » chemical conditions in the Willamette River. An analysis of data from a stormwater detention pond indicated that the facility was about 47% efficient in reducing suspended sediment loads. Precipitation samples collected at one site for a year were found to be acidic, with a median pH of 4.6. Median total lead concentration was 8 micrograms/L (ug/L) in precipitation, whereas the median total lead concentration in runoff from the 12 basins ranged from 8 to 110 ug/L. The median dissolved ammonia concentration in precipitation was larger than the median dissolved ammonia concentration at all 13 sites. In contrast, the median total Kjeldahl nitrogen concentration in precipitation samples was about half the median for streamwater concentrations. Median ratios of sulfate to chloride and nitrate to chloride in precipitation were much higher than ratios expected for sea water, suggesting anthropogenic sources for sulfate and nitrate. 24 refs., 6 figs., 7 tabs.« less

Studies in Teaching: 2011 Research Digest. Research Projects Presented at Annual Research Forum (Winston-Salem, North Carolina, June 15, 2011)

ERIC Educational Resources Information Center

McCoy, Leah P., Ed.

2011-01-01

This document presents the proceedings of 16th Annual Research Forum held June 15, 2011, at Wake Forest University in Winston-Salem, North Carolina. Included herein are the following 25 action research papers: (1) The Effects of Prompted Math Journaling on Algebra 1 Students' Achievement and Attitudes (Heidi I. Arnold); (2) Group Work and Attitude…

75 FR 20033 - Winston-Salem Southbound Railway Company-Corporate Family Transaction Exemption-High Point...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-16

... DEPARTMENT OF TRANSPORTATION Surface Transportation Board [STB Finance Docket No. 35338] Winston-Salem Southbound Railway Company--Corporate Family Transaction Exemption--High Point, Thomasville... under 49 CFR 1180.2(d)(3) for a transaction within a corporate family. Applicants state that HPTD will...

Studies in Teaching: 2007 Research Digest. Research Projects Presented at Annual Research Forum (Winston-Salem, North Carolina, December 12, 2007)

ERIC Educational Resources Information Center

McCoy, Leah P., Ed.

2007-01-01

These Proceedings document an educational research forum held at Wake Forest University (Winston-Salem, North Carolina) on December 12, 2007. Table of contents and 23 research studies of high school teaching are included. Each paper contains a literature review, methodology, results, conclusions, and references. Studies include: (1) Cultural…

Studies in Teaching: 2005 Research Digest. Research Projects Presented at Annual Research Forum (Winston-Salem, North Carolina, December 7, 2005)

ERIC Educational Resources Information Center

McCoy, Leah P., Ed.

2005-01-01

These Proceedings document an educational research forum held at Wake Forest University (Winston-Salem, North Carolina) on December 7, 2005. Table of contents and 26 research studies of high school teaching are included. Studies include: (1) Mathematical Reasoning in Multiple Representations: Connections and Confidence (Justin Allman); (2) The…

Geologic map and database of the Salem East and Turner 7.5-minute quadrangles, Marion County, Oregon: a digital database

USGS Publications Warehouse

Tolan, Terry L.; Beeson, Marvin H.; Digital Database by DuRoss, Christopher B.

2000-01-01

The Salem East and Turner 7.5-minute quadrangles are situated in the center of the Willamette Valley near the western margin of the Columbia River Basalt Group (CRBG) distribution. The terrain within the area is of low to moderate relief, ranging from about 150 to almost 1,100-ft elevation. Mill Creek flows northward from the Stayton basin (Turner quadrangle) to the northern Willamette Valley (Salem East quadrangle) through a low that dissects the Columbia River basalt that forms the Salem Hills on the west and the Waldo Hills to the east. Approximately eight flows of CRBG form a thickness of up to 700� in these two quadrangles. The Ginkgo intracanyon flow that extends from east to west through the south half of the Turner quadrangle is exposed in the hills along the southeast part of the quadrangle. Previous geologic mapping by Thayer (1939) and Bela (1981) while providing the general geologic framework did not subdivide the CRBG which limited their ability to delineate structural elements. Reconnaissance mapping of the CRBG units in the Willamette Valley indicated that these stratigraphic units could serve as a series of unique reference horizons for identifying post-Miocene folding and faulting (Beeson and others, 1985,1989; Beeson and Tolan, 1990). Crenna, et al. (1994) compiled previous mapping in the Willamette Valley in a study of the tectonics of the Salem area. The major emphasis of this study was to identify and map CRBG units within the Salem East and Turner Quadrangles and to utilize this detailed CRBG stratigraphy to identify and characterize structural features. Water well logs were used to provide better subsurface stratigraphic control. Three other quadrangles (Scotts Mills, Silverton, and Stayton NE) in the Willamette Valley have been mapped in this way (Tolan and Beeson, 1999). This area was a lowland area of weathered and eroded marine sedimentary when the Columbia River basalts encroached on this area approximately 15-16 m.y. ago. An

Beyond the Criminal Arena: The Justice Studies Program at Winston-Salem State University

ERIC Educational Resources Information Center

Harvey, Lynn K.; Mitchell, Alvin D.

2006-01-01

The justice studies program at Winston-Salem State University (WSSU) is designed to produce a new breed of justice practitioners whose understanding of justice and its administration is broadened to include the political, social, economic, and cultural conditions within which issues of crime and punishment are pursued and addressed. While it…

Studies in Teaching: 2006 Research Digest. Research Projects Presented at Annual Research Forum (Winston-Salem, North Carolina, December 6, 2006)

ERIC Educational Resources Information Center

McCoy, Leah P., Ed.

2006-01-01

These Proceedings document an educational research forum held at Wake Forest University (Winston-Salem, North Carolina) on December 6, 2006. Table of contents and 27 research studies of high school teaching are included. Studies include: (1) A Study of Teachers' Perceptions of High School Mathematics Instructional Methods (Caroline Adkisson); (2)…

Witches, History, and Microcomputers: A Computer-Assisted Course on the Salem Witch Trials.

ERIC Educational Resources Information Center

Latner, Richard B.

1988-01-01

Describes the addition of a microcomputer component to a Tulane University (Louisiana) undergraduate history course on the Salem witchcraft trials. Discusses the use of a statistical package and a data set to analyze and display data and the enhancement of the active learning approach by introducing students to quantitative methods of historical…

78 FR 11097 - Artificial Island Anchorage No. 2 Partial Closure, Delaware River; Salem, NJ

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-15

... 1625-AA00 Artificial Island Anchorage No. 2 Partial Closure, Delaware River; Salem, NJ AGENCY: Coast... safety zone around the southern portion of Anchorage 2 (Artificial Island Anchorage) below position 39... will cross the closed portion of the anchorage. This regulation is necessary to provide for the safety...

NASA Provides Coast-to-Coast Coverage of Aug. 21 Solar Eclipse (Salem, OR)

NASA Image and Video Library

2017-08-21

On Monday, Aug. 21, NASA provided coast-to-coast coverage of the solar eclipse across America – featuring views of the phenomenon from unique vantage points, including from the ground, from aircraft, and from spacecraft including the ISS, during a live broadcast seen on NASA Television and the agency’s website. This is footage from Salem, Oregon.

Development of reservoir facies and porosity-permeability compartmentalization in middle Mississippian Salem Limestone, eastern margin of the Illinois Basin

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

Ahmad, N.; Keith, B.D.; Thompson, T.A.

Shoal facies of the Salem Limestone in south-central Indiana are composed of (1) bryozoan- and echinoderm-rich (85-95%) grainstones, and (3) foraminiferal-echinodermal grainstones. Microfacies and bedform analyses indicate that deposition took place under tidal conditions where sporadic storm events mobilized the entire sand body. Strong, primarily unidirectional currents, abundant sediment supply, and rapid burial resulted in thick, massive deposits in the Salem shoal. In this hydraulic regime, shape and specific gravity determined the distribution of echinoderm grains relative to heavier bryozoan grains. The relative amount of echinoderms controlled the development of early syntaxial rim cement, which preserved porosity during early compaction.more » A significant amount of calcium carbonate ({approximately}58%) filled the micropores within the echinoderm grains before the intergranular cement could form. Areas such as toes of avalanche forests, where intermediate amounts of echinoderms are mixed with other grains, have greater porosity and permeability. Numerous stylolites also occur throughout the Salem grainstones. Variations in microfacies composition and bedforms are grouped into architectural packages that compartmentalize the porosity and permeability and form micro-to scale macroscale heterogeneities. The basal channel architectural package studied exhibits porosity-permeability compartments orthogonal to the overlying channel-fill sandwave package. In an analogous petroleum reservoir this can result in poor reservoir performance during primary and enhanced recovery leaving behind significant amounts of unrecovered mobile oil. Higher porosity and permeability at the base of a package due to the presence of toesets of a certain grain composition can result in poor sweep efficiency for enhanced recovery.« less

Studies in Teaching: 2003 Research Digest. Research Projects Presented at Annual Research Forum (Winston-Salem, North Carolina, December 10, 2003)

ERIC Educational Resources Information Center

McCoy, Leah P., Ed.

2003-01-01

This document presents the proceedings of an annual educational research forum held at Wake Forest University (Winston-Salem, North Carolina) on December 10, 2003. A table of contents and 31 research studies of high school teaching are included. The following studies are included: (1) No, Seriously: Humor Use by High School Social Studies Teachers…

Studies in Teaching: 2014 Research Digest. Action Research Projects Presented at Annual Research Forum (Winston-Salem, NC, June 26, 2014)

ERIC Educational Resources Information Center

McCoy, Leah P., Ed.

2014-01-01

This document presents the proceedings of the 19th Annual Research Forum held June 26, 2014, at Wake Forest University in Winston-Salem, North Carolina. Included are the following 17 action research papers: (1) Using Voices to Change Minds: Oral Performance and Poetry in the English Classroom (Erika Bunpermkoon), (2) Imagining Audiences: The Use…

Studies in Teaching: 2013 Research Digest. Action Research Projects Presented at Annual Research Forum (Winston-Salem, North Carolina, June 26, 2013)

ERIC Educational Resources Information Center

McCoy, Leah P., Ed.

2013-01-01

This document presents the proceedings of the 18th Annual Research Forum held June 26, 2013, at Wake Forest University in Winston-Salem, North Carolina. Included are the following 13 action research papers: (1) Developing Oral Language Ability in the Secondary Spanish Classroom Using the Interpersonal and Presentational Modes of Communication…

Supplement Analysis for the Transmission System Vegetation Management Program FEIS (DOE/EIS-0285/SA-141- Salem Albany #2)

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

Barndt, Shawn L.

Vegetation Management for the Salem Albany #2 115 kV transmission line from Salem Substation to Albany Substation. BPA proposes to remove unwanted vegetation along the right-of-way, access roads, switch platforms, microwave beam paths, and around tower structures of the subject transmission line corridor that may impede the operation and maintenance of the identified transmission lines. BPA plans to conduct vegetation control with the goal of removing tall growing vegetation that is currently or will soon be a hazard to the transmission line. BPA’s overall goal is to have low-growing plant communities along the rights-of-way to control the development of potentiallymore » threatening vegetation.« less

Studies in Teaching: 2017 Research Digest. Action Research Projects Presented at Annual Research Forum (Winston-Salem, North Carolina, June 29, 2017)

ERIC Educational Resources Information Center

McCoy, Leah P., Ed.

2017-01-01

This document presents the proceedings of the 22nd Annual Research Forum held June 29, 2017, at Wake Forest University in Winston-Salem, North Carolina. Included are the following 12 action research papers: (1) Using Captioned Video to Teach Listening Comprehension in a Spanish Classroom (Michelle Allen); (2) Multimodal Instruction: How Film…

Studies in Teaching: 2012 Research Digest. Action Research Projects Presented at Annual Research Forum (Winston-Salem, North Carolina, June 29, 2012)

ERIC Educational Resources Information Center

McCoy, Leah P., Ed.

2012-01-01

This document presents the proceedings of the 17th Annual Research Forum held June 29, 2012, at Wake Forest University in Winston-Salem, North Carolina. Included herein are the following 25 action research papers: (1) "Reading and Writing": A Study Comparing the Strengths of Peer Review and Visible Author Writing Strategies (Elizabeth Behar); (2)…

Pathogen transport and fate modeling in the Upper Salem River Watershed using SWAT model.

PubMed

Niazi, Mehran; Obropta, Christopher; Miskewitz, Robert

2015-03-15

Simulation of the fate and transport of pathogen contamination was conducted with SWAT for the Upper Salem River Watershed, located in Salem County, New Jersey. This watershed is 37 km(2) and land uses are predominantly agricultural. The watershed drains to a 32 km stretch of the Salem River upstream of the head of tide. This strech is identified on the 303(d) list as impaired for pathogens. The overall goal of this research was to use SWAT as a tool to help to better understand how two pathogen indicators (Escherichia coli and fecal coliform) are transported throughout the watershed, by determining the model parameters that control the fate and transport of these two indicator species. This effort was the first watershed modeling attempt with SWAT to successfully simulate E. coli and fecal coliform simultaneously. Sensitivity analysis has been performed for flow as well as fecal coliform and E. coli. Hydrologic calibration at six sampling locations indicate that the model provides a "good" prediction of watershed outlet flow (E = 0.69) while at certain upstream calibration locations predictions are less representative (0.32 < E < 0.70). Monthly calibration and validation of the pathogen transport and fate model was conducted for both fecal coliform (0.07 < E < 0.47 and -0.94 < E < 0.33) and E. coli (0.03 < E < 0.39 and -0.81 < E < 0.31) for the six sampling points. The fit of the model compared favorably with many similar pathogen modeling efforts. The research contributes new knowledge in E. coli and fecal coliform modeling and will help increase the understanding of sensitivity analysis and pathogen modeling with SWAT at the watershed scale. Copyright © 2014 Elsevier Ltd. All rights reserved.

Studies in Teaching: 2015 Research Digest. Action Research Projects Presented at Annual Research Forum (Winston-Salem, North Carolina, June 25, 2015)

ERIC Educational Resources Information Center

McCoy, Leah P., Ed.

2015-01-01

This document presents the proceedings of the 20th Annual Research Forum held June 25, 2015, at Wake Forest University in Winston-Salem, North Carolina. Included are the following 21 action research papers: (1) History Lives! The Use of Simulations in a High School Social Studies Classroom (Lydia Adkins); (2) Using Francophone Music in the High…

Radioactive cobalt removal from Salem liquid radwaste with cobalt selective media

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

Maza R.; Wilson, J.A.; Hetherington, R.

This paper reports results of benchtop tests using ion exchange material to selectively remove radioactive cobalt from high conductivity liquid radwaste at the Salem Nuclear Generating Station. The purpose of this test program is to reduce the number of curies in liquid releases without increasing the solid waste volume. These tests have identified two cobalt selective materials that together remove radioactive cobalt more effectively than the single component currently used. All test materials were preconditioned by conversion to the divalent calcium or sulfate form to simulate chemically exhausted media.

Low-level radwaste storage facility at Hope Creek and Salem Generating Stations

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

Oyen, L.C.; Lee, K.; Bravo, R.

Following the January 1, 1993, closure of the radwaste disposal facilities at Beatty, Nevada, and Richland, Washington (to waste generators outside the compact), only Barnwell, South Carolina, is open to waste generators in most states. Barnwell is scheduled to stay open to waste generators outside the Southeast Compact until June 30, 1994. Continued delays in opening regional radwaste disposal facilities have forced most nuclear utilities to consider on-site storage of low-level radwaste. Public Service Electric and Gas Company (PSE G) considered several different radwaste storage options before selecting the design based on the steel-frame and metal-siding building design described inmore » the Electric Power Research Institute's (EPRI's) TR-100298 Vol. 2, Project 3800 report. The storage facility will accommodate waste generated by Salem units 1 and 2 and Hope Creek unit 1 for a 5-yr period and will be located within their common protected area.« less

Comparative Evaluation of Dental and Skeletal Fluorosis in an Endemic Fluorosed District, Salem, Tamil Nadu.

PubMed

Ramesh, Maya; Malathi, N; Ramesh, K; Aruna, Rita Mary; Kuruvilla, Sarah

2017-11-01

High levels of fluoride in the drinking water, especially ground water, results in skeletal fluorosis which involves the bone and major joints. This study was conducted to assess the prevalence of skeletal fluorosis to compare with dental fluorosis in an endemically fluorosed population in the District of Salem, Tamil Nadu. Institutional ethical clearance was obtained. A total of 206 patients who reported to the Department of Hematology for blood investigations were the participants in this study. Age, sex, place, weight, height, dental fluorosis, and skeletal complaints were noted down. Body mass index was calculated, and statistical analysis was performed. Dental fluorosis was present in 63.1% and absent in 36.9% of the samples reported. Skeletal fluorosis was present in 24.8% and was absent in 75.2%. A large number of the patients had knee pain and difficulty in bending. Chi-square test was used for statistical analysis. Skeletal fluorosis and age were compared and P value was 0.00 and was significant. Dental fluorosis and skeletal fluorosis were compared and P value was found to be 0.000 and significant. There is a need to take measures to prevent dental and skeletal fluorosis among the residents of Salem district. Calcium balance should be maintained, and fluoride intake should be minimized to reduce the symptoms. The government should provide water with low fluoride level for drinking and cooking. Once the symptoms develop, treatment largely remains symptomatic, using analgesics and physiotherapy.

Proceedings of the first switch tube advanced technology meeting held at EG G, Salem, Massachusetts, May 23, 1990

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

Shuman, A.; Beavis, L.

Early in 1990, J. A. Wilder, Supervisor of Sandia National Laboratories (SNLA), Division 2565 requested that a meeting of the scientists and engineers responsible for developing and producing switch tubes be set up to discuss in a semi-formal way the science and technology of switch tubes. Programmatic and administrative issues were specifically exempted from the discussions. L. Beavis, Division 7471, SNL and A. Shuman, EG G, Salem were made responsible for organizing a program including the materials and processes of switch tubes. The purpose of the Switch Tube Advanced Technology meeting was to allow personnel from Allied Signal Kansas Citymore » Division (AS/KCD); EG G, Salem and Sandia National Laboratories (SNL) to discuss a variety of issues involved in the development and production of switch tubes. It was intended that the formal and informal discussions would allow a better understanding of the production problems by material and process engineers and of the materials and processes by production engineers. This program consisted of formal presentations on May 23 and informal discussions on May 24. The topics chosen for formal presentation were suggested by the people of AS/KCD, EG G, Salem, and SNL involved with the design, development and production of switch tubes. The topics selected were generic. They were not directed to any specific switch tube but rather to all switch tubes in production and development. This document includes summaries of the material presented at the formal presentation on May 23.« less

Catastrophic event recorded among Holocene eolianites (Sidi Salem Formation, SE Tunisia)

NASA Astrophysics Data System (ADS)

Frébourg, Gregory; Hasler, Claude-Alain; Davaud, Eric

2010-03-01

A high-energy deposit cuts through the early Holocene eolianites of the Sidi Salem Formation which forms a ridge along the southeastern coast of Tunisia. The sedimentary structures as well as the paleo-altitude and paleo-location of the outcrop state for a subaqueous deposition by an unusually large catastrophic event. Regarding its age and the related uncertainties, it could be either an exceptional storm, or a landslide or impact triggered tsunami. The mega-tsunami of the 8000 BP collapse of the Valle del Bove valley (Etna Volcano) could be this event, for its matching age and calculated run-up height.

Burner balancing Salem Harbor Station

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

Sload, A.W.; Dube, R.J.

The traditional method of burner balancing is first to determine the fuel distribution, then to measure the economizer outlet excess oxygen distribution and to adjust the burners accordingly. Fuel distribution is typically measured by clean and dirty air probing. Coal pipe flow can then be adjusted, if necessary, through the use of coal pipe orificing or by other means. Primary air flow must be adjusted to meet the design criteria of the burner. Once coal pipe flow is balanced to within the desired criteria, secondary air flow to individual burners can be changed by adjusting windbox dampers, burner registers, shroudsmore » or other devices in the secondary air stream. This paper discusses problems encountered in measuring excess O{sub 2} at the economizer outlet. It is important to recognize that O{sub 2} measurements at the economizer outlet, by themselves, can be very misleading. If measurement problems are suspected or encountered, an alternate approach similar to that described should be considered. The alternate method is not only useful for burner balancing but also can be used to help in calibrating the plant excess O{sub 2} instruments and provide an on line means of cross-checking excess air measurements. Balanced burners operate closer to their design stoichiometry, providing better NO{sub x} reduction. For Salem Harbor Station, this means a significant saving in urea consumption.« less

Comparative assessment of heat stress induced changes in carcass traits, plasma leptin profile and skeletal muscle myostatin and HSP70 gene expression patterns between indigenous Osmanabadi and Salem Black goat breeds.

PubMed

Archana, P R; Sejian, V; Ruban, Wilfred; Bagath, M; Krishnan, G; Aleena, J; Manjunathareddy, G B; Beena, V; Bhatta, Raghavendra

2018-07-01

The primary objective of the study was to compare the impact of heat stress on meat production characteristics of Osmanabadi and Salem Black breed goats based on changes in carcass characteristics, meat quality attributes, plasma leptin concentration, skeletal muscle myostatin and heat shock protein 70 (HSP70) gene expression patterns. The goats were randomly distributed into four groups: OSC (n = 6; Osmanabadi Control), OSHS (n = 6; Osmanabadi Heat Stress), SBC (n = 6; Salem Black Control) and SBHS (n = 6; Salem Black Heat Stress). The animals were slaughtered at the end of the study and their meat characteristics were assessed. This study established the impact of heat stress on a wide variety of carcass and meat quality characteristics in OS and SB goat breeds. The results from the study also provided some crucial evidence for a better resilience capacity of Salem Black breed as compared to Osmanabadi goats in maintaining the meat production during heat stress. The study also established plasma leptin and HSP70 genes to be the ideal biomarkers to reflect the impact of heat stress on meat characteristics in indigenous goats. Copyright © 2018 Elsevier Ltd. All rights reserved.

Studies in Teaching: 2002 Research Digest. Research Projects Presented at Annual Research Forum (Winston-Salem, North Carolina, December 2002)

ERIC Educational Resources Information Center

McCoy, Leah P., Ed.

2002-01-01

This document presents the of an annual educational research forum held at Wake Forest University (Winston-Salem, North Carolina) on December 11, 2002. A table of contents and 27 research studies of high school teaching are included. Studies include: Effects of the Earth/Environmental Science Requirement on High School Science Enrollment in North…

Data Center Energy Efficiency and Renewable Energy Site Assessment: Anderson Readiness Center; Salem, Oregon

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

Metzger, I.; Van Geet, O.

This report summarizes the results from the data center energy efficiency and renewable energy site assessment conducted for the Oregon Army National Guard in Salem, Oregon. A team led by NREL conducted the assessment of the Anderson Readiness Center data centers March 18-20, 2014 as part of ongoing efforts to reduce energy use and incorporate renewable energy technologies where feasible. Although the data centers in this facility account for less than 5% of the total square footage, they are estimated to be responsible for 70% of the annual electricity consumption.

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

Studies in Teaching: 2016 Research Digest. Action Research Projects Presented at Annual Research Forum (Winston-Salem, North Carolina, Jun 30, 2016)

ERIC Educational Resources Information Center

McCoy, Leah P., Ed.

2016-01-01

This document presents the proceedings of the 21st Annual Research Forum held June 30, 2016, at Wake Forest University in Winston-Salem, North Carolina. Included are the following 11 action research papers: The Use of Mexican Folk Art to Develop Oral and Written Language Ability and Cultural Awareness in the Secondary Spanish Classroom (Antonio…

JEN-1 Reactor Control System; SISTEMA DE CONTROL DEL REACTOR JEN-1

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

Cantillo, M.F.; Nuno, C.M.; Andreu, J.L.M.

1963-01-01

ABS>The JEN-1 3Mw power swimming pool reactor electrical control circuits are described. Start-up, power generation in the core, and shutdown are controlled by the reactor control system. This control system guarantees in each moment the safety conditions during reactor operation. Each circuit was represented by a scheme, complemented with a description of its function, components, and operation theory. Components described include: scram circuit; fission counter control circuit; servo control circuit; control circuit of safety sheets; control circuits of primary, secondary, and clean-up pump motors and tower fan motor; primary valve motor circuit; center cubicle alarm circuit; and process alarm circuit.more » (auth)« less

The Salem Smart Power Center: An Assessment of Battery Performance and Economic Potential

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

Balducci, Patrick J.; Alam, M. J.E.; Viswanathan, Vilayanur

This paper presents an assessment of the economic potential of a 5 MW/1.25 MWh Energy Storage System (ESS) installed at the Salem Smart Power Center (SSPC), a smart grid technology demonstration facility owned and operated by Portland General Electric (PGE) in Salem, Oregon. The ESS and the grid conditions in which it operates were modeled using Pacific Northwest National Laboratory’s Battery Storage Evaluation Tool (BSET) to explore tradeoffs between services, and to develop optimal control strategies. This assessment monetized the value derived from nine services the SSPC could provide to PGE and the customers it serves. The ESS and themore » grid conditions in which it operates were modeled using PNNL’s in-house optimization tool BSET to explore tradeoffs between services, and to develop optimal control strategies. The analysis resulted in a number of lessons that provide crucial insights into the practical application of ESS, including; The SSPC, which was originally conceived as a research and test facility and built with the prevailing maturity technology level, was built at a cost ($20.4 million) that exceeds current day prices ($5.4 million) for a similarly designed and built 5 MW/1.25 MWh system; In terms of economic operation, the SSPC is currently underutilized, deployed only for primary frequency response. PNNL modeling indicates that optimal operation of the ESS could generate an additional value of $2.3 million over 20 years. It should also be noted that primary frequency response is the highest benefit application but requires a response from the SSPC only 17 hours each year. While optimally engaged, the ESS would provide arbitrage and ancillary services 78 percent of the time, but those services generate only 27 percent of the total value; Participation in Western EIM represents an interesting opportunity for PGE with a potential to generate $2.1 million value in PV terms over 20 years in the 5-min real-time market; With an energy to power

Compressed Air System Overhaul Improves Production at a Powdered Metal Manufacturing Plant (GKN Sinter Metals in Salem, IN)

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

None

In 1998, GKN Sinter Metals completed a successful compressed air system improvement project at its Salem, Indiana manufacturing facility. The project was performed after GKN undertook a survey of its system in order to solve air quality problems and to evaluate whether the capacity of their compressed air system would meet their anticipated plant expansion. Once the project was implemented, the plant was able to increase production by 31% without having to add any additional compressor capacity.

Groundwater quality assessment using geospatial and statistical tools in Salem District, Tamil Nadu, India

NASA Astrophysics Data System (ADS)

Arulbalaji, P.; Gurugnanam, B.

2017-10-01

The water quality study of Salem district, Tamil Nadu has been carried out to assess the water quality for domestic and irrigation purposes. For this purpose, 59 groundwater samples were collected and analyzed for pH, electrical conductivity (EC), total dissolved solids (TDS), major anions (HCO3 -, CO3 -, F-, Cl-, NO2 - + NO3 -, and SO4 2-), major cations (Ca2+ Mg2+, Na+, and K+), alkalinity (ALK), and hardness (HAR). To assess the water quality, the following chemical parameters were calculated based on the analytical results, such as Piper plot, water quality index (WQI), sodium adsorption ratio (SAR), magnesium hazard (MH), Kelly index (KI), and residual sodium carbonate (RSC). Wilcox diagram represents that 23% of the samples are excellent to good, 40% of the samples are good to permissible, 10% of the samples are permissible to doubtful, 24% of the samples are doubtful unsuitable, and only 3% of the samples are unsuitable for irrigation. SAR values shows that 52% of the samples indicate high-to-very high and low-to-medium alkali water. KI values indicate good quality (30%) and not suitable (70%) for irrigation purposes. RSC values indicate that 89% of samples are suitable for irrigation purposes. MH reveals that 17% suitable and 83% samples are not suitable for irrigation purposes and for domestic purposes the excellent (8%), good (48%), and poor (44%). The agricultural waste, fertilizer used, soil leaching, urban runoff, livestock waste, and sewages are the sources of poor water quality. Some samples are not suitable for irrigation purposes due to high salinity, hardness, and magnesium concentration. In general, the groundwater of the Salem district was polluted by agricultural activities, anthropogenic activities, ion exchange, and weathering.

Lithological mapping of Kanjamalai hill using hyperspectral remote sensing tools in Salem district, Tamil Nadu, India

NASA Astrophysics Data System (ADS)

Arulbalaji, Palanisamy; Balasubramanian, Gurugnanam

2017-07-01

This study uses advanced spaceborne thermal emission and reflection radiometer (ASTER) hyperspectral remote sensing techniques to discriminate rock types composing Kanjamalai hill located in the Salem district of Tamil Nadu, India. Kanjamalai hill is of particular interest because it contains economically viable iron ore deposits. ASTER hyperspectral data were subjected to principal component analysis (PCA), independent component analysis (ICA), and minimum noise fraction (MNF) to improve identification of lithologies remotely and to compare these digital data results with published geologic maps. Hyperspectral remote sensing analysis indicates that PCA (R∶G∶B=2∶1∶3), MNF (R∶G∶B=3∶2∶1), and ICA (R∶G∶B=1∶3∶2) provide the best band combination for effective discrimination of lithological rock types composing Kanjamalai hill. The remote sensing-derived lithological map compares favorably with a published geological map from Geological Survey of India and has been verified with ground truth field investigations. Therefore, ASTER data-based lithological mapping provides fast, cost-effective, and accurate geologic data useful for lithological discrimination and identification of ore deposits.

Three-dimensional facies architecture of the Salem Limestone (middle Mississippian), Eastern Margin of Illinois basin

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

Nadeem, A.; Keith, B.D.; Thompson, T.A.

Mapping of sedimentary surfaces in the Middle Mississippian Salem Limestone exposed on sawed quarry walls in south-central Indiana has revealed a hierarchy of depositional units representative of the extremely dynamic hydrographic regime of the upper shoreface zone. The depositional units on the scale of microform and mesoform are represented by the microfacies and the facies respectively. Based on their hierarchy, genetically related depositional units and associated bounding surfaces were grouped together to construct four architectural packages (APs) of the scale of mesoforms. AP-I is dominantly an echinoderm- and bryozoan-rich grainstone and consists of bedforms ranging from small ripples bounded bymore » first-order surfaces to two- and three- dimensional megaripples bounded by the second-order surfaces. It formed as part of a giant ramp (asymmetric wavefield) within the intrashoal channel setting. AP-II, also a skeletal grainstone, is a complex of giant sandwaves that moved into the area under the infulence of a storm and partly filled the basal channel form of AP-I. Large avalanche foresets with tangential toesets prevail. AP-III is a dark-gray spatially discontinuous skeletal grainstone to packstone that laterally grades into a skeletal packstone to wackestone. It locally developed overhangs, rips-ups, and hardground on its upper surface. AP-IV is a skeletal and oolitic grainstone formed of tabular two-dimensional megaripples (planar cross-beds) and three-dimensional oscillatory megaripples (trough cross-beds). These architectural packages based on the bedform architecture and micro-and mesoscale compositional changes can be used to characterize micro-, meso, and macroscale heterogeneities. Models of facies architecture from this and similar outcrop studies can be applied to the subsurface Salem reservoirs in the Illinois Basin using cores.« less

10 CFR 1.44 - Office of New Reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement...

10 CFR 1.44 - Office of New Reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement...

10 CFR 1.44 - Office of New Reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations...

10 CFR 1.44 - Office of New Reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 1 2014-01-01 2014-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations...

10 CFR 1.44 - Office of New Reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations...

Prevalence of Oral Mucosal Lesions among Dental Patients with Mixed Habits in Salem District - A Study

PubMed Central

Karthik, R.; Mohan, N.

2017-01-01

Background: The practice of betel nut chewing with or without tobacco is still practiced in south india, salem inspite of its harmful effects. Methodology: 200 Patients visiting the outpatient department, Oral medicine and radiology from Aug 2015 to Aug 2016. Result and Conclusion: In our study, 3 women were exclusively churut smokers. Thirty-eight percent of the dental patients were beedi smoker, 32% were tobacco chewers, 12% were both betel nut and tobacco chewers, 8% were exclusively betel nut chewers, 1% of the dental population were exclusively churut smokers. Mean age group of the study population is 50.2 (14.4). There are 28 females and 172 males in the study group. Chi-square test revealed a statistically significant difference (P = 0.001) between males and females based on soft-tissue findings and no statistically significant difference (P = 0.572) between males and females based on distribution of hard-tissue findings. PMID:29284937

78 FR 4465 - PPL Bell Bend, LLC; Combined License Application for Bell Bend Nuclear Power Plant; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-22

... License Application for Bell Bend Nuclear Power Plant; Exemption 1.0 Background PPL Bell Bend, LLC... Regulations (10 CFR), Subpart C of Part 52, ``Licenses, Certifications, and Approvals for Nuclear Power Plants.'' This reactor is to be identified as Bell Bend Nuclear Power Plant (BBNPP), in Salem County...

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 1 2014-01-01 2014-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

Simulated effects of allocated and projected 2025 withdrawals from the Potomac-Raritan-Magothy aquifer system, Gloucester and Northeastern Salem Counties, New Jersey

USGS Publications Warehouse

Charles, Emmanuel; Nawyn, John P.; Voronin, Lois M.; Gordon, Alison D.

2011-01-01

-Raritan-Magothy aquifers are 49, 43, and 48 feet below NGVD 29, respectively, which are 18, 16, and 18 feet lower, respectively, than in the 2000 baseline simulation. The lowest simulated water levels are 156, 95, and 69 feet below NGVD 29, respectively, which are 79, 30, and 10 feet lower, respectively, than in the 2000 baseline simulation. Simulated net flow from the Potomac-Raritan-Magothy aquifer system to streams is 8,441 Mgal/yr in the 2000 baseline simulation but is 6,018 Mgal/yr in the full-allocation scenario, a decrease of 29 percent from the 2000 baseline simulation. Simulated net flow in the 2000 baseline simulation is 1,183 Mgal/yr from the aquifer system to the Delaware River but in the full-allocation scenario is 1,816 Mgal/yr from the river to the aquifer system. Four other simulations were conducted that incorporated full-allocation conditions at water-purveyor wells in Critical Area 2 but increased or decreased withdrawals at selected water-purveyor wells outside Critical Area 2 and agricultural-irrigation and industrial-self-supply wells in the study area. The results of the four simulations also indicate net flow from the Delaware River to the Potomac-Raritan-Magothy aquifer system. A growth scenario was developed to simulate future withdrawals in 2025 estimated from population projections for municipalities in the Salem-Gloucester study area. Simulated withdrawals for this scenario totaled 10,261 Mgal/yr, an increase of 6 percent from the 2000 baseline simulation. This total includes about 25 Mgal/yr withdrawn from the Englishtown aquifer system for domestic self-supply. This scenario incorporated full-allocation withdrawals at water-purveyor wells in Critical Area 2, and increased withdrawals at water-purveyor wells outside Critical Area 2. Results of this simulation indicate that the average simulated water levels in the Upper, Middle, and Lower Potomac-Raritan-Magothy aquifers are 32, 29, and 32 feet below NGVD 29, respectively, which are 1, 2, and

40 CFR 52.1570 - Identification of plan.

Code of Federal Regulations, 2010 CFR

2010-07-01

... County 2.0 E. I. du Pont de Nemours & Co Deepwater, Salem County 1.5 Heinz-USA Salem City, Salem County 2... County 2.0 Atlantic City Electric Deepwater Station Penns Grove, Salem County 1.5 E. I. du Pont de... County 2.0 E. I. du Pont de Nemours & Co Deepwater, Salem County 1.5 Heinz-U.S.A Salem City, Salem County...

10 CFR 1.13 - Advisory Committee on Reactor Safeguards.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Advisory Committee on Reactor Safeguards. 1.13 Section 1... Headquarters Panels, Boards, and Committees § 1.13 Advisory Committee on Reactor Safeguards. The Advisory Committee on Reactor Safeguards (ACRS) was established by section 29 of the Atomic Energy Act of 1954, as...

10 CFR 1.13 - Advisory Committee on Reactor Safeguards.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Advisory Committee on Reactor Safeguards. 1.13 Section 1... Headquarters Panels, Boards, and Committees § 1.13 Advisory Committee on Reactor Safeguards. The Advisory Committee on Reactor Safeguards (ACRS) was established by section 29 of the Atomic Energy Act of 1954, as...

Microbial Beneficiation of Salem Iron Ore Using Penicillium purpurogenum

NASA Astrophysics Data System (ADS)

Mishra, M.; Pradhan, M.; Sukla, L. B.; Mishra, B. K.

2011-02-01

High alumina and silica content in the iron ore affects coke rate, reducibility, and productivity in a blast furnace. Iron ore is being beneficiated all around the world to meet the quality requirement of iron and steel industries. Choosing a beneficiation treatment depends on the nature of the gangue present and its association with the ore structure. The advanced physicochemical methods used for the beneficiation of iron ore are generally unfriendly to the environment. Biobeneficiation is considered to be ecofriendly, promising, and revolutionary solutions to these problems. A characterization study of Salem iron ore indicates that the major iron-bearing minerals are hematite, magnetite, and goethite. Samples on average contains (pct) Fe2O3-84.40, Fe (total)-59.02, Al2O3-7.18, and SiO2-7.53. Penicillium purpurogenum (MTCC 7356) was used for the experiment . It removed 35.22 pct alumina and 39.41 pct silica in 30 days in a shake flask at 10 pct pulp density, 308 K (35 °C), and 150 rpm. In a bioreactor experiment at 2 kg scale using the same organism, it removed 23.33 pct alumina and 30.54 pct silica in 30 days at 300 rpm agitation and 2 to 3 l/min aeration. Alumina and silica dissolution follow the shrinking core model for both shake flask and bioreactor experiments.

Cover sequence stratigraphy and structure: Salem Church basement culmination, Georgia Blue Ridge

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

Li, L.; Tull, J.F.

The Salem Church anticline SW of Jasper, Georgia in the western Blue Ridge is roughly an oval shaped structural dome with its long axis trending NE-SW. The anticline is cored by the Grenville age Corbin Gneiss which represents allochthonous North American basement. In debate for decades has been the age and origin of several kilometers of poly-deformed cover sequence rocks which were metamorphosed to greenschist facies and were probably transported over a long distance inland after their deposition. The stratigraphy of the cover sequence exhibits rapid lithofacies changes. At most localities, the basement is overlain by a 500--600 m thickmore » coarse clastic unit sourced from the basement rocks, composed mainly of metaconglomerate, metasandstone and metadiamictite. A thin unit less than 20 m thick of sericite phyllite occurs between the basement and the coarse clastic unit along the SE limb of the anticline but pinches out to the MW. A relatively sharp stratigraphic contact occurs between quartzite unit and overlying dark colored metagreywackes and metadiamictites containing distinctive cobbles and boulders of granitic and gneissic basement rocks up to 1 meter in length. This unit is about 100 m thick in the SW but thins rapidly towards the NE. It grades up into a geographically widespread graphitic phyllite which encircles most of the anticline. Unlike the cover sequence above the corbin basement west of Waleska, Georgia, no carbonate is found in this area.« less

The Association between Cariogenic Factors and the Occurrence of Early Childhood Caries in Children from Salem District of India.

PubMed

Stephen, Arokiaraj; Krishnan, Ramesh; Chalakkal, Paul

2017-07-01

Various aetiological factors for Early Childhood Caries (ECC) have been suggested in literature, but the role of each factor has been disputed in various studies. To evaluate the association between ECC and its related factors in children between three to six years of age, from rural, semi-urban and urban schools in the district of Salem (India). The study consisted of children between three to six years of age, selected randomly from schools in semi-urban, urban and rural schools at Salem. Questionnaire (comprising of the factors associated with ECC) and parental consent forms were sent to each of the parents of the school children. Filled questionnaires and parental consent forms were obtained from a total of 2771 children, out of which 1771 were boys and 1000 were girls. Clinical examination for ECC was carried out according to the WHO criteria, by a single examiner and an assistant to eliminate inter-examiner variability. Statistical analysis was carried out using the student's independent t-test and ANOVA (one way analysis of variance). Out of a total of 2771 children examined, only 443 (315 boys and 128 girls) children were found to have ECC (16 %). Significant correlations were found between the occurrence of ECC and various factors like age of parents, number of siblings, on-demand breast feeding and bottle feeding habits and sweetened pacifier use. However, factors such as age and gender of the child, type of birth and duration of pregnancy, had no significant correlation with the occurrence of ECC. The need for educating parents on the ill effects of improper breast feeding and bottle feeding habits is important.

Hydrogeology of, water withdrawal from, and water levels and chloride concentrations in the major Coastal Plain aquifers of Gloucester and Salem Counties, New Jersey

USGS Publications Warehouse

Cauller, S.J.; Carleton, G.B.; Storck, M.J.

1999-01-01

Eight aquifers underlying Gloucester and Salem Counties in the southwestern Coastal Plain of New Jersey provide nearly all the drinking water for the 295,000 people who live in the area. Ground-water withdrawals in the two-county area and adjoining counties have affected water levels in several of these aquifers. Ground-water withdrawals in the two-county area also have affected the quality of water, increasing the chloride concentration in several of the aquifers as a result of saltwater intrusion. This report contains hydrologic data from the two-county area, including geometry and extent of hydrogeologic units, thickness and altitude of each aquifer, withdrawals from and water levels in major aquifers, and chloride concentrations in water from each aquifer. Reported ground-water withdrawals in Gloucester and Salem Counties during 1975-95 averaged 7,800 Mgal/yr (million gallons per year) for public supply, 4,900 Mgal/yr for industrial use, 700 Mgal/yr for irrigation, 500 Mgal/yr for power plants, 50 Mgal/yr for commercial use, and about 40 Mgal/yr for mining. Withdrawals for domestic self-supply in 1994 are estimated to be about 2,600 Mgal/yr, but only about 20 percent (520 Mgal/yr) is thought to be consumptive use; the remainder is returned to the aquifer through septic systems. The most heavily used aquifer in Salem and Gloucester Counties is the Upper Potomac-Raritan-Magothy aquifer, followed by, in decreasing order of use, the Middle Potomac-Raritan-Magothy aquifer, the Lower Potomac-Raritan-Magothy aquifer, the Kirkwood-Cohansey aquifer system, and the Wenonah-Mount Laurel aquifer. Reported withdrawals from these aquifers during 1975-95 averaged 5,000, 3,700, 3,200, and 330 Mgal/yr, respectively. Withdrawals from the Wenonah-Mount Laurel aquifer in Gloucester County increased during 1993-96 because of New Jersey Department of Environmental Protection restrictions on new withdrawals from the deeper Potomac-Raritan-Magothy aquifer system. Because of the

Groundwater resource exploration in Salem district, Tamil Nadu using GIS and remote sensing

NASA Astrophysics Data System (ADS)

Maheswaran, G.; Selvarani, A. Geetha; Elangovan, K.

2016-03-01

Since last decade, the value per barrel of potable groundwater has outpaced the value of a barrel of oil in many areas of the world. Hence, proper assessment of groundwater potential and management practices are the needs of the day. Establishing relationship between remote sensing data and hydrologic phenomenon can maximize the efficiency of water resources development projects. Present study focuses on groundwater potential assessment in Salem district, Tamil Nadu to investigate groundwater resource potential. At the same, all thematic layers important from ground water occurrence and movement point of view were digitized and integrated in the GIS environment. The weights of different parameters/themes were computed using weighed index overlay analysis (WIOA), analytic hierarchy process (AHP) and fuzzy logic technique. Through this integrated GIS analysis, groundwater prospect map of the study area was prepared qualitatively. Field verification at observation wells was used to verify identified potential zones and depth of water measured at observation wells. Generated map from weighed overlay using AHP performed very well in predicting the groundwater surface and hence this methodology proves to be a promising tool for future.

Integrated NO{sub x} control at New England Power, Salem Harbor Station

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

Frish, M.B.; Johnson, S.A.; Comer, J.P.

Selective non-catalytic reduction (SNCR) is a viable technology for reducing NO{sub x} emissions from coal-fired boilers, especially those older units where large capital expenditures for alternative technologies may not be justified. However, NO{sub x} reduction efficiency of the SNCR process is maximized when the proper amount of reagent is injected at the proper temperature and dispersed rapidly enough to avoid ammonia slip. Early NEP experience at Salem Harbor station indicated that NO{sub x} reductions of 60% were achievable with SNCR. However, less NO{sub x} reductions were tolerated to avoid NH{sub 3} slip and subsequent flyash contamination and visible stack plumemore » resulting from excess ammonia. Preliminary tests by PSI Environmental showed that ammonia slip could be monitored in real time using their patented SpectraScan{trademark}-NH{sub 3} instrument, and that furnace exit temperature could be continuously monitored and controlled using GasTemp{trademark} another PSI Environmental product. Based on this information, detailed tests were planned to show integrated control over the SNCR process. A goal of the project was to achieve lower NO{sub x} with less reagent! This paper describes the status of the project.« less

Equipment for neutron measurements at VR-1 Sparrow training reactor.

PubMed

Kolros, Antonin; Huml, Ondrej; Kríz, Martin; Kos, Josef

2010-01-01

The VR-1 sparrow reactor is an experimental nuclear facility for training, student education and teaching purposes. The sparrow reactor is an educational platform for the basic experiments at the reactor physic and dosimetry. The aim of this article is to describe the new experimental equipment EMK310 features and possibilities for neutron detection by different gas filled detectors at VR-1 reactor. Among the EMK310 equipment typical attributes belong precise set-up, simple control, resistance to electromagnetic interference, high throughput (counting rate), versatility and remote controllability. The methods for non-linearity correction of pulse neutron detection system and reactimeter application are presented. Copyright 2009. Published by Elsevier Ltd.

77 FR 12366 - JP Rail, Inc. d/b/a Southern Railroad Company of New Jersey-Operation Exemption-Rail Line in...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-29

.../a Southern Railroad Company of New Jersey-- Operation Exemption--Rail Line in Salem County, N.J. JP... known as the Salem Branch Line (the Line), which is owned by Salem County, N.J. (Salem), as follows: (1....J., and approximately milepost 28.4 at Salem, and (2) the Glass House Spur, between mileposts 0.0...

Low-NOx burner and SNCR retrofit experience at New England Power Salem Harbor Station

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

Quartucy, G.; Sload, A.; Fynan, G.

New England Power has recently installed Riley-Stoker low-NO{sub x} burners (LNB) and Nalco Fuel Tech urea-based selective non-catalytic NO{sub x} reduction (SNCR) systems on Units 1 and 3 at its Salem Harbor generating station. In addition, Unit 3 was also retrofit with a two-level overfire air (OFA) system. These two coal-fired units are front wall-fired with unequal burner spacing and have uncontrolled full-load NO{sub x} emissions of nominally 750 ppm (1.1 lb/MMBtu). Unit 1 is rated at 86 MW and has 12 burners, while Unit 3 is rated at 155 MW and has 16 burners. NO{sub x} reduction performance ofmore » the LNB, OFA and SNCR systems has been characterized both independently and in combination during the test programs while firing low-sulfur coals. Unit 1 tests showed that the LNBs provided NO{sub x} reductions of approximately 50 percent at loads above 60 MW using narrow angle coal spreaders. Corresponding ash carbon at these NO{sub x} levels varied between 16 and 35 percent. The SNCR system provided an additional 40 percent NO{sub x} reduction from the LNB baseline at a molar N/NO ratio of 1.2. The corresponding NH{sub 3} slip levels were less than 10 ppm. On Unit 3, LNB tests showed that NO{sub x} reductions of nominally 10 percent were achieved with the burners alone, using wide angle coal spreaders. The use of OFA, at design levels, provided additional NO{sub x} reductions ranging from 42 percent at full load to 4 percent a minimum load relative to the LNB baseline. Ash carbon levels doubled to levels above 30 percent when the OFA system was operated at design conditions at loads above 110 MW. The SNCR system provided NO{sub x} reductions of 33 percent relative to the LNB/OFA baseline of 0.55 lb/MMBtu, at a molar N/NO ratio of 1.3. Ammonia slip for these conditions was less than 5 ppm.« less

SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

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

None

2013-09-25

U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in amore » remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.« less

SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

ScienceCinema

None

2018-01-16

U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

Anatomy of the grainstone shoal facies of the Salem Limestone (Mississippian) of southern Indiana

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

Dodd, J.R.; Petzold, D.D.; Thompson, T.A.

In 1990 M.A. Brown described the middle Mississippian (Valmeyeran) Salem Limestone exposed on the eastern side of the Illinois Basin as consisting of a massive grainstone shoal facies behind which developed a sand flat, an open lagoon, and a restricted lagoon facies. Smaller intrashoal channels provided limited exchange between lagoon and open ocean. The authors have made detailed studies of sedimentary structures and petrography of the shoal facies in three settings: the shoal proper, an intrashoal channel, and an intershoal channel. The shoal and channel facies consists of tabular-planar and trough cross-stratified beds of grainstone containing echinoderm and fenestrate bryozoanmore » grains as their primary constituents. Prominent hardgrounds that have up to 1 m of erosional relief occur in two of the sections. Despite the apparent uniformity of composition of the shoal, porosity and especially permeability varies over a wide range, suggesting a range of cementation patterns within the shoal. Most of the cement in the shoal consists of syntaxial overgrowths on echinoderm grains. Cementation is less and thus porosity and permeability greater, in portions of the shoals containing a lower concentration of echinoderm grains and grains with thick micrite envelopes. However, some portions of the intershoal channel facies that contain a high percentage of ooids have reduced porosity and permeability due to crushing of ooids, producing micrite that clogged the pores. Primary sedimentary features of the shoal facies were produced predominantly by storm reworking of carbonate grains produced in situ and perhaps in part washed in from surrounding environments.« less

Waterborne Commerce of the United States, Calendar Year 1982. Part 1. Waterways and Harbors, Atlantic Coast.

DTIC Science & Technology

1984-07-01

TOTAL TON-HILES, 8,4110.310. SALEM HAR304, "A38. SECTION INCLUI3ED, WESTERLY SHORE OF MA44514 INCLUDINJG SOUTH RIVER. CnNTRT)LL 14...444074 400 eoToo COAO00LO, 10 40lOI O0 , PFA , ,1 6 FOOT 0’) £OCfAGF A0O,. ALL AT OAN I W aATEO* PPTJECT 10E4004 1.9 k )l.A rqOvAOAyTVf STATFOFNT mo...OLOWANS CREEK, N. J. UP~okiNfl moovnPLINA TOTAL, 6 AND LESS - 9 PEPPER CREEK, DEL. NOl WE SiL 10RIP REPORT1 .U SALEM RI VER, I..* uPON 0 oINPOUND TOT ALI

NASA Reactor Facility Hazards Summary. Volume 1

NASA Technical Reports Server (NTRS)

1959-01-01

The Lewis Research Center of the National Aeronautics and Space Administration proposes to build a nuclear research reactor which will be located in the Plum Brook Ordnance Works near Sandusky, Ohio. The purpose of this report is to inform the Advisory Committee on Reactor Safeguards of the U. S. Atomic Energy Commission in regard to the design Lq of the reactor facility, the characteristics of the site, and the hazards of operation at this location. The purpose of this research reactor is to make pumped loop studies of aircraft reactor fuel elements and other reactor components, radiation effects studies on aircraft reactor materials and equipment, shielding studies, and nuclear and solid state physics experiments. The reactor is light water cooled and moderated of the MTR-type with a primary beryllium reflector and a secondary water reflector. The core initially will be a 3 by 9 array of MTR-type fuel elements and is designed for operation up to a power of 60 megawatts. The reactor facility is described in general terms. This is followed by a discussion of the nuclear characteristics and performance of the reactor. Then details of the reactor control system are discussed. A summary of the site characteristics is then presented followed by a discussion of the larger type of experiments which may eventually be operated in this facility. The considerations for normal operation are concluded with a proposed method of handling fuel elements and radioactive wastes. The potential hazards involved with failures or malfunctions of this facility are considered in some detail. These are examined first from the standpoint of preventing them or minimizing their effects and second from the standpoint of what effect they might have on the reactor facility staff and the surrounding population. The most essential feature of the design for location at the proposed site is containment of the maximum credible accident.

52. ARAII. Support piers for SL1 reactor building. September 5, ...

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

52. ARA-II. Support piers for SL-1 reactor building. September 5, 1957. Ineel photo no. 57-4398. Photographer: Ken Mansfield. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

65. ARAII. Interior view of SL1 reactor building control piping ...

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

65. ARA-II. Interior view of SL-1 reactor building control piping for water purification system. On operating floor of building. March 21, 1958. Ineel photo no. 58-1360. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Determine Operating Reactor to Use for the 2016 PCI Level 1 Milestone

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

Clarno, Kevin T.

2016-01-30

The Consortium for Advanced Simulation of Light Water Reactors (LWRs) (CASL) Level 1 milestone to “Assess the analysis capability for core-wide [pressurized water reactor] PWR Pellet- Clad Interaction (PCI) screening and demonstrate detailed 3-D analysis on selected sub-region” (L1:CASL.P13.03) requires a particular type of nuclear power plant for the assessment. This report documents the operating reactor and cycles chosen for this assessment in completion of the physics integration (PHI) milestone to “Determine Operating Reactor to use for PCI L1 Milestone” (L3:PHI.CMD.P12.02). Watts Bar Unit 1 experienced (at least) one fuel rod failure in each of cycles 6 and 7, andmore » at least one was deemed to be duty related rather than being primarily related to a manufacturing defect or grid effects. This brief report documents that the data required to model cycles 1–12 of Watts Bar Unit 1 using VERA-CS contains sufficient data to model the PHI portion of the PCI challenge problem. A list of additional data needs is also provided that will be important for verification and validation of the BISON results.« less

A Capacitated Facility Location Approach for the Tanker Employment Problem

DTIC Science & Technology

2005-03-01

first demonstrated air refueling with an in-flight hose contact between two De Havilland DH -4B aircraft (Ianuzzi, 1997a: 22). The Army Air corps...Tool" ElseIf LongEast = True Then East = True 106 Else: East = False End If ’collect info dump into excel Application.ScreenUpdating...ALI AL SALEM AB 31 1 F16_10 F16 4 1 106 ALI AL SALEM AB 39 29 F16_10 F16 4 2 107 ALI AL SALEM AB 12 50 F16_10 F16 4 3 108 ALI AL SALEM AB 34 5

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...

Neutronic calculation of fast reactors by the EUCLID/V1 integrated code

NASA Astrophysics Data System (ADS)

Koltashev, D. A.; Stakhanova, A. A.

2017-01-01

This article considers neutronic calculation of a fast-neutron lead-cooled reactor BREST-OD-300 by the EUCLID/V1 integrated code. The main goal of development and application of integrated codes is a nuclear power plant safety justification. EUCLID/V1 is integrated code designed for coupled neutronics, thermomechanical and thermohydraulic fast reactor calculations under normal and abnormal operating conditions. EUCLID/V1 code is being developed in the Nuclear Safety Institute of the Russian Academy of Sciences. The integrated code has a modular structure and consists of three main modules: thermohydraulic module HYDRA-IBRAE/LM/V1, thermomechanical module BERKUT and neutronic module DN3D. In addition, the integrated code includes databases with fuel, coolant and structural materials properties. Neutronic module DN3D provides full-scale simulation of neutronic processes in fast reactors. Heat sources distribution, control rods movement, reactivity level changes and other processes can be simulated. Neutron transport equation in multigroup diffusion approximation is solved. This paper contains some calculations implemented as a part of EUCLID/V1 code validation. A fast-neutron lead-cooled reactor BREST-OD-300 transient simulation (fuel assembly floating, decompression of passive feedback system channel) and cross-validation with MCU-FR code results are presented in this paper. The calculations demonstrate EUCLID/V1 code application for BREST-OD-300 simulating and safety justification.

Multi-faceted monitoring of estuarine turbidity and particulate matter provenance: Case study from Salem Harbor, USA.

PubMed

Hubeny, J Bradford; Kenney, Melanie; Warren, Barbara; Louisos, Jeremy

2017-01-01

Turbidity is a water quality parameter that is known to adversely affect aquatic systems, however the causes of turbid water are often elusive. We present results of a study designed to constrain the source of particulate matter in a coastal embayment that has suffered from increased turbidity over past decades. Our approach utilized monitoring buoys to quantify turbidity at high temporal resolution complemented by geochemical isotope analysis of suspended sediment samples and meteorological data. Results reveal a complex system in which multiple sources are associated with particulate matter. Weight of evidence demonstrates that phytoplankton productivity in the water column, however, is the dominant source of particulate matter associated with elevated turbidity in Salem Harbor, Massachusetts. Allochthonous matter from the watershed was observed to mix into the pool of suspended particulate matter near river mouths, especially in spring and summer. Resuspension of harbor surface sediments likely provides additional particulates in the regions of boat moorings, especially during summer when recreational boats are attached to moorings. Our approach allows us to constrain the causes of turbidity events in this embayment, is helping with conservation efforts of environmental quality in the region, and can be used as a template for other locations. Copyright © 2016 Elsevier B.V. All rights reserved.

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.

Future Reactor Neutrino Experiments (RRNOLD)1

NASA Astrophysics Data System (ADS)

Jaffe, David E.

The prospects for future reactor neutrino experiments that would use tens of kilotons of liquid scintillator with a ∼ 50 km baseline are discussed. These experiments are generically dubbed "RRNOLD" for Radical Reactor Neutrino Oscillation Liquid scintillator Detector experiment. Such experiments are designed to resolve the neutrino mass hierarchy and make sub-percent measurements sin2θ12, Δm232 and Δm122 . RRNOLD would also be sensitive to neutrinos from other sources and have notable sensitivity to proton decay.

THE COOLING REQUIREMENTS AND PROCESS SYSTEMS OF THE SOUTH AFRICAN RESEARCH REACTOR, SAFARI 1

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

Colley, J.R.

1962-12-01

The SAFARI 1 research reactor is cooled and moderated by light water. There are three process systems, a primary water system which cools the reactor core and surroundings, a pool water system, and a secondary water system which removes the heat from the primary and pool systems. The cooling requirements for the reactor core and experimental facilities are outlined, and the cooling and purification functions of the three process systems are described. (auth)

TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

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

Ryan, B.C.

1997-05-01

This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely themore » total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.« less

Groundwater Assessment Study for 50 Communities in Southeastern New Hampshire. Appendixes

DTIC Science & Technology

1980-09-01

U.S. Dept. of Agriculture, Soil Conservation Service, 1973. 3. Groundwater Favorability Map of the Salem -Plaistow Area, New Hampshire, by James M...QUADRANGLE LOCATION AQUIFER QUADRANGLE LOCATION At-i Haverhill (7.5’ series) Ex-i Exeter Salem Depot Ba-l Dover West Fr-l Haverhill (15’ series) Mt...Qenerall? are ice contact, Ki-l and Ne-2 are generally stratified sand and silt outwash tgpe deposits B. POTENTIAL 1. WIEGLE: ( Salem -Plaistow--defines

Advance High Temperature Inspection Capabilities for Small Modular Reactors: Part 1 - Ultrasonics

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

Bond, Leonard J.; Bowler, John R.

The project objective was to investigate the development non-destructive evaluation techniques for advanced small modular reactors (aSMR), where the research sought to provide key enabling inspection technologies needed to support the design and maintenance of reactor component performance. The project tasks for the development of inspection techniques to be applied to small modular reactor are being addressed through two related activities. The first is focused on high temperature ultrasonic transducers development (this report Part 1) and the second is focused on an advanced eddy current inspection capability (Part 2). For both inspection techniques the primary aim is to develop in-servicemore » inspection techniques that can be carried out under standby condition in a fast reactor at a temperature of approximately 250°C in the presence of liquid sodium. The piezoelectric material and the bonding between layers have been recognized as key factors fundamental for development of robust ultrasonic transducers. Dielectric constant characterization of bismuth scantanate-lead titanate ((1-x)BiScO 3-xPbTiO 3) (BS-PT) has shown a high Curie temperature in excess of 450°C , suitable for hot stand-by inspection in liquid metal reactors. High temperature pulse-echo contact measurements have been performed with BS-PT bonded to 12.5 mm thick 1018-low carbon steel plate from 20C up to 260 C. High temperature air-backed immersion transducers have been developed with BS-PT, high temperature epoxy and quarter wavlength nickel plate, needed for wetting ability in liquid sodium. Ultrasonic immersion measurements have been performed in water up to 92C and in silicone oil up to 140C. Physics based models have been validated with room temperature experimental data with benchmark artifical defects.« less

Geospatial tool-based morphometric analysis using SRTM data in Sarabanga Watershed, Cauvery River, Salem district, Tamil Nadu, India

NASA Astrophysics Data System (ADS)

Arulbalaji, P.; Gurugnanam, B.

2017-11-01

A morphometric analysis of Sarabanga watershed in Salem district has been chosen for the present study. Geospatial tools, such as remote sensing and GIS, are utilized for the extraction of river basin and its drainage networks. The Shuttle Radar Topographic Mission (SRTM-30 m resolution) data have been used for morphometric analysis and evaluating various morphometric parameters. The morphometric parameters of Sarabanga watershed have been analyzed and evaluated by pioneer methods, such as Horton and Strahler. The dendritic type of drainage pattern is draining the Sarabanga watershed, which indicates that lithology and gentle slope category is controlling the study area. The Sarabanga watershed is covered an area of 1208 km2. The slope of the watershed is various from 10 to 40% and which is controlled by lithology of the watershed. The bifurcation ratio ranges from 3 to 4.66 indicating the influence of geological structure and suffered more structural disturbances. The form factor indicates elongated shape of the study area. The total stream length and area of watershed indicate that mean annual rainfall runoff is relatively moderate. The basin relief expressed that watershed has relatively high denudation rates. The drainage density of the watershed is low indicating that infiltration is more dominant. The ruggedness number shows the peak discharges that are likely to be relatively higher. The present study is very useful to plan the watershed management.

Successful multi-technology NO{sub x} reduction project experience at New England Power Company - Salem Harbor station

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

Fynan, G.A.; Sload, A.; Adamson, E.J.

This paper presents the successes and lessons learned during recent low NOx burner and SNCR projects on generating units at New England Power`s Salem Harbor Generating Station. The principals involved in the project were New England Power Company, New England Power Service Company, Stone and Webster Engineering Corp. and Deutsche-Babcock Riley Inc. One unit was retrofitted with 16 Riley CCV burners with an OFA system, the other with 12 low NOx burners only. In addition to the burners, a SNCR system was also installed on three units. Since each of the burner systems are interdependent (SNCR was treated separately duringmore » design phases and optimized along with the burner systems), close cooperation during the design stages was essential to ensuring a successful installation, startup and optimization. This paper will present the coordinated effort put forth by each company toward this goal with the hope of assisting others who may be planning a similar effort. A summary of the operating results will also be presented. The up front teamwork and advance planning that went into the design stages of the project resulted in a number of successful outcomes e.g. scanner reliability, properly operating oil supply system, compatibility of burners and burner front oil system with new Burner Management System (BMS), reliable first attempt burner ignition and more. Advance planning facilitated pre-outage work and factored into keeping schedules and budgets on track.« less

153. ARAIII Reactor building (ARA608) Foundation plan. Aerojetgeneral 880area/GCRE608S1. Date: ...

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

153. ARA-III Reactor building (ARA-608) Foundation plan. Aerojet-general 880-area/GCRE-608-S-1. Date: February 1958. Ineel index code no. 063-0608-60-013-102653. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

COMMODITY SCALE SYNTHESIS OF 1-METHYLIMIDAZOLE BASED IONIC LIQUIDS USING A SPINNING TUBE-IN-TUBE REACTOR

EPA Science Inventory

The continuous large-scale preparation of several 1-methylimidazole based ionic liquids was carried out using a Spinning Tube-in-Tube (STT) reactor (manufactured by Kreido Laboratories). This reactor, which embodies and facilitates the use of Green Chemistry principles and Proce...

Health Physics State of the RA-1 Reactor, Period Between 1-1-62 and 10-8- 62. Report No. 87; ESTADO RADIOSANIT ARIO DEL REACTOR R.A. 1, PERIODO COMPRENDIDO ENTRE EL 1-1-62 Y EL 10-8-62. Informe No. 87

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

Gaspar, R.; Moll, O.; Hermelo, C.

1963-01-01

The methods used to measure the irradiation levels and the radiation exposure of personnel of the RA-1 reactor are described. The criteria used to evaluate the risks from this exposure are reported. Typical graphs are shown of the radiation levels measured in the control room. (J.S.R.)

Reversal of OFI and CHF in Research Reactors Operating at 1 to 50 Bar. Version 1.0

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

Kalimullah, M.; Olson, A. P.; Dionne, B.

2014-02-28

The conditions at which the critical heat flux (CHF) and the heat flux at the onset of Ledinegg flow instability (OFI) are equal, are determined for a coolant channel with uniform heat flux as a function of five independent parameters: the channel exit pressure (P), heated length (Lh) , heated diameter (Dh), inlet temperature (Tin), and mass flux (G). A diagram is made by plotting the mass flux and heat flux at the OFI-CHF intersection (reversal from CHF > OFI to CHF < OFI as G increases) as a function of P (1 to 50 bar), for 36 combinations ofmore » the remaining three parameters (Lh , Dh , Tin): Lh = 0.28, 0.61, 1.18 m; Dh = 3, 4, 6, 8 mm; Tin = 30, 50, 70 °C. The use of the diagram to scope whether a research reactor is OFI-limited (below the curve) or CHF-limited based on the five parameters of its coolant channel is described. Justification for application of the diagram to research reactors with axially non-uniform heat flux is provided. Due to its limitations (uncertainties not included), the diagram cannot replace the detailed thermal-hydraulic analysis required for a reactor safety analysis. In order to make the OFI-CHF intersection diagram, two world-class CHF prediction methods (the Hall-Mudawar correlation and the extended Groeneveld 2006 table) are compared for 216 combinations of the five independent parameters. The two widely used OFI correlations (the Saha- Zuber and the Whittle-Forgan with η = 32.5) are also compared for the same combinations of the five parameters. The extended Groeneveld table and the Whittle-Forgan OFI correlation are selected for use in making the diagram. Using the above five design parameters, a research reactor can be represented by a point on the reversal diagram, and the diagram can be used to scope, without a thermal-hydraulic calculation, whether the OFI will occur before the CHF, or the CHF will occur before the OFI when the reactor power is increased keeping the five parameters fixed.« less

75 FR 60778 - Announcement of Funding Awards for Fiscal Year 2010 Historically Black Colleges and Universities...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-10-01

..., and Grant Amount Region IV 1. Winston-Salem State University, Ms. Valerie Howard, Winston Salem State... Awards for Fiscal Year 2010 Historically Black Colleges and Universities Program AGENCY: Office of the... Historically Black Colleges and Universities Program. The purpose of this document is to announce the names...

Superfund record of decision (EPA region 10): Idaho National Engineering Lab, (USDOE) Operable Unit 26 (Stationary Low-Power Reactor-1 and Boiling Water Reactor Experiment-I Burial Grounds), Idaho Falls, ID, December 1, 1995

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

NONE

1997-03-01

This document presents the selected remedial action for the Stationary Low-Power Reactor-1 (SL-1) burial ground, the Boiling Water Reactor Experiment-I (BORAX-I) burial ground, and 10 no action sites in Waste Area Group 5. Actual or threatened releases of hazardous substances from the SL-1 and BORAX-I burial grounds, if not addressed by implementing the response action selected in this Record of Decision, may present a current or potential threat to public health, welfare, or the environment. The 10 no action sites do not present a threat to human health or the environment.

Evidence of elevated pressure and temperature during burial of the Salem Limestone in south-central Indiana, USA, and its implications for surprisingly deep burial

NASA Astrophysics Data System (ADS)

Ambers, Clifford P.

2001-09-01

A minor, normal fault related to compaction of the grainstone shoal facies of the Salem Limestone in south-central Indiana provides an unusual opportunity to test the pressure and temperature of both faulting and associated stylolitization. Syn-deformational sphalerite occurs in voids along the fault where it intersects an organic-rich shale parting in the sand flat facies overlying the grainstone. The sphalerite contains fluid inclusions that can be used for microthermobaric measurements. Most fluid inclusions in the sphalerite are demonstrably cogenetic with the host sphalerite and of the two-phase aqueous type common in Indiana, although many contain petroleum and others contain gas. Crushing tests in kerosene indicate that the aqueous inclusions contain dissolved methane in varying amounts as high as 1000 ppm. Microthermometry shows that late sphalerite growth, late fault movement, and late stylolitization all occurred as conditions approached 108°C and 292 bars. This pressure is in accord with a normal, basinal, geothermal gradient of 32.5°C/km that would produce the observed temperature under hydrostatic conditions at a burial depth of 2.7 km using an average fluid density of 1.1 g/cm 3. These results serve as a reminder that fluid inclusions in diagenetic minerals hold important temperature and pressure information regarding burial diagenesis of Paleozoic rocks across the North American midcontinent. Detailed study of dissolved gases in fluids trapped in disseminated sphalerite that is common across the midcontinent could help resolve the enigma of sedimentary rocks with high thermal maturity exposed at the surface across the region.

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

Pm-1 Reactor Core Final Design Report

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

Bagley, R. O.; Cox, F. H.; Carnasale, A.

1962-01-01

The PM-1 water cooled and moderated core contains 741 highly enriched stainless steel cermet tubular fuel elements and 90 lumped B stainless steel burnable poison elements, and it is controlled by 6 Y-shaped europium titanate movable control rods. The core has a lifetime of 1.95 years when operated at its design power level of 9.37 mw of thermal energy. The control of the core is designed so that there is a positive shutdown margin at all times with either one rod stuck completely out or the core or with two rods stuck in the operating condition. The core power ismore » removed by 2125 gpm of pressurized water at an average temperature of 463 deg F and pressure of 1300 psia. In reactors of this type, the core is stable with a negative temperature coefficient of approximately 2.5 x 10/sup -4/ DELTA K/K/ deg F.« less

ARMY GAS-COOLED REACTOR SYSTEMS PROGRAM. Quarterly Progress Report, October 1-December 31, 1963

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

None

1964-02-15

The ML-1 power plant did not operate during the report period; low power reactor physics and shielding experiments were conducted with the ML-1 reactor. Evaluation of moderate corrosion observed on aluminum parts exposed to the ML-1 shield solution indicated no loss of performance capability. Preliminary tests showed that the corrosion probably was caused by heavy metal ions or chlorides in the solution, Massive corrosion observed on the ML-1 fuel element lower spiders was attributed to sub-standard material; failure of some spiders was attributed to a combination of corrosion and sub-standard fabrication. Evaluation indicated that the upper spiders will perform satisfactorilymore » for the design lifetime. Modification, repair, and reassembly of the CSN-1A t-c set was completed. Operation demonstrated bearing stability, but showed that the turbine effective flow area was too large. A bypass flow path in the turbine was being corrected. The TCS-670 t-c set will be stored indefinitely. Since a commercial alternator will be used for the ML-1A, further development of the brushless alternator was postponed indefinitely. Evaluation revealed that the ML-1 improved precooler design was not compatible with ML-1A requirements. Operntion of the IB-17R-2 and -3 test elements in the GETR continued without incident. Preliminary design of the ML-1A power plant was initiated. Design of modifications to the GCRE facility to adapt it to testing the ML-1 reactor skid was initiated. (auth)« less

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

Lifetime Neutron Fluence Analysis of the Ringhals Unit 1 Boiling Water Reactor

NASA Astrophysics Data System (ADS)

Kulesza, Joel A.; Roudén, Jenny; Green, Eva-Lena

2016-02-01

This paper describes a neutron fluence assessment considering the entire commercial operating history (35 cycles or ˜ 25 effective full power years) of the Ringhals Unit 1 reactor pressure vessel beltline region. In this assessment, neutron (E >1.0 MeV) fluence and iron atom displacement distributions were calculated on the moderator tank and reactor pressure vessel structures. To validate those calculations, five in-vessel surveillance chain dosimetry sets were evaluated as well as material samples taken from the upper core grid and wide range neutron monitor tubes to act as a form of retrospective dosimetry. During the analysis, it was recognized that delays in characterizing the retrospective dosimetry samples reduced the amount of reactions available to be counted and complicated the material composition determination. However, the comparisons between the surveillance chain dosimetry measurements (M) and calculated (C) results show similar and consistent results with the linear average M/C ratio of 1.13 which is in good agreement with the resultant least squares best estimate (BE)/C ratios of 1.10 for both neutron (E >1.0 MeV) flux and iron atom displacement rate.

Nuclear Engineering Computer Modules: Reactor Dynamics, RD-1 and RD-2.

ERIC Educational Resources Information Center

Onega, Ronald J.

The objective of the Reactor Dynamics Module, RD-1, is to obtain the kinetics equation without feedback and solve the kinetics equations numerically for one to six delayed neutron groups for time varying reactivity insertions. The computer code FUMOKI (Fundamental Mode Kinetics) will calculate the power as a function of time for either uranium or…

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)

Ethanol Production from Wet-Exploded Wheat Straw Hydrolysate by Thermophilic Anaerobic Bacterium Thermoanaerobacter BG1L1 in a Continuous Immobilized Reactor

NASA Astrophysics Data System (ADS)

Georgieva, Tania I.; Mikkelsen, Marie J.; Ahring, Birgitte K.

Thermophilic ethanol fermentation of wet-exploded wheat straw hydrolysate was investigated in a continuous immobilized reactor system. The experiments were carried out in a lab-scale fluidized bed reactor (FBR) at 70°C. Undetoxified wheat straw hydrolysate was used (3-12% dry matter), corresponding to sugar mixtures of glucose and xylose ranging from 12 to 41 g/1. The organism, thermophilic anaerobic bacterium Thermoanaerobacter BG1L1, exhibited significant resistance to high levels of acetic acid (up to 10 g/1) and other metabolic inhibitors present in the hydrolysate. Although the hydrolysate was not detoxified, ethanol yield in a range of 0.39-0.42 g/g was obtained. Overall, sugar efficiency to ethanol was 68-76%. The reactor was operated continuously for approximately 143 days, and no contamination was seen without the use of any agent for preventing bacterial infections. The tested microorganism has considerable potential to be a novel candidate for lignocellulose bioconversion into ethanol. The work reported here also demonstrates that the use of FBR configuration might be a viable approach for thermophilic anaerobic ethanol fermentation.

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.

IEA-R1 Nuclear Research Reactor: 58 Years of Operating Experience and Utilization for Research, Teaching and Radioisotopes Production

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

Cardenas, Jose Patricio Nahuel; Filho, Tufic Madi; Saxena, Rajendra

IEA-R1 research reactor at the Instituto de Pesquisas Energeticas e Nucleares (Nuclear and Energy Research Institute) IPEN, Sao Paulo, Brazil is the largest power research reactor in Brazil, with a maximum power rating of 5 MWth. It is being used for basic and applied research in the nuclear and neutron related sciences, for the production of radioisotopes for medical and industrial applications, and for providing services of neutron activation analysis, real time neutron radiography, and neutron transmutation doping of silicon. IEA-R1 is a swimming pool reactor, with light water as the coolant and moderator, and graphite and beryllium as reflectors.more » The reactor was commissioned on September 16, 1957 and achieved its first criticality. It is currently operating at 4.5 MWth with a 60-hour cycle per week. In the early sixties, IPEN produced {sup 131}I, {sup 32}P, {sup 198}Au, {sup 24}Na, {sup 35}S, {sup 51}Cr and labeled compounds for medical use. During the past several years, a concerted effort has been made in order to upgrade the reactor power to 5 MWth through refurbishment and modernization programs. One of the reasons for this decision was to produce {sup 99}Mo at IPEN. The reactor cycle will be gradually increased to 120 hours per week continuous operation. It is anticipated that these programs will assure the safe and sustainable operation of the IEA-R1 reactor for several more years, to produce important primary radioisotopes {sup 99}Mo, {sup 125}I, {sup 131}I, {sup 153}Sm and {sup 192}Ir. Currently, all aspects of dealing with fuel element fabrication, fuel transportation, isotope processing, and spent fuel storage are handled by IPEN at the site. The reactor modernization program is slated for completion by 2015. This paper describes 58 years of operating experience and utilization of the IEA-R1 research reactor for research, teaching and radioisotopes production. (authors)« less

Zircon U-Pb age, Lu-Hf isotope, mineral chemistry and geochemistry of Sundamalai peralkaline pluton from the Salem Block, southern India: Implications for Cryogenian adakite-like magmatism in an aborted-rift

NASA Astrophysics Data System (ADS)

Renjith, M. L.; Santosh, M.; Li, Tang; Satyanarayanan, M.; Korakoppa, M. M.; Tsunogae, T.; Subba Rao, D. V.; Kesav Krishna, A.; Nirmal Charan, S.

2016-01-01

The Sundamalai peralkaline pluton is one among the Cryogenian alkaline plutons occurring in the Dharmapuri Rift Zone (DRZ) of the Salem Block in the Southern Granulite Terrane (SGT) of India. Here we present zircon U-Pb age and Lu-Hf isotopic composition, mineral chemistry and geochemistry of the pluton to explore the petrogenesis and geodynamic implications. Systematic modal variation of orthoclase, Na-plagioclase, Ca-amphibole (ferro-edenite and hastingsite) and quartz developed quartz-monzonite and granite litho units in the Sundamalai pluton. Thermometry based on amphibole-plagioclase pair suggests that the pluton was emplaced and solidified at around 4.6 kbar pressure with crystallization of the major phases between 748 and 661 °C. Estimated saturation temperature of zircon (712-698 °C) is also well within this range. However, apatite saturation occurred at higher temperatures between 835 and 870 °C, in contrast with monazite saturation (718-613 °C) that continued up to the late stage of crystallization. Estimated oxygen fugacity values (log fO2: -14 to -17) indicate high oxidation state for the magma that stabilized titanite and magnetite. The magmatic zircons from Sundamalai pluton yielded a weighted mean 206Pb/238U age of 832.6 ± 3.2 Ma. Geochemically, the Sundamalai rocks are high-K to shoshonitic, persodic (Na2O/K2O ratio > 1), silica-saturated (SiO2:65-72 wt.%), and peralkaline in composition (aluminum saturation index, ASI < 1; Alkalinity index, AI < 0). The initial magma was mildly metaluminous which evolved to strongly peralkaline as result of fractional crystallization (plagioclase effect) controlled differentiation between quartz-monzonite and granite. Both rock types have high content of Na2O (5.1-6.3 wt.%), Ba (350-2589 ppm) and Sr (264-1036 ppm); low content of Y (8.7-17 ppm) and Yb (0.96-1.69 ppm); elevated ratios of La/Yb (11-46) and Sr/Y (46-69) and are depleted in Ti, with a positive Sr anomaly suggesting an adakite-like composition and

Solvent refined coal reactor quench system

DOEpatents

Thorogood, R.M.

1983-11-08

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

The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 1: Basic Models

NASA Astrophysics Data System (ADS)

Mosunova, N. A.

2018-05-01

The article describes the basic models included in the EUCLID/V1 integrated code intended for safety analysis of liquid metal (sodium, lead, and lead-bismuth) cooled fast reactors using fuel rods with a gas gap and pellet dioxide, mixed oxide or nitride uranium-plutonium fuel under normal operation, under anticipated operational occurrences and accident conditions by carrying out interconnected thermal-hydraulic, neutronics, and thermal-mechanical calculations. Information about the Russian and foreign analogs of the EUCLID/V1 integrated code is given. Modeled objects, equation systems in differential form solved in each module of the EUCLID/V1 integrated code (the thermal-hydraulic, neutronics, fuel rod analysis module, and the burnup and decay heat calculation modules), the main calculated quantities, and also the limitations on application of the code are presented. The article also gives data on the scope of functions performed by the integrated code's thermal-hydraulic module, using which it is possible to describe both one- and twophase processes occurring in the coolant. It is shown that, owing to the availability of the fuel rod analysis module in the integrated code, it becomes possible to estimate the performance of fuel rods in different regimes of the reactor operation. It is also shown that the models implemented in the code for calculating neutron-physical processes make it possible to take into account the neutron field distribution over the fuel assembly cross section as well as other features important for the safety assessment of fast reactors.

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

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

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

2015-08-20

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

RAZORBACK - A Research Reactor Transient Analysis Code Version 1.0 - Volume 3: Verification and Validation Report.

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

Talley, Darren G.

2017-04-01

This report describes the work and results of the verification and validation (V&V) of the version 1.0 release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This V&V effort was intended to confirm that the code showsmore » good agreement between simulation and actual ACRR operations.« less

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.

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.

Fukushima Daiichi Unit 1 Accident Progression Uncertainty Analysis and Implications for Decommissioning of Fukushima Reactors - Volume I.

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

Gauntt, Randall O.; Mattie, Patrick D.

Sandia National Laboratories (SNL) has conducted an uncertainty analysis (UA) on the Fukushima Daiichi unit (1F1) accident progression with the MELCOR code. The model used was developed for a previous accident reconstruction investigation jointly sponsored by the US Department of Energy (DOE) and Nuclear Regulatory Commission (NRC). That study focused on reconstructing the accident progressions, as postulated by the limited plant data. This work was focused evaluation of uncertainty in core damage progression behavior and its effect on key figures-of-merit (e.g., hydrogen production, reactor damage state, fraction of intact fuel, vessel lower head failure). The primary intent of this studymore » was to characterize the range of predicted damage states in the 1F1 reactor considering state of knowledge uncertainties associated with MELCOR modeling of core damage progression and to generate information that may be useful in informing the decommissioning activities that will be employed to defuel the damaged reactors at the Fukushima Daiichi Nuclear Power Plant. Additionally, core damage progression variability inherent in MELCOR modeling numerics is investigated.« less

Remediation of 1,2,3-trichlorobenzene contaminated soil using a combined thermal desorption-molten salt oxidation reactor system.

PubMed

Li, Jin-hui; Sun, Xiao-fei; Yao, Zhi-tong; Zhao, Xiang-yang

2014-02-01

A combined thermal desorption (TD)-molten salt oxidation (MSO) reactor system was applied to remediate the 1,2,3-trichlorobenzene (1,2,3-TCB) contaminated soil. The TD reactor was used to enrich the contaminant from soil, and its dechlorination of the contaminant was achieved in the MSO reactor. The optimum operating conditions of TD, and the effects of MSO reactor temperatures, additive amounts of the TCB on destruction and removal efficiency (DRE) of TCB and chlorine retention efficiency (CRE) were investigated. The reaction mechanism and pathway were proposed as well. The combined system could remediate the contaminated soil at a large scale of concentration from 5 to 25gkg(-1), and the DRE and CRE reached more than 99% and 95%, respectively, at temperatures above 850°C. The reaction emissions included C6H6, CH4, CO and CO2, and chlorinated species were not detected. It was found that a little increase in the temperature can considerably reduce the emission of C6H6, CH4, and CO, while the CO2 level increased. Copyright © 2014. Published by Elsevier Ltd.

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

Introduction to Reactor Statics Modules, RS-1. Nuclear Engineering Computer Modules.

ERIC Educational Resources Information Center

Edlund, Milton C.

The nine Reactor Statics Modules are designed to introduce students to the use of numerical methods and digital computers for calculation of neutron flux distributions in space and energy which are needed to calculate criticality, power distribution, and fuel burn-up for both slow neutron and fast neutron fission reactors. The diffusion…

A search for neutrino oscillations using the CHOOZ 1 km baseline reactor neutrino experiment

NASA Astrophysics Data System (ADS)

George, Jean

1999-10-01

Neutrino oscillation searches are an active field of research due to the implications their discovery may have for the solar neutrino anomaly as well as for the atmospheric neutrino anomaly. Their discovery may also have broad ramifications for the Standard Model of Particle Physics as a whole. Results from an oscillation search using the CHOOZ long baseline reactor neutrino experiment are presented in this thesis. These results are based on the data taken from June 1997 through April 1998 when the two reactors ran at combined thermal power levels ranging from zero power to their full power level of 8.5 GW. Electron flavored antineutrinos emanating from the reactors were detected through the inverse beta decay channel using a liquid scintillating calorimeter located at a distance of approximately 1 km from the reactor sources. The underground experimental site (300 MWE) provided natural shielding from the background of cosmic ray muons-leading to a background rate more than an order of magnitude lower than the full power signal rate. From the agreement between the detected and expected neutrino event rates no evidence for neutrino oscillations was found (at the 90% C.L.) for the oscillation parameter space governed by Δm 2 > 0.8 × 10-3 eV2 for maximal mixing and by sin2 2Θ > 0.18 for large values of Δm2.

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

9 CFR 78.31 - Brucellosis reactor swine.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 9 Animals and Animal Products 1 2011-01-01 2011-01-01 false Brucellosis reactor swine. 78.31... Restrictions on Interstate Movement of Swine Because of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for immediate slaughter as follows: (1...

9 CFR 78.22 - Brucellosis reactor bison.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 9 Animals and Animal Products 1 2011-01-01 2011-01-01 false Brucellosis reactor bison. 78.22... Restrictions on Interstate Movement of Bison Because of Brucellosis § 78.22 Brucellosis reactor bison. (a) Destination. Brucellosis reactor bison may be moved interstate only for immediate slaughter as follows: (1...

9 CFR 78.31 - Brucellosis reactor swine.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Brucellosis reactor swine. 78.31... Restrictions on Interstate Movement of Swine Because of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for immediate slaughter as follows: (1...

9 CFR 78.7 - Brucellosis reactor cattle.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 9 Animals and Animal Products 1 2011-01-01 2011-01-01 false Brucellosis reactor cattle. 78.7... Restrictions on Interstate Movement of Cattle Because of Brucellosis § 78.7 Brucellosis reactor cattle. (a) Destination. Brucellosis reactor cattle may be moved interstate only for immediate slaughter as follows: (1...

9 CFR 78.7 - Brucellosis reactor cattle.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Brucellosis reactor cattle. 78.7... Restrictions on Interstate Movement of Cattle Because of Brucellosis § 78.7 Brucellosis reactor cattle. (a) Destination. Brucellosis reactor cattle may be moved interstate only for immediate slaughter as follows: (1...

9 CFR 78.22 - Brucellosis reactor bison.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Brucellosis reactor bison. 78.22... Restrictions on Interstate Movement of Bison Because of Brucellosis § 78.22 Brucellosis reactor bison. (a) Destination. Brucellosis reactor bison may be moved interstate only for immediate slaughter as follows: (1...

9 CFR 78.31 - Brucellosis reactor swine.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 9 Animals and Animal Products 1 2012-01-01 2012-01-01 false Brucellosis reactor swine. 78.31... Restrictions on Interstate Movement of Swine Because of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for immediate slaughter as follows: (1...

9 CFR 78.31 - Brucellosis reactor swine.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 9 Animals and Animal Products 1 2013-01-01 2013-01-01 false Brucellosis reactor swine. 78.31... Restrictions on Interstate Movement of Swine Because of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for immediate slaughter as follows: (1...

9 CFR 78.31 - Brucellosis reactor swine.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 9 Animals and Animal Products 1 2014-01-01 2014-01-01 false Brucellosis reactor swine. 78.31... Restrictions on Interstate Movement of Swine Because of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for immediate slaughter as follows: (1...

Impacts of large dams on downstream flow conditions of rivers: Aggradation and reduction of the Medjerda channel capacity downstream of the Sidi Salem dam (Tunisia)

NASA Astrophysics Data System (ADS)

Zahar, Yadh; Ghorbel, Abdelmajid; Albergel, Jean

2008-04-01

SummarySince the opening of the Sidi Salem dam on the watercourse of the Medjerda, in 1981, an alarming narrowing of the riverbed in the lower valley has been observed. This geo-morphological change is attributed to different factors ranking from the reduction in the discharge flows, which used to clean out the riverbed to the periodic releases of turbid water undertaken to remove the silt deposition inside the reservoir, which increased the sediment deposition in the downstream channel. Other smaller hydraulic projects are also held responsible for the loss of the water velocity including a series of concrete sills meant to raise water levels, numerous cross bridges and the management of the downstream Laroussia dam regulating the discharge from the Cap Bon canal. The above anthropogenic factors, in conjunction with natural topographical conditions characterized by a generally shallow slope and a very sinuous watercourse, led to an extremely rapid aggradation of the downstream channel-bed. This paper proposes an analysis of this process and argues that the resulting reduction in channel capacity is one of the major causes of the large floods experienced in the country since 1996.

Reactor operations informal monthly report, May 1, 1995--May 31, 1995

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

Hauptman, H.M.; Petro, J.N.; Jacobi, O.

1995-05-01

This document is an informal progress report for the operational performance of the Brookhaven Medical Research Reactor, and the Brookhaven High Flux Beam Reactor, for the month of May, 1995. Both machines ran well during this period, with no reportable instrumentation problems, all scheduled maintenance performed, and only one reportable occurance, involving a particle on Vest Button, Personnel Radioactive Contamination.

A Neutronic Program for Critical and Nonequilibrium Study of Mobile Fuel Reactors: The Cinsf1D Code

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

Lecarpentier, David; Carpentier, Vincent

2003-01-15

Molten salt reactors (MSRs) have the distinction of having a liquid fuel that is also the coolant. The transport of delayed-neutron precursors by the fuel modifies the precursors' equation. As a consequence, it is necessary to adapt the methods currently used for solid fuel reactors to achieve critical or kinetics calculations for an MSR. A program is presented for which this adaptation has been carried out within the framework of the two-energy-group diffusion theory with one dimension of space. This program has been called Cinsf1D (Cinetique pour reacteur a sels fondus 1D)

Validation of light water reactor calculation methods and JEF-1-based data libraries by TRX and BAPL critical experiments

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

Paratte, J.M.; Pelloni, S.; Grimm, P.

1991-04-01

This paper analyzes the capability of various code systems and JEF-1-based nuclear data libraries to compute light water reactor lattices by comparing calculations with results from thermal reactor benchmark experiments TRX and BAPL and with previously published values. With the JEF-1 evaluation, eigenvalues are generally well predicted within 8 mk (1 mk = 0.001) or less by all code systems, and all methods give reasonable results for the measured reaction rate ratios within, or not too far from, the experimental uncertainty.

NORTHROP REACTOR. REVISION NO. 1 TO THE FINAL SAFEGUARDS REPORT

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

Duncan, J.M.; Shimizu, B.; Romine, R.A.

1962-10-01

Additions and changes related to the original application concerning construction and operation of the Northrop Reactor are given. Answers to 13 questions relative to the Final Safeguards Report are included. Answers are also included concerning 12 questions regarding receipt, possession, and storage of fuel elements. Other information is included concerning personnel changes and changes from Al-clad fuel elements to improved stainless steelclad hydride elements. It is concluded that the improved Northrop Reactor does not present any undue hazard to the health and safety of the operating personnel or the public. (J.R.D.)

Geochemical constraints on crustal anatexis: a case study from the Pan-African Damara granitoids of Namibia

NASA Astrophysics Data System (ADS)

McDermott, F.; Harris, N. B. W.; Hawkesworth, C. J.

1996-05-01

Major and trace element models of recently published vapour-absent mica dehydration melting experiments are used to identify granitoids generated by muscovite and biotite dehydration melting, and to distinguish between plagioclase-limited and biotite-limited, biotite dehydration melting. In the case of granitoids from the Pan-African Damara mobile belt (Namibia), many of the leucogranites and Salem-type granitoids may be modelled by biotite dehydration melting. The low Rb/Sr granitoids (e.g. Donkerhuk Onanis, Salem Onanis, Donkerhuk Nomatsaus, Salem Goas) probably reflect feldspar-limited, biotite dehydration melting (a pelitic source) whereas the high Rb/Sr suites (e.g. Bloedkoppie leucogranite, Stinkbank leucogranite, Salem Swakopmund, Leucocratic Stink bank granite) reflect biotite-limited, biotite dehydration melting (a greywacke source). Alaskites from the Damara belt have major element compositions which are consistent with muscovite dehydration melting, and their positive Eu anomalies are linked to high K2O reflecting K-feldspar entrainment. Combined Zr and LREE (light rare earth element) solubility models indicate that insufficient time (probably less than 104 years) had elapsed between melt generation and melt extraction to ensure that the alaskite melts attained their equilibrium concentrations of Zr and the LREEs. In contrast, the leucogranites and Salem-type granites have attained their equilibrium inventories of these trace elements. Combined Fe2O3 and MgO contents in some samples from two granitoids (the Salem Goas and Donkerhuk Onanis intrusions) are higher than those readily attainable by biotite dehydration melting indicating either: (1) that they contain a contribution from melts generated by incipient garnet breakdown or; (2) that they contain small amounts of an entrained ferromagnesian phase.

Potential use of sugar binding proteins in reactors for regeneration of CO2 fixation acceptor D-Ribulose-1,5-bisphosphate

PubMed Central

Mahato, Sourav; De, Debojyoti; Dutta, Debajyoti; Kundu, Moloy; Bhattacharya, Sumana; Schiavone, Marc T; Bhattacharya, Sanjoy K

2004-01-01

Sugar binding proteins and binders of intermediate sugar metabolites derived from microbes are increasingly being used as reagents in new and expanding areas of biotechnology. The fixation of carbon dioxide at emission source has recently emerged as a technology with potentially significant implications for environmental biotechnology. Carbon dioxide is fixed onto a five carbon sugar D-ribulose-1,5-bisphosphate. We present a review of enzymatic and non-enzymatic binding proteins, for 3-phosphoglycerate (3PGA), 3-phosphoglyceraldehyde (3PGAL), dihydroxyacetone phosphate (DHAP), xylulose-5-phosphate (X5P) and ribulose-1,5-bisphosphate (RuBP) which could be potentially used in reactors regenerating RuBP from 3PGA. A series of reactors combined in a linear fashion has been previously shown to convert 3-PGA, (the product of fixed CO2 on RuBP as starting material) into RuBP (Bhattacharya et al., 2004; Bhattacharya, 2001). This was the basis for designing reactors harboring enzyme complexes/mixtures instead of linear combination of single-enzyme reactors for conversion of 3PGA into RuBP. Specific sugars in such enzyme-complex harboring reactors requires removal at key steps and fed to different reactors necessitating reversible sugar binders. In this review we present an account of existing microbial sugar binding proteins and their potential utility in these operations. PMID:15175111

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

The Egyptian Military Elite: An Operational Code

DTIC Science & Technology

1977-09-01

17 Jan 1955. 6. Salem, Major Salah, "Mehalla’s Great Welcome For Major Salem," Egyptian Gazette, p. 1, 7 Jan 1955. 7. Muhieldin, Lt. Col. Zakaria, "The...Wheel of Reform Will Not Turn Back," Egyptian Gazette, p. 1, 26 Jan 1955. 8. Gohar, Lt. Col. Salah, "Imaginary Incident Explored," Egyptian Gazette...Says Army is Fully Prepared," Al Akhbar , FBIS Vol. 172, p. Al, 4 Sept 1956. 12. Sadat, Anwar al, "Western Powers Determined to Grab Canal," Al

Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

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

NONE

1995-08-01

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized intomore » six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.« less

Nuclear reactor fuel containment safety structure

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

Rosewell, M.P.

A nuclear reactor fuel containment safety structure is disclosed and is shown to include an atomic reactor fuel shield with a fuel containment chamber and exhaust passage means, and a deactivating containment base attached beneath the fuel reactor shield and having exhaust passages, manifold, and fluxing and control material and vessels. 1 claim, 8 figures.

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.

MIDURA (Minefield Detection Using Reconnaissance Assets) 1982-1983 Experimental Test Plan.

DTIC Science & Technology

1982-04-01

3.2.4.2 Subjection Validation at the Salem ONG 27 3.2.4.3 Objective Validity at Fort Huachuca 28 4. TEST FLIGHTS AT ARRAYS IIa, lib, Ilia AND IIIb...subjective validation at the Salem ONG; (3) objective validation at Fort Huachuca. 3.2.4.1 Subjective Image Interpretation at ERIM The initial phase...The ERIM II’s will provide for each image estimate of PD’ Pc and PFA on a 0.00 to 1.00 scale. P is defined as the subjective probability estimate that

International Research Reactor Decommissioning Project

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

Leopando, Leonardo; Warnecke, Ernst

2008-01-15

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

A mini-cavity probe reactor.

NASA Technical Reports Server (NTRS)

Hyland, R. E.

1971-01-01

The mini-cavity reactor is a rocket engine concept which combines the high specific impulse from a central gaseous fueled cavity (0.6 m diam) and NERVA type fuel elements in a driver region that is external to a moderator-reflector zone to produce a compact light weight reactor. The overall dimension including a pressure vessel that is located outside of the spherical reactor is approximately 1.21 m in diameter. Specific impulses up to 2000 sec are obtainable for 220 to 890 N of thrust with pressures less than 1000 atm. Powerplant weights including a radiator for disposing of the power in the driver region are between 4600 and 32,000 kg - less than payloads of the shuttle. This reactor could also be used as a test reactor for gas-core, MHD, breeding and materials research.

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.

A comparison of the technological effectiveness of dairy wastewater treatment in anaerobic UASB reactor and anaerobic reactor with an innovative design.

PubMed

Jedrzejewska-Cicinska, M; Kozak, K; Krzemieniewski, M

2007-10-01

The present research was an investigation of the influence of an innovative design of reactor filled with polyethylene (PE) granulate on model dairy wastewater treatment efficiency under anaerobic conditions compared to that obtained in a typical UASB reactor. The experiment was conducted at laboratory scale. An innovative reactor was designed with the reaction chamber inclined 30 degrees in relation to the ground with upward waste flow and was filled with PE granular material. Raw model dairy wastewater was fed to two anaerobic reactors of different design at the organic loading rate of 4 kg COD m(-3)d(-1). Throughout the experiment, a higher removal efficiency of organic compounds was observed in the reactor with an innovative design and it was higher by 7.1% on average than in the UASB reactor. The total suspended solids was lower in the wastewater treated in the anaerobic reactor with the innovative design. Applying a PE granulated filling in the chamber of the innovative reactor contributed to an even distribution of sludge biomass in the reactor, reducing washout of anaerobic sludge biomass from the reaction chamber and giving a higher organic compounds removal efficiency.

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.

Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP.

PubMed

Abrefah, R G; Sogbadji, R B M; Ampomah-Amoako, E; Birikorang, S A; Odoi, H C; Nyarko, B J B

2011-01-01

The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model, which has a boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245 × 10(-4)s was recorded for the new boron carbide designed model while a value of 1.5571 × 10(-7)s was recorded for the original MCNP design of the GHARR-1. Copyright © 2010 Elsevier Ltd. All rights reserved.

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.

Detecting Dark Photons with Reactor Neutrino Experiments

NASA Astrophysics Data System (ADS)

Park, H. K.

2017-08-01

We propose to search for light U (1 ) dark photons, A', produced via kinetically mixing with ordinary photons via the Compton-like process, γ e-→A'e-, in a nuclear reactor and detected by their interactions with the material in the active volumes of reactor neutrino experiments. We derive 95% confidence-level upper limits on ɛ , the A'-γ mixing parameter, ɛ , for dark-photon masses below 1 MeV of ɛ <1.3 ×10-5 and ɛ <2.1 ×10-5, from NEOS and TEXONO experimental data, respectively. This study demonstrates the applicability of nuclear reactors as potential sources of intense fluxes of low-mass dark photons.

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.

7. Details of northeast elevation, demonstrating the relationship of basement ...

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

7. Details of northeast elevation, demonstrating the relationship of basement windows to column bases. - Roanoke Veterans Administration Hospital, Building No. 6, 1970 Roanoke Boulevard, Salem, Salem, VA

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…

The 235U Prompt Fission Neutron Spectrum in the BR1 Reactor at SCK•CEN

NASA Astrophysics Data System (ADS)

Wagemans, Jan; Malambu, Edouard; Borms, Luc; Fiorito, Luca

2016-02-01

The BR1 research reactor at SCK•CEN has a spherical cavity in the graphite above the reactor core. In this cavity an accurately characterised Maxwellian thermal neutron field is present. Different converters can be loaded in the cavity in order to obtain other types of neutron (and gamma) irradiation fields. Inside the so-called MARK III converter a fast 235U(n,f) prompt fission neutron field can be obtained. With the support of MCNP calculations, irradiations in MARK III can be directly related to the pure 235U(n,f) prompt fission neutron spectrum. For this purpose MARK III spectrum averaged cross sections for the most relevant fluence dosimetry reactions have been determined. A calibration factor for absolute measurements has been determined applying activation dosimetry following ISO/IEC 17025 standards.

``Sleeping reactor`` irradiations: Shutdown reactor determination of short-lived activation products

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

Jerde, E.A.; Glasgow, D.C.

1998-09-01

At the High-Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory, the principal irradiation system has a thermal neutron flux ({phi}) of {approximately} 4 {times} 10{sup 14} n/cm{sup 2} {center_dot} s, permitting the detection of elements via irradiation of 60 s or less. Irradiations of 6 or 7 s are acceptable for detection of elements with half-lives of as little as 30 min. However, important elements such as Al, Mg, Ti, and V have half-lives of only a few minutes. At HFIR, these can be determined with irradiation times of {approximately} 6 s, but the requirement of immediate countingmore » leads to increased exposure to the high activity produced by irradiation in the high flux. In addition, pneumatic system timing uncertainties (about {+-} 0.5 s) make irradiations of < 6 s less reliable. Therefore, the determination of these ultra-short-lived species in mixed matrices has not generally been made at HFIR. The authors have found that very short lived activation products can be produced easily during the period after reactor shutdown (SCRAM), but prior to the removal of spent fuel elements. During this 24- to 36-h period (dubbed the ``sleeping reactor``), neutrons are produced in the beryllium reflector by the reaction {sup 9}Be({gamma},n){sup 8}Be, the gamma rays principally originating in the spent fuel. Upon reactor SCRAM, the flux drops to {approximately} 1 {times} 10{sup 10} n/cm{sup 2} {center_dot} s within 1 h. By the time the fuel elements are removed, the flux has dropped to {approximately} 6 {times} 10{sup 8}. Such fluxes are ideal for the determination of short-lived elements such as Al, Ti, Mg, and V. An important feature of the sleeping reactor is a flux that is not constant.« less

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.

Development of Neutron Imaging System for Neutron Tomography at Thai Research Reactor TRR-1/M1

NASA Astrophysics Data System (ADS)

Wonglee, S.; Khaweerat, S.; Channuie, J.; Picha, R.; Liamsuwan, T.; Ratanatongchai, W.

2017-09-01

The neutron imaging is a powerful non-destructive technique to investigate the internal structure and provides the information which is different from the conventional X-ray/Gamma radiography. By reconstruction of the obtained 2-dimentional (2D) images from the taken different angle around the specimen, the tomographic image can be obtained and it can provide the information in more detail. The neutron imaging system at Thai Research Reactor TRR-1/M1 of Thailand Institute of Nuclear Technology (Public Organization) has been developed to conduct the neutron tomography since 2014. The primary goal of this work is to serve the investigation of archeological samples, however, this technique can also be applied to various fields, such as investigation of industrial specimen and others. This research paper presents the performance study of a compact neutron camera manufactured by Neutron Optics such as speed and sensitivity. Furthermore, the 3-dimentional (3D) neutron image was successfully reconstructed at the developed neutron imaging system of TRR-1/M1.

Detecting Dark Photons with Reactor Neutrino Experiments.

PubMed

Park, H K

2017-08-25

We propose to search for light U(1) dark photons, A^{'}, produced via kinetically mixing with ordinary photons via the Compton-like process, γe^{-}→A^{'}e^{-}, in a nuclear reactor and detected by their interactions with the material in the active volumes of reactor neutrino experiments. We derive 95% confidence-level upper limits on ε, the A^{'}-γ mixing parameter, ε, for dark-photon masses below 1 MeV of ε<1.3×10^{-5} and ε<2.1×10^{-5}, from NEOS and TEXONO experimental data, respectively. This study demonstrates the applicability of nuclear reactors as potential sources of intense fluxes of low-mass dark photons.

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.

Generic Stellarator-like Magnetic Fusion Reactor

NASA Astrophysics Data System (ADS)

Sheffield, John; Spong, Donald

2015-11-01

The Generic Magnetic Fusion Reactor paper, published in 1985, has been updated, reflecting the improved science and technology base in the magnetic fusion program. Key changes beyond inflation are driven by important benchmark numbers for technologies and costs from ITER construction, and the use of a more conservative neutron wall flux and fluence in modern fusion reactor designs. In this paper the generic approach is applied to a catalyzed D-D stellarator-like reactor. It is shown that an interesting power plant might be possible if the following parameters could be achieved for a reference reactor: R/ < a > ~ 4 , confinement factor, fren = 0.9-1.15, < β > ~ 8 . 0 -11.5 %, Zeff ~ 1.45 plus a relativistic temperature correction, fraction of fast ions lost ~ 0.07, Bm ~ 14-16 T, and R ~ 18-24 m. J. Sheffield was supported under ORNL subcontract 4000088999 with the University of Tennessee.

Reactor Design and Decommissioning - An Overview of International Activities in Post Fukushima Era1 - 12396

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

Devgun, Jas S.; Laraia, Michele; Pescatore, Claudio

Accidents at the Fukushima Dai-ichi reactors as a result of the devastating earthquake and tsunami of March 11, 2011 have not only dampened the nuclear renaissance but have also initiated a re-examination of the design and safety features for the existing and planned nuclear reactors. Even though failures of some of the key site features at Fukushima can be attributed to events that in the past would have been considered as beyond the design basis, the industry as well as the regulatory authorities are analyzing what features, especially passive features, should be designed into the new reactor designs to minimizemore » the potential for catastrophic failures. It is also recognized that since the design of the Fukushima BWR reactors which were commissioned in 1971, many advanced safety features are now a part of the newer reactor designs. As the recovery efforts at the Fukushima site are still underway, decisions with respect to the dismantlement and decommissioning of the damaged reactors and structures have not yet been finalized. As it was with Three Mile Island, it could take several decades for dismantlement, decommissioning and clean up, and the project poses especially tough challenges. Near-term assessments have been issued by several organizations, including the IAEA, the USNRC and others. Results of such investigations will lead to additional improvements in system and site design measures including strengthening of the anti-tsunami defenses, more defense-in-depth features in reactor design, and better response planning and preparation involving reactor sites. The question also arises what would the effect be on the decommissioning scene worldwide, and what would the effect be on the new reactors when they are eventually retired and dismantled. This paper provides an overview of the US and international activities related to recovery and decommissioning including the decommissioning features in the reactor design process and examines these from a new

Reactor core isolation cooling system

DOEpatents

Cooke, F.E.

1992-12-08

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

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

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.

1977-01-01

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

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)

Nuclear Energy Enabling Technologies (NEET) Reactor Materials: News for the Reactor Materials Crosscut, May 2016

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

Maloy, Stuart Andrew

In this newsletter for Nuclear Energy Enabling Technologies (NEET) Reactor Materials, pages 1-3 cover highlights from the DOE-NE (Nuclear Energy) programs, pages 4-6 cover determining the stress-strain response of ion-irradiated metallic materials via spherical nanoindentation, and pages 7-8 cover theoretical approaches to understanding long-term materials behavior in light water reactors.

Nuclear propulsion apparatus with alternate reactor segments

DOEpatents

Szekely, Thomas

1979-04-03

1. Nuclear propulsion apparatus comprising: A. means for compressing incoming air; B. nuclear fission reactor means for heating said air; C. means for expanding a portion of the heated air to drive said compressing means; D. said nuclear fission reactor means being divided into a plurality of radially extending segments; E. means for directing a portion of the compressed air for heating through alternate segments of said reactor means and another portion of the compressed air for heating through the remaining segments of said reactor means; and F. means for further expanding the heated air from said drive means and the remaining heated air from said reactor means through nozzle means to effect reactive thrust on said apparatus.

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.

Feasibility study of a magnetic fusion production reactor

NASA Astrophysics Data System (ADS)

Moir, R. W.

1986-12-01

A magnetic fusion reactor can produce 10.8 kg of tritium at a fusion power of only 400 MW —an order of magnitude lower power than that of a fission production reactor. Alternatively, the same fusion reactor can produce 995 kg of plutonium. Either a tokamak or a tandem mirror production plant can be used for this purpose; the cost is estimated at about 1.4 billion (1982 dollars) in either case. (The direct costs are estimated at 1.1 billion.) The production cost is calculated to be 22,000/g for tritium and 260/g for plutonium of quite high purity (1%240Pu). Because of the lack of demonstrated technology, such a plant could not be constructed today without significant risk. However, good progress is being made in fusion technology and, although success in magnetic fusion science and engineering is hard to predict with assurance, it seems possible that the physics basis and much of the needed technology could be demonstrated in facilities now under construction. Most of the remaining technology could be demonstrated in the early 1990s in a fusion test reactor of a few tens of megawatts. If the Magnetic Fusion Energy Program constructs a fusion test reactor of approximately 400 MW of fusion power as a next step in fusion power development, such a facility could be used later as a production reactor in a spinoff application. A construction decision in the late 1980s could result in an operating production reactor in the late 1990s. A magnetic fusion production reactor (MFPR) has four potential advantages over a fission production reactor: (1) no fissile material input is needed; (2) no fissioning exists in the tritium mode and very low fissioning exists in the plutonium mode thus avoiding the meltdown hazard; (3) the cost will probably be lower because of the smaller thermal power required; (4) and no reprocessing plant is needed in the tritium mode. The MFPR also has two disadvantages: (1) it will be more costly to operate because it consumes rather than sells

Moon base reactor system

NASA Technical Reports Server (NTRS)

Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

1989-01-01

The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

Treatment of fruit-juice industry wastewater in a two-stage anaerobic hybrid (AH) reactor system followed by a sequencing batch reactor (SBR).

PubMed

Tawfik, A; El-Kamah, H

2012-01-01

This study has been carried out to assess the performance of a combined system consisting of an anaerobic hybrid (AH) reactor followed by a sequencing batch reactor (SBR) for treatment of fruit-juice industry wastewater at a temperature of 26 degrees C. Three experimental runs were conducted in this investigation. In the first experiment, a single-stage AH reactor was operated at a hydraulic retention time (HRT) of 10.2 h and organic loading rate (OLR) of 11.8 kg COD m(-3) d(-1). The reactor achieved a removal efficiency of 42% for chemical oxygen demand (COD), 50.8% for biochemical oxygen demand (BOD5), 50.3% for volatile fatty acids (VFA) and 56.4% for total suspended solids (TSS). In the second experiment, two AH reactors connected in series achieved a higher removal efficiency for COD (67.4%), BOD5 (77%), and TSS (71.5%) at a total HRT of 20 h and an OLR of 5.9 kg COD m(-3) d(-1). For removal of the remaining portions of COD, BOD5 and TSS from the effluent of the two-stage AH system, a sequencing batch reactor (SBR) was investigated as a post-treatment unit. The reactor achieved a substantial reduction in total COD, resulting in an average effluent concentration of 50 mg L(-1) at an HRT of 11 h and OLR of 5.3 kg COD m(-3) d(-1). Almost complete removal of total BOD5 and oil and grease was achieved, i.e. 10 mg L(-1) and 1.2 mg L(-1), respectively, remained in the final effluent of the SBR.

Evaluation of power density on the bioethanol production using mesoscale oscillatory baffled reactor and stirred tank reactor

NASA Astrophysics Data System (ADS)

Yussof, H. W.; Bahri, S. S.; Mazlan, N. A.

2018-03-01

A recent development in oscillatory baffled reactor technology is down-scaling the reactor, so that it can be used for production of small-scale bioproduct. In the present study, a mesoscale oscillatory baffled reactor (MOBR) with central baffle system was developed. The reactor performance of the MOBR was compared with conventional stirred tank reactor (STR) to evaluate the performance of bioethanol fermentation using Saccharomyces cerevisiae. Evaluation was made at similar power density of 24.21, 57.38, 112.35 and 193.67 Wm-3 by varying frequency (f), amplitude (xo) and agitation speed (rpm). It was found that the MOBR improved the mixing intensity resulted in lower glucose concentration (0.988 gL-1) and higher bioethanol concentration (38.98 gL-1) after 12 hours fermentation at power density of 193.67 Wm-3. Based on the results, the bioethanol yield obtained using MOBR was 39% higher than the maximum achieved in STR. Bioethanol production using MOBR proved to be feasible as it is not only able to compete with conventional STR but also offers advantages of straight-forward scale-up, whereas it is complicated and difficult in STR. Overall, MOBR offers great prospective over the conventional STR.

Inactive enzymatic mutant proteins (phosphoglycerate mutase and enolase) as sugar binders for ribulose-1,5-bisphosphate regeneration reactors

PubMed Central

De, Debojyoti; Dutta, Debajyoti; Kundu, Moloy; Mahato, Sourav; Schiavone, Marc T; Chaudhuri, Surabhi; Giri, Ashok; Gupta, Vidya; Bhattacharya, Sanjoy K

2005-01-01

Background Carbon dioxide fixation bioprocess in reactors necessitates recycling of D-ribulose1,5-bisphosphate (RuBP) for continuous operation. A radically new close loop of RuBP regenerating reactor design has been proposed that will harbor enzyme-complexes instead of purified enzymes. These reactors will need binders enabling selective capture and release of sugar and intermediate metabolites enabling specific conversions during regeneration. In the current manuscript we describe properties of proteins that will act as potential binders in RuBP regeneration reactors. Results We demonstrate specific binding of 3-phosphoglycerate (3PGA) and 3-phosphoglyceraldehyde (3PGAL) from sugar mixtures by inactive mutant of yeast enzymes phosphoglycerate mutase and enolase. The reversibility in binding with respect to pH and EDTA has also been shown. No chemical conversion of incubated sugars or sugar intermediate metabolites were found by the inactive enzymatic proteins. The dissociation constants for sugar metabolites are in the micromolar range, both proteins showed lower dissociation constant (Kd) for 3-phosphoglycerate (655–796 μM) compared to 3-phosphoglyceraldehyde (822–966 μM) indicating higher affinity for 3PGA. The proteins did not show binding to glucose, sucrose or fructose within the sensitivity limits of detection. Phosphoglycerate mutase showed slightly lower stability on repeated use than enolase mutants. Conclusions The sugar and their intermediate metabolite binders may have a useful role in RuBP regeneration reactors. The reversibility of binding with respect to changes in physicochemical factors and stability when subjected to repeated changes in these conditions are expected to make the mutant proteins candidates for in-situ removal of sugar intermediate metabolites for forward driving of specific reactions in enzyme-complex reactors. PMID:15689239

78 FR 64028 - Decommissioning of Nuclear Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0035] Decommissioning of Nuclear Power Reactors AGENCY... Commission (NRC) is issuing Revision 1 of regulatory guide (RG) 1.184 ``Decommissioning of Nuclear Power... the NRC's regulations relating to the decommissioning process for nuclear power reactors. The revision...

EBT reactor systems analysis and cost code: description and users guide (Version 1)

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

Santoro, R.T.; Uckan, N.A.; Barnes, J.M.

1984-06-01

An ELMO Bumpy Torus (EBT) reactor systems analysis and cost code that incorporates the most recent advances in EBT physics has been written. The code determines a set of reactors that fall within an allowed operating window determined from the coupling of ring and core plasma properties and the self-consistent treatment of the coupled ring-core stability and power balance requirements. The essential elements of the systems analysis and cost code are described, along with the calculational sequences leading to the specification of the reactor options and their associated costs. The input parameters, the constraints imposed upon them, and the operatingmore » range over which the code provides valid results are discussed. A sample problem and the interpretation of the results are also presented.« less

Strengthening IAEA Safeguards for Research Reactors

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

Reid, Bruce D.; Anzelon, George A.; Budlong-Sylvester, Kory

During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half amore » dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors could pose a safeguards challenge. To strengthen the effectiveness of safeguards at the State level, this paper advocates that the IAEA consider ways to focus additional attention and broaden its safeguards toolbox for research reactors. This increase in focus on the research reactors could begin with the recognition that the research reactor (of any size) could be a common path element on a large number of technically plausible pathways that must be considered when performing acquisition pathway analysis (APA) for developing a State Level Approach (SLA) and Annual Implementation Plan

Health physics aspects of advanced reactor licensing reviews

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

Hinson, C.S.

1995-03-01

The last Construction Permit to be issued by the U.S. Nuclear Regulatory Commission (NRC) for a U.S. light water reactor (LWR) was granted in the late 1970s. In 1989 the NRC issued 10 CFR Part 52 which is intended to serve as a framework for the licensing of future reactor designs. The NRC is currently reviewing four different future on {open_quotes}next-generation{close_quotes} reactor designs. Two of these designs are classified as evolutionary designs (modified versions of current generation LWRs) and two are advanced designs (reactors incorporating simplified designs and passive means for accident mitigation). These {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovativemore » design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs currently being reviewed by the NRC.« less

Design options for a bunsen reactor.

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

Moore, Robert Charles

2013-10-01

This work is being performed for Matt Channon Consulting as part of the Sandia National Laboratories New Mexico Small Business Assistance Program (NMSBA). Matt Channon Consulting has requested Sandia's assistance in the design of a chemical Bunsen reactor for the reaction of SO2, I2 and H2O to produce H2SO4 and HI with a SO2 feed rate to the reactor of 50 kg/hour. Based on this value, an assumed reactor efficiency of 33%, and kinetic data from the literature, a plug flow reactor approximately 1%E2%80%9D diameter and and 12 inches long would be needed to meet the specification of the project.more » Because the Bunsen reaction is exothermic, heat in the amount of approximately 128,000 kJ/hr would need to be removed using a cooling jacket placed around the tubular reactor. The available literature information on Bunsen reactor design and operation, certain support equipment needed for process operation and a design that meet the specification of Matt Channon Consulting are presented.« 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

75 FR 28590 - Procurement List Additions

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-21

... Container. NPA: Winston-Salem Industries for the Blind, Winston-Salem, NC. Contracting Activity: DEPT OF THE... DEPT OF THE ARMY, ACA NATICK, Natick, MA. Tape, Insulation, Electrical NSN: 5970-00-240-0617. NSN: 5970...

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.

Research Program of a Super Fast Reactor

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

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie

2006-07-01

Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is notmore » breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)« less

Efficient Preparation of Enantiopure D-Phenylalanine through Asymmetric Resolution Using Immobilized Phenylalanine Ammonia-Lyase from Rhodotorula glutinis JN-1 in a Recirculating Packed-Bed Reactor

PubMed Central

Huang, Nan; Cui, Wenjing; Liu, Zhongmei; Xiao, Ke; Zhou, Zhemin

2014-01-01

An efficient enzymatic process was developed to produce optically pure D-phenylalanine through asymmetric resolution of the racemic DL-phenylalanine using immobilized phenylalanine ammonia-lyase (RgPAL) from Rhodotorula glutinis JN-1. RgPAL was immobilized on a modified mesoporous silica support (MCM-41-NH-GA). The resulting MCM-41-NH-GA-RgPAL showed high activity and stability. The resolution efficiency using MCM-41-NH-GA-RgPAL in a recirculating packed-bed reactor (RPBR) was higher than that in a stirred-tank reactor. Under optimal operational conditions, the volumetric conversion rate of L-phenylalanine and the productivity of D-phenylalanine reached 96.7 mM h−1 and 0.32 g L−1 h−1, respectively. The optical purity (ee D) of D-phenylalanine exceeded 99%. The RPBR ran continuously for 16 batches, the conversion ratio did not decrease. The reactor was scaled up 25-fold, and the productivity of D-phenylalanine (ee D>99%) in the scaled-up reactor reached 7.2 g L−1 h−1. These results suggest that the resolution process is an alternative method to produce highly pure D-phenylalanine. PMID:25268937

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

Control console replacement at the WPI Reactor

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

Not Available

1992-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

An investigation of reactivity effect due to inadvertent filling of the irradiation channels with water in NIRR-1 Nigeria Research Reactor-1.

PubMed

Iliyasu, U; Ibrahim, Y V; Umar, Sadiq; Agbo, S A; Jibrin, Y

2017-05-01

Investigation of reactivity variation due to flooding of the irradiation channels of Nigeria Research Reactor (NIRR-1) a low power miniature neutron source reactor (MNSR) located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria Nigeria using the MCNP code for High Enrich Uranium (HEU) and Low Enrich Uranium (LEU) core has been simulated in this present study. In this work, the excess reactivity worth of flooding HEU core for 1 inner, 2 inner, 3 inner, 4 inner and all inner are 0.318mk, 0.577mk, 0.318mk, 1.204mk and 1.503mk respectively, and outer irradiation channels are 0.119mk, 0.169mk, 0.348mk, 0.438mk and 0.418mk respectively, the highest excess reactivity result from flooding both inner and outer irradiation channels is 2.04mk (±1.72×10 -7 ), the excess reactivity for LEU core was 0.299mk, 0.568mk, 0.896mk, 1.195mk and 1.524mk in the inner irradiation channels, and the outer irradiation channels are 0.129mk, 0.189mk, 0.219mk, 0.269mk and 0.548mk where the highest excess reactivity was 1.942mk (±1.64×10 -7 ) resulting from flooding inner and outer irradiation channels. The reactivity induced by flooding of the irradiation channels of NIRR-1 with water is within design safety limit enshrined in Safety Analysis Report of NIRR-1. The results also compare well with literature. Copyright © 2017 Elsevier Ltd. All rights reserved.

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. .

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.

Pressurized-water reactor internals aging degradation study. Phase 1

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

Luk, K.H.

1993-09-01

This report documents the results of a Phase I study on the effects of aging degradations on pr internals. Primary stressers for internals an generated by the primary coolant flow in the they include unsteady hydrodynamic forces and pump-generated pressure pulsations. Other stressors are applied loads, manufacturing processes, impurities in the coolant and exposures to fast neutron fluxes. A survey of reported aging-related failure information indicates that fatigue, stress corrosion cracking (SCC) and mechanical wear are the three major aging-related degradation mechanisms for PWR internals. Significant reported failures include thermal shield flow-induced vibration problems, SCC in guide tube support pinsmore » and core support structure bolts, fatigue-induced core baffle water-jet impingement problems and excess wear in flux thimbles. Many of the reported problems have been resolved by accepted engineering practices. Uncertainties remain in the assessment of long-term neutron irradiation effects and environmental factors in high-cycle fatigue failures. Reactor internals are examined by visual inspections and the technique is access limited. Improved inspection methods, especially one with an early failure detection capability, can enhance the safety and efficiency of reactor operations.« 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.

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

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.

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.

The Experimental Breeder Reactor II seismic probabilistic risk assessment

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

Roglans, J; Hill, D J

1994-02-01

The Experimental Breeder Reactor II (EBR-II) is a US Department of Energy (DOE) Category A research reactor located at Argonne National Laboratory (ANL)-West in Idaho. EBR-II is a 62.5 MW-thermal Liquid Metal Reactor (LMR) that started operation in 1964 and it is currently being used as a testbed in the Integral Fast Reactor (IFR) Program. ANL has completed a Level 1 Probabilistic Risk Assessment (PRA) for EBR-II. The Level 1 PRA for internal events and most external events was completed in June 1991. The seismic PRA for EBR-H has recently been completed. The EBR-II reactor building contains the reactor, themore » primary system, and the decay heat removal systems. The reactor vessel, which contains the core, and the primary system, consisting of two primary pumps and an intermediate heat exchanger, are immersed in the sodium-filled primary tank, which is suspended by six hangers from a beam support structure. Three systems or functions in EBR-II were identified as the most significant from the standpoint of risk of seismic-induced fuel damage: (1) the reactor shutdown system, (2) the structural integrity of the passive decay heat removal systems, and (3) the integrity of major structures, like the primary tank containing the reactor that could threaten both the reactivity control and decay heat removal functions. As part of the seismic PRA, efforts were concentrated in studying these three functions or systems. The passive safety response of EBR-II reactor -- both passive reactivity shutdown and passive decay heat removal, demonstrated in a series of tests in 1986 -- was explicitly accounted for in the seismic PRA as it had been included in the internal events assessment.« less

Assessment of Sensor Technologies for Advanced Reactors

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

Korsah, Kofi; Kisner, R. A.; Britton Jr., C. L.

This paper provides an assessment of sensor technologies and a determination of measurement needs for advanced reactors (AdvRx). It is a summary of a study performed to provide the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program. The study covered two broad reactor technology categories: High Temperature Reactors and Fast Reactors. The scope of “High temperature reactors” included Gen IV reactors whose coolant exit temperatures exceed ≈650 °C and are moderated (as opposed to fast reactors). To bound the scope formore » fast reactors, this report reviewed relevant operating experience from US-operated Sodium Fast Reactor (SFR) and relevant test experience from the Fast Flux Test Facility (FFTF). For high temperature reactors the study showed that in many cases instrumentation have performed reasonably well in research and demonstration reactors. However, even in cases where the technology is “mature” (such as thermocouples), HTGRs can benefit from improved technologies. Current HTGR instrumentation is generally based on decades-old technology and adapting newer technologies could provide significant advantages. For sodium fast reactors, the study found that several key research needs arise around (1) radiation-tolerant sensor design for in-vessel or in-core applications, where possible non-invasive sensing approaches for key parameters that minimize the need to deploy sensors in-vessel, (2) approaches to exfiltrating data from in-vessel sensors while minimizing penetrations, (3) calibration of sensors in-situ, and (4) optimizing sensor placements to maximize the information content while minimizing the number of sensors needed.« less

Evaluation of a Method for Remote Detection of Fuel Relocation Outside the Original Core Volumes of Fukushima Reactor Units 1-3

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

Douglas W. Akers; Edwin A. Harvego

2012-08-01

This paper presents the results of a study to evaluate the feasibility of remotely detecting and quantifying fuel relocation from the core to the lower head, and to regions outside the reactor vessel primary containment of the Fukushima 1-3 reactors. The goals of this study were to determine measurement conditions and requirements, and to perform initial radiation transport sensitivity analyses for several potential measurement locations inside the reactor building. The radiation transport sensitivity analyses were performed based on reactor design information for boiling water reactors (BWRs) similar to the Fukushima reactors, ORIGEN2 analyses of 3-cycle BWR fuel inventories, and datamore » on previously molten fuel characteristics from TMI- 2. A 100 kg mass of previously molten fuel material located on the lower head of the reactor vessel was chosen as a fuel interrogation sensitivity target. Two measurement locations were chosen for the transport analyses, one inside the drywell and one outside the concrete biological shield surrounding the drywell. Results of these initial radiation transport analyses indicate that the 100 kg of previously molten fuel material may be detectable at the measurement location inside the drywell, but that it is highly unlikely that any amount of fuel material inside the RPV will be detectable from a location outside the concrete biological shield surrounding the drywell. Three additional fuel relocation scenarios were also analyzed to assess detection sensitivity for varying amount of relocated material in the lower head of the reactor vessel, in the control rods perpendicular to the detector system, and on the lower head of the drywell. Results of these analyses along with an assessment of background radiation effects and a discussion of measurement issues, such as the detector/collimator design, are included in the paper.« less

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

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...

Reactor for exothermic reactions

DOEpatents

Smith, Jr., Lawrence A.; Hearn, Dennis; Jones, Jr., Edward M.

1993-01-01

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

Reactor for exothermic reactions

DOEpatents

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

1993-03-02

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

Biological hydrogen production by Clostridium acetobutylicum in an unsaturated flow reactor.

PubMed

Zhang, Husen; Bruns, Mary Ann; Logan, Bruce E

2006-02-01

A mesophilic unsaturated flow (trickle bed) reactor was designed and tested for H2 production via fermentation of glucose. The reactor consisted of a column packed with glass beads and inoculated with a pure culture (Clostridium acetobutylicum ATCC 824). A defined medium containing glucose was fed at a flow rate of 1.6 mL/min (0.096 L/h) into the capped reactor, producing a hydraulic retention time of 2.1 min. Gas-phase H2 concentrations were constant, averaging 74 +/- 3% for all conditions tested. H2 production rates increased from 89 to 220 mL/hL of reactor when influent glucose concentrations were varied from 1.0 to 10.5 g/L. Specific H2 production rate ranged from 680 to 1270 mL/g glucose per liter of reactor (total volume). The H2 yield was 15-27%, based on a theoretical limit by fermentation of 4 moles of H2 from 1 mole of glucose. The major fermentation by-products in the liquid effluent were acetate and butyrate. The reactor rapidly (within 60-72 h) became clogged with biomass, requiring manual cleaning of the system. In order to make long-term operation of the reactor feasible, biofilm accumulation in the reactor will need to be controlled through some process such as backwashing. These tests using an unsaturated flow reactor demonstrate the feasibility of the process to produce high H2 gas concentrations in a trickle-bed type of reactor. A likely application of this reactor technology could be H2 gas recovery from pre-treatment of high carbohydrate-containing wastewaters.

Nitrate removal with lateral flow sulphur autotrophic denitrification reactor.

PubMed

Lv, Xiaomei; Shao, Mingfei; Li, Ji; Xie, Chuanbo

2014-01-01

An innovative lateral flow sulphur autotrophic denitrification (LFSAD) reactor was developed in this study; the treatment performance was evaluated and compared with traditional sulphur/limestone autotrophic denitrification (SLAD) reactor. Results showed that nitrite accumulation in the LFSAD reactor was less than 1.0 mg/L during the whole operation. Denitrification rate increased with the increased initial alkalinity and was approaching saturation when initial alkalinity exceeded 2.5 times the theoretical value. Higher influent nitrate concentration could facilitate nitrate removal capacity. In addition, denitrification efficiency could be promoted under an appropriate reflux ratio, and the highest nitrate removal percentage was achieved under reflux ratio of 200%, increased by 23.8% than that without reflux. Running resistance was only about 1/9 of that in SLAD reactor with equal amount of nitrate removed, which was the prominent excellence of the new reactor. In short, this study indicated that the developed reactor was feasible for nitrate removal from waters with lower concentrations, including contaminated surface water, groundwater or secondary effluent of municipal wastewater treatment with fairly low running resistance. The innovation in reactor design in this study may bring forth new ideas of reactor development of sulphur autotrophic denitrification for nitrate-contaminated water treatment.

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.

ETRCF, TRA654, INTERIOR. REACTOR OPERATED IN WATERFILLED TANK. CAMERA LOOKS ...

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

ETR-CF, TRA-654, INTERIOR. REACTOR OPERATED IN WATER-FILLED TANK. CAMERA LOOKS DOWN FROM ABOVE UPON LATER (NON-NUCLEAR) EXPERIMENTAL GEAR. INL NEGATIVE NO. HD24-1-1. Mike Crane, Photographer, ca. 2003 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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.

9 CFR 78.22 - Brucellosis reactor bison.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 9 Animals and Animal Products 1 2014-01-01 2014-01-01 false Brucellosis reactor bison. 78.22... AGRICULTURE INTERSTATE TRANSPORTATION OF ANIMALS (INCLUDING POULTRY) AND ANIMAL PRODUCTS BRUCELLOSIS Restrictions on Interstate Movement of Bison Because of Brucellosis § 78.22 Brucellosis reactor bison. (a...

9 CFR 78.22 - Brucellosis reactor bison.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 9 Animals and Animal Products 1 2013-01-01 2013-01-01 false Brucellosis reactor bison. 78.22... AGRICULTURE INTERSTATE TRANSPORTATION OF ANIMALS (INCLUDING POULTRY) AND ANIMAL PRODUCTS BRUCELLOSIS Restrictions on Interstate Movement of Bison Because of Brucellosis § 78.22 Brucellosis reactor bison. (a...

9 CFR 78.7 - Brucellosis reactor cattle.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 9 Animals and Animal Products 1 2014-01-01 2014-01-01 false Brucellosis reactor cattle. 78.7... AGRICULTURE INTERSTATE TRANSPORTATION OF ANIMALS (INCLUDING POULTRY) AND ANIMAL PRODUCTS BRUCELLOSIS Restrictions on Interstate Movement of Cattle Because of Brucellosis § 78.7 Brucellosis reactor cattle. (a...

9 CFR 78.7 - Brucellosis reactor cattle.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 9 Animals and Animal Products 1 2013-01-01 2013-01-01 false Brucellosis reactor cattle. 78.7... AGRICULTURE INTERSTATE TRANSPORTATION OF ANIMALS (INCLUDING POULTRY) AND ANIMAL PRODUCTS BRUCELLOSIS Restrictions on Interstate Movement of Cattle Because of Brucellosis § 78.7 Brucellosis reactor cattle. (a...

9 CFR 78.7 - Brucellosis reactor cattle.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 9 Animals and Animal Products 1 2012-01-01 2012-01-01 false Brucellosis reactor cattle. 78.7... AGRICULTURE INTERSTATE TRANSPORTATION OF ANIMALS (INCLUDING POULTRY) AND ANIMAL PRODUCTS BRUCELLOSIS Restrictions on Interstate Movement of Cattle Because of Brucellosis § 78.7 Brucellosis reactor cattle. (a...

9 CFR 78.22 - Brucellosis reactor bison.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 9 Animals and Animal Products 1 2012-01-01 2012-01-01 false Brucellosis reactor bison. 78.22... AGRICULTURE INTERSTATE TRANSPORTATION OF ANIMALS (INCLUDING POULTRY) AND ANIMAL PRODUCTS BRUCELLOSIS Restrictions on Interstate Movement of Bison Because of Brucellosis § 78.22 Brucellosis reactor bison. (a...

Target-fueled nuclear reactor for medical isotope production

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

Coats, Richard L.; Parma, Edward J.

A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7more » to 21 days.« less

Pentachlorophenol (PCP) dechlorination in horizontal-flow anaerobic immobilized biomass (HAIB) reactors.

PubMed

Damianovic, M H R Z; Moraes, E M; Zaiat, M; Foresti, E

2009-10-01

This study verifies the potential applicability of horizontal-flow anaerobic immobilized biomass (HAIB) reactors to pentachlorophenol (PCP) dechlorination. Two bench-scale HAIB reactors (R1 and R2) were filled with cubic polyurethane foam matrices containing immobilized anaerobic sludge. The reactors were then continuously fed with synthetic wastewater consisting of PCP, glucose, acetic acid, and formic acid as co-substrates for PCP anaerobic degradation. Before being immobilized in polyurethane foam matrices, the biomass was exposed to wastewater containing PCP in reactors fed at a semi-continuous rate of 2.0 microg PCP g(-1) VS. The applied PCP loading rate was increased from 0.05 to 2.59 mg PCP l(-1)day(-1) for R1, and from 0.06 to 4.15 mg PCP l(-1)day(-1) for R2. The organic loading rates (OLR) were 1.1 and 1.7 kg COD m(-3)day(-1) at hydraulic retention times (HRT) of 24h for R1 and 18 h for R2. Under such conditions, chemical oxygen demand (COD) removal efficiencies of up to 98% were achieved in the HAIB reactors. Both reactors exhibited the ability to remove 97% of the loaded PCP. Dichlorophenol (DCP) was the primary chlorophenol detected in the effluent. The adsorption of PCP and metabolites formed during PCP degradation in the packed bed was negligible for PCP removal efficiency.

Nuclear reactors built, being built, or planned, 1991

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

Simpson, B.

1992-07-01

This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor ismore » an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).« less

SPERT1. Contextual aerial view of SPERTI Reactor Pit Building (PER605) ...

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

SPERT-1. Contextual aerial view of SPERT-I Reactor Pit Building (PER-605) at top of view, and its accessories: the earth-shielded instrument cell (PER-606) immediately adjacent to it; the Guard House (PER-607) to its right; and the Terminal Building in lower center of view (PER-604). Camera faces west. Road and buried line leaving view at right lead to Control Building (PER-601) out of view. Sagebrush vegetation has been scraped from around buildings. Photographer: R.G. Larsen. Date: June 6, 1955. INEEL negative no. 55-1477. - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Risk Management for Sodium Fast Reactors.

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

Denman, Matthew R.; Groth, Katrina; Cardoni, Jeffrey N.

2015-01-01

Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event withmore » the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.« less

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.

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

In-reactor performance of LWR-type tritium target rods

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

Lanning, D.D.; Paxton, M.M.; Crumbaugh, L.

Pacific Northwest Laboratory has conducted several 1-yr irradiation tests of light water reactor-type tritium target rods. These tests have been sponsored by the U.S. Department of Energy's Office of New Production Reactors. The first test, designated water capsule-1 (WC-1), was conducted in the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory from November 1989 to December 1990. The test vehicle contained a single 4-ft target rod within a pressurized water capsule. The capsule maintained the rod at pressurized water reactor (PWR)-type water temperature and pressure conditions. Posttest nondestructive examinations of the WC-1 rod involved visual examinations, dimensional checks,more » gamma scanning, and neutron radiography. The results indicate that the rod maintained the integrity of its pressure seal and was otherwise unaltered both mechanically and dimensionally by its irradiation and posttest handling.« less

Vulnerbility of production wells in the Potomac-Raritan-Magothy aquifer system to saltwater intrusion from the Delaware River in Camden, Gloucester, and Salem Counties, New Jersey

USGS Publications Warehouse

Navoy, Anthony S.; Voronin, Lois M.; Modica, Edward

2005-01-01

The Potomac-Raritan-Magothy aquifer system is hydraulically connected to the Delaware River in parts of Camden and Gloucester Counties, New Jersey, and has more limited contact with the river in Salem County, New Jersey. The aquifer system is used widely for water supply, and 122 production wells that are permitted by the New Jersey Department of Environmental Protection to pump more than 100,000 gallons per year in the three counties are within 2 miles of the river. During drought, saltwater may encroach upstream from the Atlantic Ocean and Delaware Bay to areas where the aquifer system is recharged by induced infiltration through the Delaware River streambed. During the drought of the mid-1960's, water with a chloride concentration in excess of potability standards (250 mg/L (milligrams per liter)) encroached into the reach of the river that recharges the aquifer system. The vulnerability of the major production wells in the area to similar saltwater encroachment in the future is a concern to water managers. This vulnerability was evaluated by investigating two scenarios: (1) a one-time recurrence of the conditions approximating those that occurred in the1960's, and (2) the recurrence of those same conditions on an annual basis. Results of ground-water-flow simulation in conjunction with particle tracking and one-dimensional transport analysis indicate that the wells that are most vulnerable to saltwater intrusion are those in the Morris and Delair well fields in Camden County. A single 30-day event during which the concentration of dissolved chloride or sodium exceeds 2,098 mg/L or 407 mg/L, respectively, in the Delaware River would threaten the potability of water from these wells, given New Jersey drinking-water standards of 250 mg/L for dissolved chloride and 50 mg/L for dissolved sodium. This chloride concentration is about six times that observed in the river during the 1960's drought. An annually occurring 1-month event during which the concentrations of

18. Photocopy of scaled drawing (from the Engineering Office of ...

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

18. Photocopy of scaled drawing (from the Engineering Office of the Veterans Administration) originally drawn by Trone, November 9, 1933. Final revisions date December 10, 1957. Basement plan. - Roanoke Veterans Administration Hospital, Building No. 6, 1970 Roanoke Boulevard, Salem, Salem, VA

16. Photocopy of scaled drawing (from the Engineering Office of ...

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

16. Photocopy of scaled drawing (from the Engineering Office of the Veterans Administration) drawn by Ayars, November 9, 1933. Original drawings of the northeast and southwest elevations. - Roanoke Veterans Administration Hospital, Building No. 6, 1970 Roanoke Boulevard, Salem, Salem, VA

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.

Effect of temperature on selenium removal from wastewater by UASB reactors.

PubMed

Dessì, Paolo; Jain, Rohan; Singh, Satyendra; Seder-Colomina, Marina; van Hullebusch, Eric D; Rene, Eldon R; Ahammad, Shaikh Ziauddin; Carucci, Alessandra; Lens, Piet N L

2016-05-01

The effect of temperature on selenium (Se) removal by upflow anaerobic sludge blanket (UASB) reactors treating selenate and nitrate containing wastewater was investigated by comparing the performance of a thermophilic (55 °C) versus a mesophilic (30 °C) UASB reactor. When only selenate (50 μM) was fed to the UASB reactors (pH 7.3; hydraulic retention time 8 h) with excess electron donor (lactate at 1.38 mM corresponding to an organic loading rate of 0.5 g COD L(-1) d(-1)), the thermophilic UASB reactor achieved a higher total Se removal efficiency (94.4 ± 2.4%) than the mesophilic UASB reactor (82.0 ± 3.8%). When 5000 μM nitrate was further added to the influent, total Se removal was again better under thermophilic (70.1 ± 6.6%) when compared to mesophilic (43.6 ± 8.8%) conditions. The higher total effluent Se concentration in the mesophilic UASB reactor was due to the higher concentrations of biogenic elemental Se nanoparticles (BioSeNPs). The shape of the BioSeNPs observed in both UASB reactors was different: nanospheres and nanorods, respectively, in the mesophilic and thermophilic UASB reactors. Microbial community analysis showed the presence of selenate respirers as well as denitrifying microorganisms. Copyright © 2016 Elsevier Ltd. All rights reserved.

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear... requirements for immediate notification of the NRC by licensed operating nuclear power reactors are contained...

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)

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Amounts of financial protection for certain reactors. 140... reactors. (a) Each licensee is required to have and maintain financial protection: (1) In the amount of $1,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding...

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Amounts of financial protection for certain reactors. 140... reactors. (a) Each licensee is required to have and maintain financial protection: (1) In the amount of $1,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding...

Mapping of groundwater potential zones in Salem Chalk Hills, Tamil Nadu, India, using remote sensing and GIS techniques.

PubMed

Thilagavathi, N; Subramani, T; Suresh, M; Karunanidhi, D

2015-04-01

This study proposes to introduce the remote sensing and geographic information system (GIS) techniques in mapping the groundwater potential zones. Remote sensing and GIS techniques have been used to map the groundwater potential zones in Salem Chalk Hills, Tamil Nadu, India. Charnockites and fissile hornblende biotite gneiss are the major rock types in this region. Dunites and peridodites are the ultramafic rocks which cut across the foliation planes of the gneisses and are highly weathered. It comprises magnesite and chromite deposits which are excavated by five mining companies by adopting bench mining. The thickness of weathered and fracture zone varies from 2.2 to 50 m in gneissic formation and 5.8 to 55 m in charnockite. At the contacts of gneiss and charnockite, the thickness ranges from 9.0 to 90.8 m favoring good groundwater potential. The mine lease area is underlined by fractured and sheared hornblende biotite gneiss where groundwater potential is good. Water catchment tanks in this area of 5 km radius are small to moderate in size and are only seasonal. They remain dry during summer seasons. As perennial water resources are remote, the domestic and agricultural activities in this region depend mainly upon the groundwater resources. The mines are located in gently slope area, and accumulation of water is not observed except in mine pits even during the monsoon period. Therefore, it is essential to map the groundwater potential zones for proper management of the aquifer system. Satellite imageries were also used to extract lineaments, hydrogeomorphic landforms, drainage patterns, and land use, which are the major controlling factors for the occurrence of groundwater. Various thematic layers pertaining to groundwater existence such as geology, geomorphology, land use/land cover, lineament, lineament density, drainage, drainage density, slope, and soil were generated using GIS tools. By integrating all the above thematic layers based on the ranks and

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 %).

Characterization of cartridge filters from the IEA-R1 Nuclear Reactor

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

NONE

The management of radioactive waste ensures safety to human health and the environment nowadays and for the future, without overwhelming the upcoming generations. The primary characterization of radioactive waste is one of the main steps in the management of radioactive waste. This step permits to choose the best treatment for the radioactive waste before forwarding it to its final disposal. The aim of the present work is the primary characterization of cartridge filters from the IEA-R1 nuclear reactor utilizing gamma-ray spectrometry, and the method of Monte Carlo for calibration. The IEA-R1 is located in the Nuclear and Energy Research Institutemore » (IPEN - CNEN) in the city of Sao Paulo, Brazil. Cartridge filters are used for purification of the cooling water that is pumped through the core of the pool type nuclear research reactors. Once worn out, these filters are replaced and then become radioactive waste. Determination of the radioactive inventory is of paramount importance in the management of such radioactive waste, and one of the main methods for doing so is the gamma-ray spectrometry, which can identify and quantify high energy photon emitters. The technique chosen for the characterization of radioactive waste in the present work is the gamma-ray spectrometry with High purity Germanium (HPGe) detectors. From the energy identified in the experimental spectrum, three radioisotopes were identified in the cartridge filter: {sup 108m}Ag, {sup 110m}Ag, {sup 60}Co. For the estimated activity of the filter, the calibration in efficiency was made utilizing the MCNP4C code of the Monte Carlo method. Such method was chosen because there is no standard source available in the same geometry of the cartridge filter, therefore a simulation had to be developed in order to reach a calibration equation, necessary to estimate the activity of the radioactive waste. The results presented an activity value in the order of MBq for all radioisotopes. (authors)« less

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.

19. Photocopy of scaled drawing (from the Engineering Office of ...

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

19. Photocopy of scaled drawing (from the Engineering Office of the Veterans Administration) originally drawn by Trone, November 9, 1933. Final revisions date December 9, 1957. First floor plan. - Roanoke Veterans Administration Hospital, Building No. 6, 1970 Roanoke Boulevard, Salem, Salem, VA

20. Photocopy of scaled drawing (from the Engineering Office of ...

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

20. Photocopy of scaled drawing (from the Engineering Office of the Veterans Administration) originally drawn by Trone, November 9, 1933. Final revisions date December 6, 1957. Second floor plan. - Roanoke Veterans Administration Hospital, Building No. 6, 1970 Roanoke Boulevard, Salem, Salem, VA

77 FR 2048 - Procurement List Proposed Additions

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-13

..., Vandenberg AFB, CA. Service Type/Location: Base Supply Center, National Maritime Intelligence Center/Office of Naval Intelligence, 4251 Suitland Road, Suitland, MD. NPA: Winston-Salem Industries for the Blind, Inc., Winston-Salem, NC. Contracting Activity: Dept. of the Navy, Office of Naval Intelligence...

Membrane-aerated biofilm reactor for the removal of 1,2-dichloroethane by Pseudomonas sp. strain DCA1.

PubMed

Hage, J C; Van Houten, R T; Tramper, J; Hartmans, S

2004-06-01

A membrane-aerated biofilm reactor (MBR) with a biofilm of Pseudomonas sp. strain DCA1 was studied for the removal of 1,2-dichloroethane (DCA) from water. A hydrophobic membrane was used to create a barrier between the liquid and the gas phase. Inoculation of the MBR with cells of strain DCA1 grown in a continuous culture resulted in the formation of a stable and active DCA-degrading biofilm on the membrane. The maximum removal rate of the MBR was reached at a DCA concentration of approximately 80 micro M. Simulation of the DCA fluxes into the biofilm showed that the MBR performance at lower concentrations was limited by the DCA diffusion rate rather than by kinetic constraints of strain DCA1. Aerobic biodegradation of DCA present in anoxic water could be achieved by supplying oxygen solely from the gas phase to the biofilm grown on the liquid side of the membrane. As a result, direct aeration of the water, which leads to undesired coagulation of iron oxides, could be avoided.

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.

Indicators of AEI applied to the Delaware Estuary.

PubMed

Barnthouse, Lawrence W; Heimbuch, Douglas G; Anthony, Vaughn C; Hilborn, Ray W; Myers, Ransom A

2002-05-18

We evaluated the impacts of entrainment and impingement at the Salem Generating Station on fish populations and communities in the Delaware Estuary. In the absence of an agreed-upon regulatory definition of "adverse environmental impact" (AEI), we developed three independent benchmarks of AEI based on observed or predicted changes that could threaten the sustainability of a population or the integrity of a community. Our benchmarks of AEI included: (1) disruption of the balanced indigenous community of fish in the vicinity of Salem (the "BIC" analysis); (2) a continued downward trend in the abundance of one or more susceptible fish species (the "Trends" analysis); and (3) occurrence of entrainment/impingement mortality sufficient, in combination with fishing mortality, to jeopardize the future sustainability of one or more populations (the "Stock Jeopardy" analysis). The BIC analysis utilized nearly 30 years of species presence/absence data collected in the immediate vicinity of Salem. The Trends analysis examined three independent data sets that document trends in the abundance of juvenile fish throughout the estuary over the past 20 years. The Stock Jeopardy analysis used two different assessment models to quantify potential long-term impacts of entrainment and impingement on susceptible fish populations. For one of these models, the compensatory capacities of the modeled species were quantified through meta-analysis of spawner-recruit data available for several hundred fish stocks. All three analyses indicated that the fish populations and communities of the Delaware Estuary are healthy and show no evidence of an adverse impact due to Salem. Although the specific models and analyses used at Salem are not applicable to every facility, we believe that a weight of evidence approach that evaluates multiple benchmarks of AEI using both retrospective and predictive methods is the best approach for assessing entrainment and impingement impacts at existing facilities.

Extending the maximum operation time of the MNSR reactor.

PubMed

Dawahra, S; Khattab, K; Saba, G

2016-09-01

An effective modification to extend the maximum operation time of the Miniature Neutron Source Reactor (MNSR) to enhance the utilization of the reactor has been tested using the MCNP4C code. This modification consisted of inserting manually in each of the reactor inner irradiation tube a chain of three polyethylene-connected containers filled of water. The total height of the chain was 11.5cm. The replacement of the actual cadmium absorber with B(10) absorber was needed as well. The rest of the core structure materials and dimensions remained unchanged. A 3-D neutronic model with the new modifications was developed to compare the neutronic parameters of the old and modified cores. The results of the old and modified core excess reactivities (ρex) were: 3.954, 6.241 mk respectively. The maximum reactor operation times were: 428, 1025min and the safety reactivity factors were: 1.654 and 1.595 respectively. Therefore, a 139% increase in the maximum reactor operation time was noticed for the modified core. This increase enhanced the utilization of the MNSR reactor to conduct a long time irradiation of the unknown samples using the NAA technique and increase the amount of radioisotope production in the reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

Comparative performance of fixed-film biological filters: Application of reactor theory

USGS Publications Warehouse

Watten, B.J.; Sibrell, P.L.

2006-01-01

Nitrification is classified as a two-step consecutive reaction where R1 represents the rate of formation of the intermediate product NO2-N and R2 represents the rate of formation of the final product NO3-N. The relative rates of R1 and R2 are influenced by reactor type characterized hydraulically as plug-flow, plug-flow with dispersion and mixed-flow. We develop substrate conversion models for fixed-film biofilters operating in the first-order kinetic regime based on application of chemical reactor theory. Reactor type, inlet conditions and the biofilm kinetic constants Ki (h-1) are used to predict changes in NH4-N, NO2-N, NO3-N and BOD5. The inhibiting effects of the latter on R1 and R2 were established based on the ?? relation, e.g.:{A formula is presented}where BOD5,max is the concentration that causes nitrification to cease and N is a variable relating Ki to increasing BOD5. Conversion models were incorporated in spreadsheet programs that provided steady-state concentrations of nitrogen and BOD5 at several points in a recirculating aquaculture system operating with input values for fish feed rate, reactor volume, microscreen performance, make-up and recirculating flow rates. When rate constants are standardized, spreadsheet use demonstrates plug-flow reactors provide higher rates of R1 and R2 than mixed-flow reactors thereby reducing volume requirements for target concentrations of NH4-N and NO2-N. The benefit provided by the plug-flow reactor varies with hydraulic residence time t as well as the effective vessel dispersion number, D/??L. Both reactor types are capable of providing net increases in NO2-N during treatment but the rate of decrease in the mixed-flow case falls well behind that predicted for plug-flow operation. We show the potential for a positive net change in NO2-N increases with decreases in the dimensionless ratios K2, (R2 )/K1,( R1 ) and [NO2-N]/[NH4-N] and when the product K1, (R1) t provides low to moderate NH4-N conversions. Maintaining

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

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...

Reactor Monitoring with Antineutrinos - A Progress Report

NASA Astrophysics Data System (ADS)

Bernstein, Adam

2012-08-01

The Reactor Safeguards regime is the name given to a set of protocols and technologies used to monitor the consumption and production of fissile materials in nuclear reactors. The Safeguards regime is administered by the International Atomic Energy Agency (IAEA), and is an essential component of the global Treaty on Nuclear Nonproliferation, recently renewed by its 189 remaining signators. (The 190th, North Korea, withdrew from the Treaty in 2003). Beginning in Russia in the 1980s, a number of researchers worldwide have experimentally demonstrated the potential of cubic meter scale antineutrino detectors for non-intrusive real-time monitoring of fissile inventories and power output of reactors. The detectors built so far have operated tens of meters from a reactor core, outside of the containment dome, largely unattended and with remote data acquisition for an entire 1.5 year reactor cycle, and have achieved levels of sensitivity to fissile content of potential interest for the IAEA safeguards regime. In this article, I will describe the unique advantages of antineutrino detectors for cooperative monitoring, consider the prospects and benefits of increasing the range of detectability for small reactors, and provide a partial survey of ongoing global research aimed at improving near-field and far field monitoring and discovery of nuclear reactors.

PARTIAL ECONOMIC STUDY OF STEAM COOLED HEAVY WATER MODERATED REACTORS

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

None

1960-04-01

Steam-cooled reactors are compared with CAHDU for costs of Calandria tubes, pressure tubes. heavy water moderator, heavy water reflector, fuel supply, heat exchanger, and turbine generator. A direct-cycle lightsteam-cooled heavy- water-moderated pressure-tube reactor formed the basic reactor design for the study. Two methods of steam circulation through the reactor were examined. In both cases the steam was generated outside the reactor and superheated in the reactor core. One method consisted of a series of reactor and steam generator passes. The second method consisted of the Loeffler cycle and its modifications. The fuel was assumed to be natural cylindrical UO/sub 2/more » pellets sheathed in a hypothetical material with the nuclear properties of Zircaloy, but able to function at temperatures to 900 deg F. For the conditions assumed, the longer the rod, the higher the outlet temperature and therefore the higher the efficiency. The turbine cycle efficiency was calculated on the assumption that suitable steam generators are available. As the neutron losses to the pressure tubes were significant, an economic analysis of insulated pressure tubes is included. A description of the physics program for steam-cooled reactors is included. Results indicated that power from the steam-cooled reactor would cost 1.4 mills/ kwh compared with 1.25 mills/kwh for CANDU. (M.C.G.)« less

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.

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear power reactor licensee licensed under §§ 50.21(b) or 50.22 holding an operating license under this part...

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear power reactor licensee licensed under §§ 50.21(b) or 50.22 holding an operating license under this part...

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.

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.

75 FR 14575 - Procurement List; Proposed Additions

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-26

... Cost Container. NPA: Winston-Salem Industries for the Blind, Winston-Salem, NC. Contracting Activity... the Department of the Army, Natick, MA. Tape, Insulation, Electrical NSN: 5970-00-240-0617. NSN: 5970... Base, Warner Robins, GA. NPA: Georgia Industries for the Blind, Bainbridge, GA. Contracting Activity...

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.

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

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)

Determination of neutron spectra within the energy of 1 keV to 1 MeV by means of reactor dosimetry

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

Sergeyeva, Victoria; Destouches, Christophe; Lyoussi, Abdallah

2015-07-01

The standard procedure for neutron reactor dosimetry is based on neutron irradiation of a target and its post-irradiation analysis by Gamma and/or X-ray spectrometry. Nowadays, the neutron spectra can be easily characterized for thermal and fast energies (respectively 0.025 eV and >1 MeV). In this work we propose a new target and an innovating post-irradiation technique of analysis in order to detect the neutron spectra within the energy of 1 keV to 1 MeV. This article will present the calculations performed for the selection of a suitable nuclear reaction and isotope, the results predicted by simulations, the irradiation campaign thatmore » is proposed and the post-irradiation technique of analysis. (authors)« less

Nuclear reactors built, being built, or planned 1993

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

Not Available

1993-08-01

Nuclear Reactors Built, Being Built, or Planned contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1993. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical datamore » that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: civilian, production, military, export, and critical assembly.« less

Boiling-Water Reactor internals aging degradation study. Phase 1

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

Luk, K.H.

1993-09-01

This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor drymore » tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR.« less

NETL - Chemical Looping Reactor

ScienceCinema

None

2018-02-14

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

Degradation of Acid Orange 7 Dye in Two Hybrid Plasma Discharge Reactors

NASA Astrophysics Data System (ADS)

Shen, Yongjun; Lei, Lecheng; Zhang, Xingwang; Ding, Jiandong

2014-11-01

To get an optimized pulsed electrical plasma discharge reactor and to increase the energy utilization efficiency in the removal of pollutants, two hybrid plasma discharge reactors were designed and optimized. The reactors were compared via the discharge characteristics, energy transfer efficiency, the yields of the active species and the energy utilization in dye wastewater degradation. The results showed that under the same AC input power, the characteristics of the discharge waveform of the point-to-plate reactor were better. Under the same AC input power, the two reactors both had almost the same peak voltage of 22 kV. The peak current of the point-to-plate reactor was 146 A, while that of the wire-to-cylinder reactor was only 48.8 A. The peak powers of the point-to-plate reactor and the wire-to-cylinder reactor were 1.38 MW and 1.01 MW, respectively. The energy per pulse of the point-to-plate reactor was 0.2221 J, which was about 29.4% higher than that of the wire-to-cylinder reactor (0.1716 J). To remove 50% Acid Orange 7 (AO7), the energy utilizations of the point-to-plate reactor and the wire-to-cylinder reactor were 1.02 × 10-9 mol/L and 0.61 × 10-9 mol/L, respectively. In the point-to-plate reactor, the concentration of hydrogen peroxide in pure water was 3.6 mmol/L after 40 min of discharge, which was higher than that of the wire-to-cylinder reactor (2.5 mmol/L). The concentration of liquid phase ozone in the point-to-plate reactor (5.7 × 10-2 mmol/L) was about 26.7% higher than that in the wire-to-cylinder reactor (4.5 × 10-2 mmol/L). The analysis results of the variance showed that the type of reactor and reaction time had significant impacts on the yields of the hydrogen peroxide and ozone. The main degradation intermediates of AO7 identified by gas chromatography and mass spectrometry (GCMS) were acetic acid, maleic anhydride, p-benzoquinone, phenol, benzoic acid, phthalic anhydride, coumarin and 2-naphthol. Proposed degradation pathways were

NUCLEAR REACTOR COOLANT

DOEpatents

Colichman, E.L.

1959-10-20

The formation of new reactor coolants which suppress polymerization resulting from pyrolitic and radiation decomposition is described. The coolants consist of polyphenyls and condensed ring compounds having from two to about four carbon rings and from 0.1 to about 10% of an alkall metal dispersed in the hydrocarbon.

NUCLEAR REACTOR COOLANT

DOEpatents

Colichman, E.L.

1959-10-20

The formation of new reactor coolants which suppress polymerization resulting from pyrolytic and radiation decomposition is described. The coolants consist of polyphenyls and condensed ring compounds having from two to about four carbon rings and from 0.1 to about 5% of beryllium or magnesium dispersed in the hydrocarbon.

Laboratory instrumentation modernization at the WPI Nuclear Reactor Facility

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

Not Available

1995-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Program several laboratory instruments utilized by students and researchers at the WPI Nuclear Reactor Facility have been upgraded or replaced. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduate use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The low power output of the reactor and an ergonomicmore » facility design make it an ideal tool for undergraduate nuclear engineering education and other training. The reactor, its control system, and the associate laboratory equipment are all located in the same room. Over the years, several important milestones have taken place at the WPI reactor. In 1969, the reactor power level was upgraded from 1 kW to 10 kW. The reactor`s Nuclear Regulatory Commission operating license was renewed for 20 years in 1983. In 1988, under DOE Grant No. DE-FG07-86ER75271, the reactor was converted to low-enriched uranium fuel. In 1992, again with partial funding from DOE (Grant No. DE-FG02-90ER12982), the original control console was replaced.« less

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.

10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems required...

10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems required...

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.

Two conceptual designs of helical fusion reactor FFHR-d1A based on ITER technologies and challenging ideas

NASA Astrophysics Data System (ADS)

Sagara, A.; Miyazawa, J.; Tamura, H.; Tanaka, T.; Goto, T.; Yanagi, N.; Sakamoto, R.; Masuzaki, S.; Ohtani, H.; The FFHR Design Group

2017-08-01

The Fusion Engineering Research Project (FERP) at the National Institute for Fusion Science (NIFS) is conducting conceptual design activities for the LHD-type helical fusion reactor FFHR-d1A. This paper newly defines two design options, ‘basic’ and ‘challenging.’ Conservative technologies, including those that will be demonstrated in ITER, are chosen in the basic option in which two helical coils are made of continuously wound cable-in-conduit superconductors of Nb3Sn strands, the divertor is composed of water-cooled tungsten monoblocks, and the blanket is composed of water-cooled ceramic breeders. In contrast, new ideas that would possibly be beneficial for making the reactor design more attractive are boldly included in the challenging option in which the helical coils are wound by connecting high-temperature REBCO superconductors using mechanical joints, the divertor is composed of a shower of molten tin jets, and the blanket is composed of molten salt FLiNaBe including Ti powers to increase hydrogen solubility. The main targets of the challenging option are early construction and easy maintenance of a large and three-dimensionally complicated helical structure, high thermal efficiency, and, in particular, realistic feasibility of the helical reactor.

Nuclear reactors built, being built, or planned: 1995

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

NONE

1996-08-01

This report contains unclassified information about facilities built, being built, or planned in the US for domestic use or export as of December 31, 1995. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristicmore » and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company--working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).« 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

17. Photocopy of scaled drawing (from the Engineering Office of ...

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

17. Photocopy of scaled drawing (from the Engineering Office of the Veterans Administration) drawn by Ayars, November 9, 1933. Original drawings of the northwest and southwest elevations with details of roof and lobbies. - Roanoke Veterans Administration Hospital, Building No. 6, 1970 Roanoke Boulevard, Salem, Salem, VA

Micro-reactors for characterization of nanostructure-based sensors.

PubMed

Savu, R; Silveira, J V; Flacker, A; Vaz, A R; Joanni, E; Pinto, A C; Gobbi, A L; Santos, T E A; Rotondaro, A L P; Moshkalev, S A

2012-05-01

Fabrication and testing of micro-reactors for the characterization of nanosensors is presented in this work. The reactors have a small volume (100 μl) and are equipped with gas input/output channels. They were machined from a single piece of kovar in order to avoid leaks in the system due to additional welding. The contact pins were electrically insulated from the body of the reactor using a borosilicate sealing glass and the reactor was hermetically sealed using a lid and an elastomeric o-ring. One of the advantages of the reactor lies in its simple assembly and ease of use with any vacuum/gas system, allowing the connection of more than one device. Moreover, the lid can be modified in order to fit a window for in situ optical characterization. In order to prove its versatility, carbon nanotube-based sensors were tested using this micro-reactor. The devices were fabricated by depositing carbon nanotubes over 1 μm thick gold electrodes patterned onto Si/SiO(2) substrates. The sensors were tested using oxygen and nitrogen atmospheres, in the pressure range between 10(-5) and 10(-1) mbar. The small chamber volume allowed the measurement of fast sensor characteristic times, with the sensors showing good sensitivity towards gas and pressure as well as high reproducibility.

Micro-reactors for characterization of nanostructure-based sensors

NASA Astrophysics Data System (ADS)

Savu, R.; Silveira, J. V.; Flacker, A.; Vaz, A. R.; Joanni, E.; Pinto, A. C.; Gobbi, A. L.; Santos, T. E. A.; Rotondaro, A. L. P.; Moshkalev, S. A.

2012-05-01

Fabrication and testing of micro-reactors for the characterization of nanosensors is presented in this work. The reactors have a small volume (100 μl) and are equipped with gas input/output channels. They were machined from a single piece of kovar in order to avoid leaks in the system due to additional welding. The contact pins were electrically insulated from the body of the reactor using a borosilicate sealing glass and the reactor was hermetically sealed using a lid and an elastomeric o-ring. One of the advantages of the reactor lies in its simple assembly and ease of use with any vacuum/gas system, allowing the connection of more than one device. Moreover, the lid can be modified in order to fit a window for in situ optical characterization. In order to prove its versatility, carbon nanotube-based sensors were tested using this micro-reactor. The devices were fabricated by depositing carbon nanotubes over 1 μm thick gold electrodes patterned onto Si/SiO2 substrates. The sensors were tested using oxygen and nitrogen atmospheres, in the pressure range between 10-5 and 10-1 mbar. The small chamber volume allowed the measurement of fast sensor characteristic times, with the sensors showing good sensitivity towards gas and pressure as well as high reproducibility.

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...

Results on the neutron energy distribution measurements at the RECH-1 Chilean nuclear reactor

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

Aguilera, P., E-mail: paguilera87@gmail.com; Romero-Barrientos, J.; Universidad de Chile, Dpto. de Física, Facultad de Ciencias, Las Palmeras 3425, Nuñoa, Santiago

2016-07-07

Neutron activations experiments has been perform at the RECH-1 Chilean Nuclear Reactor to measure its neutron flux energy distribution. Samples of pure elements was activated to obtain the saturation activities for each reaction. Using - ray spectroscopy we identify and measure the activity of the reaction product nuclei, obtaining the saturation activities of 20 reactions. GEANT4 and MCNP was used to compute the self shielding factor to correct the cross section for each element. With the Expectation-Maximization algorithm (EM) we were able to unfold the neutron flux energy distribution at dry tube position, near the RECH-1 core. In this work,more » we present the unfolding results using the EM algorithm.« less

Convective cooling in a pool-type research reactor

NASA Astrophysics Data System (ADS)

Sipaun, Susan; Usman, Shoaib

2016-01-01

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

Experiences in utilization of research reactors in Yugoslavia

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

Copic, M.; Gabrovsek, Z.; Pop-Jordanov, J.

1971-06-15

The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied bymore » means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro

Performance of intermittent aeration reactor on NH4-N removal from groundwater resources.

PubMed

Khanitchaidecha, W; Nakamura, T; Sumino, T; Kazama, F

2010-01-01

To study the effect of intermittent aeration period on ammonium-nitrogen (NH4-N) removal from groundwater resources, synthetic groundwater was prepared and three reactors were operated under different conditions--"reactor A" under continuous aeration, "reactor B" under 6 h intermittent aeration, and "reactor C" under 2 h intermittent aeration. To facilitate denitrification simultaneously with nitrification, "acetate" was added as an external carbon source with step-wise increase from 0.5 to 1.5 C/N ratio, where C stands for total carbon content in the system, and N for NH4-N concentration in the synthetic groundwater. Results show that complete NH4-N removal was obtained in "reactor B" and "reactor C" at 1.3 and 1.5 C/N ratio respectively; and partial NH4-N removal in "reactor A". These results suggest that intermittent aeration at longer interval could enhance the reactor performance on NH4-N removal in terms of efficiency and low external carbon requirement. Because of consumption of internal carbon by the process, less amount of external carbon is required. Further increase in carbon in a form of acetate (1.5 to 2.5 C/N ratios) increases removal rate (represented by reaction rate coefficient (k) of kinetic equation) as well as occurrence of free cells. It suggests that the operating condition at reactor B with 1.3 C/N ratio is more appropriate for long-term operation at a pilot-scale.

Measurement and Analysis of Structural Integrity of Reactor Core Support Structure in Pressurized Water Reactor (PWR) Plant

NASA Astrophysics Data System (ADS)

Ansari, Saleem A.; Haroon, Muhammad; Rashid, Atif; Kazmi, Zafar

2017-02-01

Extensive calculation and measurements of flow-induced vibrations (FIV) of reactor internals were made in a PWR plant to assess the structural integrity of reactor core support structure against coolant flow. The work was done to meet the requirements of the Fukushima Response Action Plan (FRAP) for enhancement of reactor safety, and the regulatory guide RG-1.20. For the core surveillance measurements the Reactor Internals Vibration Monitoring System (IVMS) has been developed based on detailed neutron noise analysis of the flux signals from the four ex-core neutron detectors. The natural frequencies, displacement and mode shapes of the reactor core barrel (CB) motion were determined with the help of IVMS. The random pressure fluctuations in reactor coolant flow due to turbulence force have been identified as the predominant cause of beam-mode deflection of CB. The dynamic FIV calculations were also made to supplement the core surveillance measurements. The calculational package employed the computational fluid dynamics, mode shape analysis, calculation of power spectral densities of flow & pressure fields and the structural response to random flow excitation forces. The dynamic loads and stiffness of the Hold-Down Spring that keeps the core structure in position against upward coolant thrust were also determined by noise measurements. Also, the boron concentration in primary coolant at any time of the core cycle has been determined with the IVMS.

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)

77 FR 64563 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-22

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In... Advisory Committee on Reactor Safeguards (ACRS) will hold a meeting on November 1-3, 2012, 11545 Rockville...-Term Core Cooling Approach for the Advanced Boiling Water Reactor (ABWR) Design for South Texas Project...

Airport Activity Statistics of Certificated Route Air Carriers.

DTIC Science & Technology

1985-12-31

Houlton, ME............ Presque Isle /Houlton Bluefield, WV........... Princeon/Bhzefield Huntington, WV......... Ashland/ Huntington Bonn, Germany...Benito Wheeling, WV ......... Pttaburgh/ Wheeling (Under Pennsylvania ) San Bernardino, CA ... Ontario & San Bernardino White River Junction, VT... Pennsylvania ) Spartanburg, SC........ Greenville/Spartanbury Winston-Salem, NC ... Greensboro/High Point/Winston-Salem Springfield, MA......... Hartford

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

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.

Flow rate analysis of wastewater inside reactor tanks on tofu wastewater treatment plant

NASA Astrophysics Data System (ADS)

Mamat; Sintawardani, N.; Astuti, J. T.; Nilawati, D.; Wulan, D. R.; Muchlis; Sriwuryandari, L.; Sembiring, T.; Jern, N. W.

2017-03-01

The research aimed to analyse the flow rate of the wastewater inside reactor tanks which were placed a number of bamboo cutting. The resistance of wastewater flow inside reactor tanks might not be occurred and produce biogas fuel optimally. Wastewater from eleven tofu factories was treated by multi-stages anaerobic process to reduce its organic pollutant and produce biogas. Biogas plant has six reactor tanks of which its capacity for waste water and gas dome was 18 m3 and 4.5 m3, respectively. Wastewater was pumped from collecting ponds to reactors by either serial or parallel way. Maximum pump capacity, head, and electrical motor power was 5m3/h, 50m, and 0.75HP, consecutively. Maximum pressure of biogas inside the reactor tanks was 55 mbar higher than atmosphere pressure. A number of 1,400 pieces of cutting bamboo at 50-60 mm diameter and 100 mm length were used as bacteria growth media inside each reactor tank, covering around 14,287 m2 bamboo area, and cross section area of inner reactor was 4,9 m2. In each reactor, a 6 inches PVC pipe was installed vertically as channel. When channels inside reactor were opened, flow rate of wastewater was 6x10-1 L.sec-1. Contrary, when channels were closed on the upper part, wastewater flow inside the first reactor affected and increased gas dome. Initially, wastewater flowed into each reactor by a gravity mode with head difference between the second and third reactor was 15x10-2m. However, head loss at the second reactor was equal to the third reactor by 8,422 x 10-4m. As result, wastewater flow at the second and third reactors were stagnant. To overcome the problem pump in each reactor should be installed in serial mode. In order to reach the output from the first reactor and the others would be equal, and biogas space was not filled by wastewater, therefore biogas production will be optimum.

Neutronic reactor construction

DOEpatents

Huston, Norman E.

1976-07-06

1. A neutronic reactor comprising a moderator including horizontal layers formed of horizontal rows of graphite blocks, alternate layers of blocks having the rows extending in one direction, the remaining alternate layers having the rows extending transversely to the said one direction, alternate rows of blocks in one set of alternate layers having longitudinal ducts, the moderator further including slotted graphite tubes positioned in the ducts, the reactor further comprising an aluminum coolant tube positioned within the slotted tube in spaced relation thereto, bodies of thermal-neutron-fissionable material, and jackets enclosing the bodies and being formed of a corrosion-resistant material having a low neutron-capture cross section, the bodies and jackets being positioned within the coolant tube so that the jackets are spaced from the coolant tube.

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.

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.

Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

NASA Astrophysics Data System (ADS)

Hamzah, Amir; Suwoto; Rohanda, Anis; Adrial, Hery; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

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.

Research and proposal on selective catalytic reduction reactor optimization for industrial boiler.

PubMed

Yang, Yiming; Li, Jian; He, Hong

2017-08-24

The advanced computational fluid dynamics (CFD) software STAR-CCM+ was used to simulate a denitrification (De-NOx) project for a boiler in this paper, and the simulation result was verified based on a physical model. Two selective catalytic reduction (SCR) reactors were developed: reactor 1 was optimized and reactor 2 was developed based on reactor 1. Various indicators, including gas flow field, ammonia concentration distribution, temperature distribution, gas incident angle, and system pressure drop were analyzed. The analysis indicated that reactor 2 was of outstanding performance and could simplify developing greatly. Ammonia injection grid (AIG), the core component of the reactor, was studied; three AIGs were developed and their performances were compared and analyzed. The result indicated that AIG 3 was of the best performance. The technical indicators were proposed for SCR reactor based on the study. Flow filed distribution, gas incident angle, and temperature distribution are subjected to SCR reactor shape to a great extent, and reactor 2 proposed in this paper was of outstanding performance; ammonia concentration distribution is subjected to ammonia injection grid (AIG) shape, and AIG 3 could meet the technical indicator of ammonia concentration without mounting ammonia mixer. The developments above on the reactor and the AIG are both of great application value and social efficiency.

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.

Noncatalytic hydrogenation of decene-1 with hydrogen accumulated in a hybrid carbon nanostructure in nanosized membrane reactors

NASA Astrophysics Data System (ADS)

Soldatov, A. P.

2014-08-01

Studies on the creation of nanosized membrane reactors (NMRs) of a new generation with accumulated hydrogen and a regulated volume of reaction zone were continued at the next stage. Hydrogenation was performed in the pores of ceramic membranes with hydrogen preliminarily adsorbed in mono- and multilayered orientated carbon nanotubes with graphene walls (OCNTGs)—a new hybrid carbon nanostructure formed on the inner pore surface. Quantitative determination of hydrogen adsorption in OCNTGs was performed using TRUMEM ultrafiltration membranes with D av = 50 and 90 nm and showed that hydrogen adsorption was up to ˜1.5% of the mass of OCNTG. The instrumentation and procedure for noncatalytic hydrogenation of decene-1 at 250-350°C using hydrogen accumulated and stored in OCNTG were developed. The conversion of decene-1 into decane was ˜0.2-1.8% at hydrogenation temperatures of 250 and 350°C, respectively. The rate constants and activation energy of hydrogenation were determined. The latter was found to be 94.5 kJ/mol, which is much smaller than the values typical for noncatalytic hydrogenations and very close to the values characteristic for catalytic reactions. The quantitative distribution of the reacting compounds in each pore regarded as a nanosized membrane reactor was determined. The activity of hydrogen adsorbed in a 2D carbon nanostructure was evaluated. Possible mechanisms of noncatalytic hydrogenation were discussed.

Control console replacement at the WPI Reactor. [Final report

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

Not Available

1992-12-31

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

International Conference on Composite Structures (7th), Held in United Kingdom, July 1993. Volume 25, Numbers 1 - 4

DTIC Science & Technology

1993-07-01

the code number at the end of this paragraph. to the Copýright Clearance Center. 21 Congress Street. Salem . MA 01970. USA. If no code appears in an...failure always occurs at theo Selapex: the values of the critical loads are very low, It is in verv grood agreement with the theore- 0. PFa iAismn

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

Nuclear reactors built, being built, or planned 1996

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

NONE

1997-08-01

This publication contains unclassified information about facilities, built, being built, or planned in the United States for domestic use or export as of December 31, 1996. The Office of Scientific and Technical Information, U.S. Department of Energy, gathers this information annually from Washington headquarters, and field offices of DOE; from the U.S. Nuclear Regulatory Commission (NRC); from the U. S. reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from U.S. and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables ofmore » the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled.« less

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.

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.

Catalyzed D-D stellarator reactor

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

Sheffield, John; Spong, Donald A.

The advantages of using the catalyzed deuterium-deuterium (D-D) approach for a fusion reactor—lower and less energetic neutron flux and no need for a tritium breeding blanket—have been evaluated in previous papers, giving examples of both tokamak and stellarator reactors. This paper presents an update for the stellarator example, taking account of more recent empirical transport scaling results and design studies of lower-aspect-ratio stellarators. We use a modified version of the Generic Magnetic Fusion Reactor model to cost a stellarator-type reactor. Recently, this model has been updated to reflect the improved science and technology base and costs in the magnetic fusionmore » program. Furthermore, it is shown that an interesting catalyzed D-D, stellarator power plant might be possible if the following parameters could be achieved: R/ ≈ 4, required improvement factor to ISS04 scaling, F R = 0.9 to 1.15, ≈ 8.0% to 11.5%, Z eff ≈ 1.45 plus a relativistic temperature correction, fraction of fast ions lost ≈ 0.07, B m ≈ 14 to 16 T, and R ≈ 18 to 24 m.« less

Catalyzed D-D stellarator reactor

DOE PAGES

Sheffield, John; Spong, Donald A.

2016-05-12

The advantages of using the catalyzed deuterium-deuterium (D-D) approach for a fusion reactor—lower and less energetic neutron flux and no need for a tritium breeding blanket—have been evaluated in previous papers, giving examples of both tokamak and stellarator reactors. This paper presents an update for the stellarator example, taking account of more recent empirical transport scaling results and design studies of lower-aspect-ratio stellarators. We use a modified version of the Generic Magnetic Fusion Reactor model to cost a stellarator-type reactor. Recently, this model has been updated to reflect the improved science and technology base and costs in the magnetic fusionmore » program. Furthermore, it is shown that an interesting catalyzed D-D, stellarator power plant might be possible if the following parameters could be achieved: R/ ≈ 4, required improvement factor to ISS04 scaling, F R = 0.9 to 1.15, ≈ 8.0% to 11.5%, Z eff ≈ 1.45 plus a relativistic temperature correction, fraction of fast ions lost ≈ 0.07, B m ≈ 14 to 16 T, and R ≈ 18 to 24 m.« less

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

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.

NEUTRONIC REACTOR CONTROL ROD DRIVE APPARATUS

DOEpatents

Oakes, L.C.; Walker, C.S.

1959-12-15

ABS>A suspension mechanism between a vertically movable nuclear reactor control rod and a rod extension, which also provides information for the operator or an automatic control signal, is described. A spring connects the rod extension to a drive shift. The extension of the spring indicates whether (1) the rod is at rest on the reactor, (2) the rod and extension are suspended, or (3) the extension alone is suspended, the spring controlling a 3-position electrical switch.

Silicon production in a fluidized bed reactor

NASA Technical Reports Server (NTRS)

Rohatgi, N. K.

1986-01-01

Part of the development effort of the JPL in-house technology involved in the Flat-Plate Solar Array (FSA) Project was the investigation of a low-cost process to produce semiconductor-grade silicon for terrestrial photovoltaic cell applications. The process selected was based on pyrolysis of silane in a fluidized-bed reactor (FBR). Following initial investigations involving 1- and 2-in. diameter reactors, a 6-in. diameter, engineering-scale FBR was constructed to establish reactor performance, mechanism of silicon deposition, product morphology, and product purity. The overall mass balance for all experiments indicates that more than 90% of the total silicon fed into the reactor is deposited on silicon seed particles and the remaining 10% becomes elutriated fines. Silicon production rates were demonstrated of 1.5 kg/h at 30% silane concentration and 3.5 kg/h at 80% silane concentration. The mechanism of silicon deposition is described by a six-path process: heterogeneous deposition, homogeneous decomposition, coalescence, coagulation, scavenging, and heterogeneous growth on fines. The bulk of the growth silicon layer appears to be made up of small diameter particles. This product morphology lends support to the concept of the scavenging of homogeneously nucleated silicon.

15. Photocopy of scaled drawing (from the Engineering Office of ...

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

15. Photocopy of scaled drawing (from the Engineering Office of the Veterans Administration) drawn by A.B. Metcalf, issued February 5, 1935). Tree planting plan of R-VAH site, showing Building #6 in relation to its surroundings. - Roanoke Veterans Administration Hospital, Building No. 6, 1970 Roanoke Boulevard, Salem, Salem, VA

PBF Reactor Building (PER620). Camera faces north into highbay/reactor pit ...

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

PBF Reactor Building (PER-620). Camera faces north into high-bay/reactor pit area. Inside from for reactor enclosure is in place. Photographer: John Capek. Date: March 15, 1967. INEEL negative no. 67-1769 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

DR Reactor VSR channel damage

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

Kempf, F.J.; Rawlins, J.K.

1961-10-30

On July 11, 1961 the Ball 3X System at DR Reactor was inadventently tripped. All vertical safety rods dropped and all channels were filled with balls. This report has the twofold purpose of documenting borescope observations of ten vertical rod channels at DR Reactor and recording the estimated extent of graphite damage resulting from the above incident. Channel damage data are presented on appended drawings. With suitable notations, the tracings of these drawings may be revised to reflect any future graphite damage. All vertical rod channels at DR Reactor were visually examined with a closed circuit television system during ballmore » removal efforts. Typical photographs of trapped balls and ledges, as viewed on the television monitor, are shown. Photographs of typical graphite damage, obtained through the borescope are also included in this report. 3 refs., 8 figs., 1 tab.« less

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.

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.

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.

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.

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.

Convective cooling in a pool-type research reactor

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

Sipaun, Susan, E-mail: susan@nm.gov.my; Usman, Shoaib, E-mail: usmans@mst.edu

2016-01-22

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to passmore » through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.« less

Neutronic reactor thermal shield

DOEpatents

Wende, Charles W. J.

1976-06-15

1. The method of operating a water-cooled neutronic reactor having a graphite moderator which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40-60 volume percent of the mixture, in contact with the graphite moderator.

A PC-based high temperature gas reactor simulator for Indonesian conceptual HTR reactor basic training

NASA Astrophysics Data System (ADS)

Syarip; Po, L. C. C.

2018-05-01

In planning for nuclear power plant construction in Indonesia, helium cooled high temperature reactor (HTR) is favorable for not relying upon water supply that might be interrupted by earthquake. In order to train its personnel, BATAN has cooperated with Micro-Simulation Technology of USA to develop a 200 MWt PC-based simulation model PCTRAN/HTR. It operates in Win10 environment with graphic user interface (GUI). Normal operation of startup, power maneuvering, shutdown and accidents including pipe breaks and complete loss of AC power have been conducted. A sample case of safety analysis simulation to demonstrate the inherent safety features of HTR was done for helium pipe break malfunction scenario. The analysis was done for the variation of primary coolant pipe break i.e. from 0,1% - 0,5 % and 1% - 10 % helium gas leakages, while the reactor was operated at the maximum constant power of 10 MWt. The result shows that the highest temperature of HTR fuel centerline and coolant were 1150 °C and 1296 °C respectively. With 10 kg/s of helium flow in the reactor core, the thermal power will back to the startup position after 1287 s of helium pipe break malfunction.

Fischer-Tropsch Slurry Reactor modeling

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

Soong, Y.; Gamwo, I.K.; Harke, F.W.

1995-12-31

This paper reports experimental and theoretical results on hydrodynamic studies. The experiments were conducted in a hot-pressurized Slurry-Bubble Column Reactor (SBCR). It includes experimental results of Drakeol-10 oil/nitrogen/glass beads hydrodynamic study and the development of an ultrasonic technique for measuring solids concentration. A model to describe the flow behavior in reactors was developed. The hydrodynamic properties in a 10.16 cm diameter bubble column with a perforated-plate gas distributor were studied at pressures ranging from 0.1 to 1.36 MPa, and at temperatures from 20 to 200{degrees}C, using a dual hot-wire probe with nitrogen, glass beads, and Drakeol-10 oil as the gas,more » solid, and liquid phase, respectively. It was found that the addition of 20 oil wt% glass beads in the system has a slight effect on the average gas holdup and bubble size. A well-posed three-dimensional model for bed dynamics was developed from an ill-posed model. The new model has computed solid holdup distributions consistent with experimental observations with no artificial {open_quotes}fountain{close_quotes} as predicted by the earlier model. The model can be applied to a variety of multiphase flows of practical interest. An ultrasonic technique is being developed to measure solids concentration in a three-phase slurry reactor. Preliminary measurements have been made on slurries consisting of molten paraffin wax, glass beads, and nitrogen bubbles at 180 {degrees}C and 0.1 MPa. The data show that both the sound speed and attenuation are well-defined functions of both the solid and gas concentrations in the slurries. The results suggest possibilities to directly measure solids concentration during the operation of an autoclave reactor containing molten wax.« less

Enhanced biodegradation of hexachlorocyclohexane in upflow anaerobic sludge blanket reactor using methanol as an electron donor.

PubMed

Bhatt, Praveena; Kumar, M Suresh; Mudliar, Sandeep; Chakrabarti, Tapan

2008-05-01

Anaerobic dechlorination of technical grade hexachlorocyclohexane (THCH) was studied in a continuous upflow anaerobic sludge blanket (UASB) reactor with methanol as a supplementary substrate and electron donor. A reactor without methanol served as the experimental control. The inlet feed concentration of THCH in both the experimental and the control UASB reactor was 100 mg l(-1). After 60 days of continuous operation, the removal of THCH was >99% in the methanol-supplemented reactor as compared to 20-35% in the control reactor. THCH was completely dechlorinated in the methanol fed reactor at 48 h HRT after 2 months of continuous operation. This period was also accompanied by increase in biomass in the reactor, which was not observed in the experimental control. Batch studies using other supplementary substrates as well as electron donors namely acetate, butyrate, formate and ethanol showed lower % dechlorination (<85%) and dechlorination rates (<3 mg g(-1)d(-1)) as compared to methanol (98%, 5 mg g(-1)d(-1)). The optimum concentration of methanol required, for stable dechlorination of THCH (100 mg l(-1)) in the UASB reactor, was found to be 500 mg l(-1). Results indicate that addition of methanol as electron donor enhances dechlorination of THCH at high inlet concentration, and is also required for stable UASB reactor performance.

78 FR 24308 - Notice of Application for Special Permits

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-24

... 1, 2, 3, 4) 15838-N Primo Water, 49 CFR 171.2(k)........ To authorize the Winston-Salem, NC... amount qualifying as hazardous material. (modes 1, 2, 3, 4) 15839-N Sensors, Inc., 49 CFR 172.202, To...

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.

VERA Core Simulator methodology for pressurized water reactor cycle depletion

DOE PAGES

Kochunas, Brendan; Collins, Benjamin; Stimpson, Shane; ...

2017-01-12

This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclidemore » transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. Lastly, these results provide confidence in VERA-CS’s capability to perform high-fidelity calculations for practical PWR reactor problems.« less

Period meter for reactors

DOEpatents

Rusch, Gordon K.

1976-01-06

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

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.

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.

PBF Reactor Building (PER620). After lowering reactor vessel onto blocks, ...

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

PBF Reactor Building (PER-620). After lowering reactor vessel onto blocks, it is rolled on logs into PBF. Metal framework under vessel is handling device. Various penetrations in reactor bottom were for instrumentation, poison injection, drains. Large one, below center "manhole" was for primary coolant. Photographer: Larry Page. Date: February 13, 1970. INEEL negative no. 70-736 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

75 FR 26787 - National Register of Historic Places; Notification of Pending Nominations and Related Actions

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-12

... Lydia Plantation, 703 W Lydia Hwy (US HWY 15/SC HWY 34), Lydia, 10000299 VIRGINIA Campbell County... District, Roughly bounded by E and W Main, N and S Front Sts, W Vine, and SW Third Sts, Walnut Ridge...-103 W Cherry St, Salem, 10000291 Cook-Johnson House, 3091 Franklin Ave, Salem, 10000292 Farmers...

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.

Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-1: Pressurized Water Reactors.

ERIC Educational Resources Information Center

Reihman, Thomas C.

This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical pressurized water reactor (PWR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module is the PWR…

33 CFR 117.737 - Oldmans Creek.

Code of Federal Regulations, 2012 CFR

2012-07-01

... US30 bridge, mile 3.1 at Nortonville, the Conrail railroad bridge, mile 4.0 at Jumbo, and the Salem County bridge, mile 5.1 at Pedricktown, need not be opened for the passage of vessels. However, the draws...

33 CFR 117.737 - Oldmans Creek.

Code of Federal Regulations, 2013 CFR

2013-07-01

... US30 bridge, mile 3.1 at Nortonville, the Conrail railroad bridge, mile 4.0 at Jumbo, and the Salem County bridge, mile 5.1 at Pedricktown, need not be opened for the passage of vessels. However, the draws...

33 CFR 117.737 - Oldmans Creek.

Code of Federal Regulations, 2014 CFR

2014-07-01

... US30 bridge, mile 3.1 at Nortonville, the Conrail railroad bridge, mile 4.0 at Jumbo, and the Salem County bridge, mile 5.1 at Pedricktown, need not be opened for the passage of vessels. However, the draws...

33 CFR 117.737 - Oldmans Creek.

Code of Federal Regulations, 2011 CFR

2011-07-01

... US30 bridge, mile 3.1 at Nortonville, the Conrail railroad bridge, mile 4.0 at Jumbo, and the Salem County bridge, mile 5.1 at Pedricktown, need not be opened for the passage of vessels. However, the draws...

33 CFR 117.737 - Oldmans Creek.

Code of Federal Regulations, 2010 CFR

2010-07-01

... US30 bridge, mile 3.1 at Nortonville, the Conrail railroad bridge, mile 4.0 at Jumbo, and the Salem County bridge, mile 5.1 at Pedricktown, need not be opened for the passage of vessels. However, the draws...

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

75 FR 21220 - Hood/Willamette Resource Advisory Committee (RAC)

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-23

... Management Office; 1717 Fabry Road SE; Salem, Oregon; (503) 375-5646. The tentative agenda includes: (1... encouraged, particularly if the material cannot be presented within the time limits for the Public Forum...

REUSABLE ADSORBENTS FOR DILUTE SOLUTIONS SEPARATION. 6. BATCH AND CONTINUOUS REACTORS FOR ADSORPTION AND DEGRADATION OF 1,2-DICHLOROBENZENE FROM DILUTE WASTEWATER STREAMS USING TITANIA AS A PHOTOCATALYST. (R828598C753)

EPA Science Inventory

Two types of external lamp reactors were investigated for the titania catalyzed photodegradation of 1,2-dichlorobenzene (DCB) from a dilute water stream. The first one was a batch mixed slurry reactor and the second one was a semi-batch reactor with continuous feed recycle wit...

Controlled nitric oxide production via O(1D) + N2O reactions for use in oxidation flow reactor studies

NASA Astrophysics Data System (ADS)

Lambe, Andrew; Massoli, Paola; Zhang, Xuan; Canagaratna, Manjula; Nowak, John; Daube, Conner; Yan, Chao; Nie, Wei; Onasch, Timothy; Jayne, John; Kolb, Charles; Davidovits, Paul; Worsnop, Douglas; Brune, William

2017-06-01

Oxidation flow reactors that use low-pressure mercury lamps to produce hydroxyl (OH) radicals are an emerging technique for studying the oxidative aging of organic aerosols. Here, ozone (O3) is photolyzed at 254 nm to produce O(1D) radicals, which react with water vapor to produce OH. However, the need to use parts-per-million levels of O3 hinders the ability of oxidation flow reactors to simulate NOx-dependent secondary organic aerosol (SOA) formation pathways. Simple addition of nitric oxide (NO) results in fast conversion of NOx (NO + NO2) to nitric acid (HNO3), making it impossible to sustain NOx at levels that are sufficient to compete with hydroperoxy (HO2) radicals as a sink for organic peroxy (RO2) radicals. We developed a new method that is well suited to the characterization of NOx-dependent SOA formation pathways in oxidation flow reactors. NO and NO2 are produced via the reaction O(1D) + N2O → 2NO, followed by the reaction NO + O3 → NO2 + O2. Laboratory measurements coupled with photochemical model simulations suggest that O(1D) + N2O reactions can be used to systematically vary the relative branching ratio of RO2 + NO reactions relative to RO2 + HO2 and/or RO2 + RO2 reactions over a range of conditions relevant to atmospheric SOA formation. We demonstrate proof of concept using high-resolution time-of-flight chemical ionization mass spectrometer (HR-ToF-CIMS) measurements with nitrate (NO3-) reagent ion to detect gas-phase oxidation products of isoprene and α-pinene previously observed in NOx-influenced environments and in laboratory chamber experiments.

Controlled Nitric Oxide Production via O(1D) + N2O Reactions for Use in Oxidation Flow Reactor Studies

NASA Technical Reports Server (NTRS)

Lambe, Andrew; Massoli, Paola; Zhang, Xuan; Canagaratna, Manjula; Nowak, John; Daube, Conner; Yan, Chao; Nie, Wei; Onasch, Timothy; Jayne, John;

Carbofuran removal in continuous-photocatalytic reactor: Reactor optimization, rate-constant determination and carbofuran degradation pathway analysis.

PubMed

Vishnuganth, M A; Remya, Neelancherry; Kumar, Mathava; Selvaraju, N

2017-05-04

Carbofuran (CBF) removal in a continuous-flow photocatalytic reactor with granular activated carbon supported titanium dioxide (GAC-TiO 2 ) catalyst was investigated. The effects of feed flow rate, TiO 2 concentration and addition of supplementary oxidants on CBF removal were investigated. The central composite design (CCD) was used to design the experiments and to estimate the effects of feed flow rate and TiO 2 concentration on CBF removal. The outcome of CCD experiments demonstrated that reactor performance was influenced mainly by feed flow rate compared to TiO 2 concentration. A second-order polynomial model developed based on CCD experiments fitted the experimental data with good correlation (R 2 ∼ 0.964). The addition of 1 mL min -1 hydrogen peroxide has shown complete CBF degradation and 76% chemical oxygen demand removal under the following operating conditions of CBF ∼50 mg L -1 , TiO 2 ∼5 mg L -1 and feed flow rate ∼82.5 mL min -1 . Rate constant of the photodegradation process was also calculated by applying the kinetic data in pseudo-first-order kinetics. Four major degradation intermediates of CBF were identified using GC-MS analysis. As a whole, the reactor system and GAC-TiO 2 catalyst used could be constructive in cost-effective CBF removal with no impact to receiving environment through getaway of photocatalyst.

THERMAL NEUTRONIC REACTOR

DOEpatents

Spinrad, B.I.

1960-01-12

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

Biofilm reactors for industrial bioconversion processes: employing potential of enhanced reaction rates

PubMed Central

Qureshi, Nasib; Annous, Bassam A; Ezeji, Thaddeus C; Karcher, Patrick; Maddox, Ian S

2005-01-01

This article describes the use of biofilm reactors for the production of various chemicals by fermentation and wastewater treatment. Biofilm formation is a natural process where microbial cells attach to the support (adsorbent) or form flocs/aggregates (also called granules) without use of chemicals and form thick layers of cells known as "biofilms." As a result of biofilm formation, cell densities in the reactor increase and cell concentrations as high as 74 gL-1 can be achieved. The reactor configurations can be as simple as a batch reactor, continuous stirred tank reactor (CSTR), packed bed reactor (PBR), fluidized bed reactor (FBR), airlift reactor (ALR), upflow anaerobic sludge blanket (UASB) reactor, or any other suitable configuration. In UASB granular biofilm particles are used. This article demonstrates that reactor productivities in these reactors have been superior to any other reactor types. This article describes production of ethanol, butanol, lactic acid, acetic acid/vinegar, succinic acid, and fumaric acid in addition to wastewater treatment in the biofilm reactors. As the title suggests, biofilm reactors have high potential to be employed in biotechnology/bioconversion industry for viable economic reasons. In this article, various reactor types have been compared for the above bioconversion processes. PMID:16122390

Bioconversion reactor

DOEpatents

McCarty, Perry L.; Bachmann, Andre

1992-01-01

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

Modelling of the anti-neutrino production and spectra from a Magnox reactor

NASA Astrophysics Data System (ADS)

Mills, Robert W.; Mountford, David J.; Coleman, Jonathon P.; Metelko, Carl; Murdoch, Matthew; Schnellbach, Yan-Jie

2018-01-01

The anti-neutrino source properties of a fission reactor are governed by the production and beta decay of the radionuclides present and the summation of their individual anti-neutrino spectra. The fission product radionuclide production changes during reactor operation and different fissioning species give rise to different product distributions. It is thus possible to determine some details of reactor operation, such as power, from the anti-neutrino emission to confirm safeguards records. Also according to some published calculations, it may be feasible to observe different anti-neutrino spectra depending on the fissile contents of the reactor fuel and thus determine the reactor's fissile material inventory during operation which could considerable improve safeguards. In mid-2014 the University of Liverpool deployed a prototype anti-neutrino detector at the Wylfa R1 station in Anglesey, United Kingdom based upon plastic scintillator technology developed for the T2K project. The deployment was used to develop the detector electronics and software until the reactor was finally shutdown in December 2015. To support the development of this detector technology for reactor monitoring and to understand its capabilities, the National Nuclear Laboratory modelled this graphite moderated and natural uranium fuelled reactor with existing codes used to support Magnox reactor operations and waste management. The 3D multi-physics code PANTHER was used to determine the individual powers of each fuel element (8×6152) during the year and a half period of monitoring based upon reactor records. The WIMS/TRAIL/FISPIN code route was then used to determine the radionuclide inventory of each nuclide on a daily basis in each element. These nuclide inventories were then used with the BTSPEC code to determine the anti-neutrino spectra and source strength using JEFF-3.1.1 data. Finally the anti-neutrino source from the reactor for each day during the year and a half of monitored reactor

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

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

Bahri, Che Nor Aniza Che Zainul, E-mail: anizazainul@gmail.com; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

2015-04-29

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 nuclearmore » 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.« less

Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

NASA Astrophysics Data System (ADS)

Cisneros, Anselmo Tomas, Jr.

The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP

Hybrid plasmachemical reactor

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

Lelevkin, V. M., E-mail: lelevkin44@mail.ru; Smirnova, Yu. G.; Tokarev, A. V.

2015-04-15

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

SM-1 REACTOR VESSEL COVER AND FLANGE STRESS ANALYSIS

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

Sayre, M.F.

1962-02-19

The maximum stress calculated for the SMl-1 reactor vessel closure studs occurs during operation at full power. This value is 27,180 psi of which 19,800 psi is tension and 7380 psi bending. This stress does not include a stress concentration factor for effect of threads. It was eonservatively assumed the studs were initially tightened to a code allowable stress of 20,000 psi as specified in the ASME Code rather than the lesser stress obtained by the normal operating procedure. The maximum calculated stress occurs at the outside surface of the cover where the stress ranges from 318 psi in tensionmore » to 90,660 psi in compression. The alternating stress is 50,000 psi. According to the Navy Code for a stress range of 50,000 psi, the eover material ean safely undergo a maximum of 1600 cycles. It was estimated that the SM-1 will go through approximately 000 startup and shutdown cycles during a 20-yr life period, so the calculated stress is regarded as safe. For a transient eondition of 30 deg F/hr during heat-up, approximate temperature differences between the inside and outside surfaces of the cover were obtained. Temperature differentials between the inside and outside surfaces of the cover are increased by roughly 10%; above the steady state condition. More exact calculations of the transient stresses did not appear necessary siuce they would be not more than 10% greater than the steady state thermal stress. (auth)« less

A dual purpose packed-bed reactor for biogas scrubbing and methane-dependent water quality improvement applying to a wastewater treatment system consisting of UASB reactor and trickling filter.

PubMed

Tanaka, Yasuo

2002-08-01

A wastewater treatment system employing a UASB reactor in temperate regions requires biogas as a heat source for the UASB reactor during low temperature seasons. In this case, removal of H2S in the biogas by means of a scrubber before burning is necessary in order to prevent the boilers from corroding. Heating of the UASB reactor is, however, unnecessary in a warm season, and the scrubber and biogas become useless. Methane-dependent water quality improvement using the scrubber and biogas would be one way to use them efficiently during the warm season. The possible dual-purpose use of a packed-bed reactor was examined, with one of its uses being the scrubbing of biogas during the cold season and the other being the methane-dependent improvement of effluent water quality during the warm season. A bench scale packed-bed filled with plastic latticed-ring media was installed in a livestock wastewater treatment plant consisting of a UASB reactor and a trickling filter for post-treatment. The packed-bed was operated with biogas flowing at a superficial velocity of 0.14-0.39 m h(-1) and the hydraulic loading of trickling filter effluent sprayed onto the media 9.4-26.1 m3 m2 day(-1). H2S in the biogas from the UASB reactor was reduced from 1,200-2,500 ppm to less than 2 ppm by the reactor. Methane-dependent water quality improvement was examined using a laboratory scale reactor to which methane and/or air was supplied from the bottom, while plant effluent was spread from the top of the reactor. When the mixture gas of methane and air (volume ratio 1:3) was added to the reactor, biofilm grew on the surface of the media. Accompanying this growth, ammonium and phosphate in the spread water decreased, probably due to assimilation by the methane-oxidizing bacteria. Though assimilation activity dropped after the accumulation of biomass, it could be reactivated by washing out the excess biomass. Periodical backwash at a rate of more than once a week seemed to efficiently maintain

HOMOGENEOUS NUCLEAR POWER REACTOR

DOEpatents

King, L.D.P.

1959-09-01

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

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 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Immediate notification requirements for operating nuclear power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear...

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Immediate notification requirements for operating nuclear power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear...

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.

Solvent refined coal reactor quench system

DOEpatents

Thorogood, Robert M.

1983-01-01

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

148. ARAIII Reactor building (ARA608) Floor plan. Shows location of ...

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

148. ARA-III Reactor building (ARA-608) Floor plan. Shows location of reactor, heater, and mechanical loop pits; mechanical and electrical equipment rooms; and other work areas. Aerojet-general 880-area/GCRE-608-A-1. Date: February 1958. Ineel index code no. 063-0608-00-013-102612. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

REACTOR SERVICES BUILDING, TRA635, INTERIOR. ALSO KNOWN AS MATERIAL RECEIVING ...

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

REACTOR SERVICES BUILDING, TRA-635, INTERIOR. ALSO KNOWN AS MATERIAL RECEIVING AREA AND LABORATORY. CAMERA ON FIRST FLOOR FACING NORTH TOWARD MTR BUILDING. MOCK-UP AREA WAS TO THE RIGHT OF VIEW. INL NEGATIVE NO. HD46-10-1. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere with...

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

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere with...

Spherical torus fusion reactor

DOEpatents

Peng, Yueng-Kay M.

1989-04-04

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

Spherical torus fusion reactor

DOEpatents

Peng, Yueng-Kay M.

1989-01-01

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

Fluorescence-based monitoring of tracer and substrate distribution in an UASB reactor.

PubMed

Lou, S J; Tartakovsky, B; Zeng, Y; Wu, P; Guiot, S R

2006-11-01

In this work, rhodamine-related fluorescence was measured on-line at four reactor heights in order to study hydrodynamics within an upflow anaerobic sludge bed reactor. A linear dependence of the dispersion coefficient (D) on the upflow velocity was observed, while the influence of the organic loading rate (OLR) was insignificant. Furthermore, the Bodenstein number of the reactor loaded with granulated sludge was found to be position-dependent with the largest values measured at the bottom of the sludge bed. This trend was not observed in the reactor without sludge. Chemical oxygen demand (COD) and volatile fatty acid (VFA) concentrations were measured at the same reactor heights as in rhodamine tests using conventional off-line analytical methods and on-line multiwavelength fluorometry. Significant spatial COD and VFA gradients were observed at organic loading rates above 6g COD l(R)(-1)d(-1) and linear upflow velocities below 0.8m h(-1).

Multicapillary Flow Reactor: Synthesis of 1,2,5-Thiadiazepane 1,1-Dioxide Library Utilizing One-Pot Elimination and Inter-/Intramolecular Double aza-Michael Addition Via Microwave-Assisted, Continuous-Flow Organic Synthesis (MACOS)

PubMed Central

Ullah, Farman; Zang, Qin; Javed, Salim; Zhou, Aihua; Knudtson, Christopher A.; Bi, Danse; Hanson, Paul R.; Organ, Michael G.

2013-01-01

A microwave-assisted, continuous-flow organic synthesis (MACOS) protocol for the synthesis of functionalized 1,2,5-thiadiazepane 1,1-dioxide library, utilizing a one-pot elimination and inter-/intramolecular double aza-Michael addition strategy is reported. The optimized protocol in MACOS was utilized for scale-out and further extended for library production using a multicapillary flow reactor. A 50-member library of 1,2,5-thiadiazepane 1,1-dioxides was prepared on a 100- to 300-mg scale with overall yields between 50 and 80% and over 90 % purity determined by proton nuclear magnetic resonance (1H-NMR) spectroscopy. PMID:24244871

The effect of cover use on plastic pyrolysis reactor heating process

NASA Astrophysics Data System (ADS)

Armadi, Benny H.; Rangkuti, Chalilullah; Fauzi, M. D.; Permatasari, R.

2017-03-01

Plastic pyrolysis process to produce liquid fuel is an endothermic process that uses heat from the combustion of fuel as heat source. The reactor used is usually a vertical cylindrical in shape, with LPG fuel combustion under the flat bottom of the reactor, and the combustion gases is dispersed into the surrounding environment, so that heat transferred to the plastic inside the reactor is not effective, causing high LPG consumption. In this study, the reactor is made of stainless steel plate, with a vertical cylindrical shape, with a basic cylindrical conical truncated by a pit pass hot flue gas in the middle that serves to deliver flue gas into the chimney. The contact area between the hot combusted LPG gases to the processed plastic inside the reactor becomes bigger and gets better heat transfer, and required less LPG consumption. For more effective heat transfer process, an outer cover of this reactor was made and the relatively hot combustion gases are used to heat the outside of the reactor by directing the flow of the flue gas from the chimney down along the outer wall of the reactor and out the bottom lid. This construction makes the heating process to be faster and the LPG fuel is used more efficiently. From the measurements, it was found to raise 1°C of temperature inside the covered reactor, the LPG consumed is 0.59 gram, and if the reactor cover is removed, the gas demand will rise nearly threefold to 1.43 grams. With this method, in addition to reducing the rate of heat loss will also help reduce LPG consumption significantly.

FOOD IRRADIATION REACTOR

DOEpatents

Leyse, C.F.; Putnam, G.E.

1961-05-01

An irradiation apparatus is described. It comprises a pressure vessel, a neutronic reactor active portion having a substantially greater height than diameter in the pressure vessel, an annular tank surrounding and spaced from the pressure vessel containing an aqueous indium/sup 1//sup 1//sup 5/ sulfate solution of approximately 600 grams per liter concentration, means for circulating separate coolants through the active portion and the space between the annular tank and the pressure vessel, radiator means adapted to receive the materials to be irradiated, and means for flowing the indium/sup 1//sup 1//sup 5/ sulfate solution through the radiator means.

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

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

Carbajo, Juan J; Qualls, A L

2016-01-01

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

REACTOR FUEL SCAVENGING MEANS

DOEpatents

Coffinberry, A.S.

1962-04-10

A process for removing fission products from reactor liquid fuel without interfering with the reactor's normal operation or causing a significant change in its fuel composition is described. The process consists of mixing a liquid scavenger alloy composed of about 44 at.% plutoniunm, 33 at.% lanthanum, and 23 at.% nickel or cobalt with a plutonium alloy reactor fuel containing about 3 at.% lanthanum; removing a portion of the fuel and scavenger alloy from the reactor core and replacing it with an equal amount of the fresh scavenger alloy; transferring the portion to a quiescent zone where the scavenger and the plutonium fuel form two distinct liquid layers with the fission products being dissolved in the lanthanum-rich scavenger layer; and the clean plutonium-rich fuel layer being returned to the reactor core. (AEC)

Reactor-Produced Medical Radionuclides

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

Mirzadeh, Saed; Mausner, Leonard; Garland, Marc A

2011-01-01

The therapeutic use of radionuclides in nuclear medicine, oncology and cardiology is the most rapidly growing use of medical radionuclides. Since most therapeutic radionuclides are neutron rich and decay by beta emission, they are reactor-produced. This chapter deals mainly with production approaches with neutrons. Neutron interactions with matter, neutron transmission and activation rates, and neutron spectra of nuclear reactors are discussed in some detail. Further, a short discussion of the neutron-energy dependence of cross sections, reaction rates in thermal reactors, cross section measurements and flux monitoring, and general equations governing the reactor production of radionuclides are presented. Finally, the chaptermore » is concluded by providing a number of examples encompassing the various possible reaction routes for production of a number of medical radionuclides in a reactor.« less

Preliminary validation of computational model for neutron flux prediction of Thai Research Reactor (TRR-1/M1)

NASA Astrophysics Data System (ADS)

Sabaibang, S.; Lekchaum, S.; Tipayakul, C.

2015-05-01

This study is a part of an on-going work to develop a computational model of Thai Research Reactor (TRR-1/M1) which is capable of accurately predicting the neutron flux level and spectrum. The computational model was created by MCNPX program and the CT (Central Thimble) in-core irradiation facility was selected as the location for validation. The comparison was performed with the typical flux measurement method routinely practiced at TRR-1/M1, that is, the foil activation technique. In this technique, gold foil is irradiated for a certain period of time and the activity of the irradiated target is measured to derive the thermal neutron flux. Additionally, the flux measurement with SPND (self-powered neutron detector) was also performed for comparison. The thermal neutron flux from the MCNPX simulation was found to be 1.79×1013 neutron/cm2s while that from the foil activation measurement was 4.68×1013 neutron/cm2s. On the other hand, the thermal neutron flux from the measurement using SPND was 2.47×1013 neutron/cm2s. An assessment of the differences among the three methods was done. The difference of the MCNPX with the foil activation technique was found to be 67.8% and the difference of the MCNPX with the SPND was found to be 27.8%.

Fiber-Optical Sensors: Basics and Applications in Multiphase Reactors

PubMed Central

Li, Xiangyang; Yang, Chao; Yang, Shifang; Li, Guozheng

2012-01-01

This work presents a brief introduction on the basics of fiber-optical sensors and an overview focused on the applications to measurements in multiphase reactors. The most commonly principle utilized is laser back scattering, which is also the foundation for almost all current probes used in multiphase reactors. The fiber-optical probe techniques in two-phase reactors are more developed than those in three-phase reactors. There are many studies on the measurement of gas holdup using fiber-optical probes in three-phase fluidized beds, but negative interference of particles on probe function was less studied. The interactions between solids and probe tips were less studied because glass beads etc. were always used as the solid phase. The vision probes may be the most promising for simultaneous measurements of gas dispersion and solids suspension in three-phase reactors. Thus, the following techniques of the fiber-optical probes in multiphase reactors should be developed further: (1) online measuring techniques under nearly industrial operating conditions; (2) corresponding signal data processing techniques; (3) joint application with other measuring techniques.

NEUTRONIC REACTOR

DOEpatents

Metcalf, H.E.

1957-10-01

A reactor of the type which preferably uses plutonium as the fuel and a liquid moderator, preferably ordinary water, and which produces steam within the reactor core due to the heat of the chain reaction is described. In the reactor shown the fuel elements are essentially in the form of trays and are ventically stacked in spaced relationship. The water moderator is continuously supplied to the trays to maintain a constant level on the upper surfaces of the fuel element as it is continually evaporated by the heat. The steam passes out through the spaces between the fuel elements and is drawn off at the top of the core. The fuel elements are clad in aluminum to prevent deterioration thereof with consequent contamimation of the water.

REACTOR CONTROL

DOEpatents

Fortescue, P.; Nicoll, D.

1962-04-24

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

NUCLEAR REACTOR CONTROL SYSTEM

DOEpatents

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

1959-11-01

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

Reactor vessel support system. [LMFBR

DOEpatents

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

1980-05-09

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

Identification of major sources controlling groundwater chemistry from a hard rock terrain — A case study from Mettur taluk, Salem district, Tamil Nadu, India

NASA Astrophysics Data System (ADS)

Srinivasamoorthy, K.; Chidambaram, S.; Prasanna, M. V.; Vasanthavihar, M.; Peter, John; Anandhan, P.

2008-02-01

The study area Mettur forms an important industrial town situated NW of Salem district. The geology of the area is mainly composed of Archean crystalline metamorphic complexes. To identify the major process activated for controlling the groundwater chemistry an attempt has been made by collecting a total of 46 groundwater samples for two different seasons, viz., pre-monsoon and post-monsoon. The groundwater chemistry is dominated by silicate weathering and (Na + Mg) and (Cl + SO4) accounts of about 90% of cations and anions. The contribution of (Ca + Mg) and (Na + K) to total cations and HCO3 indicates the domination of silicate weathering as major sources for cations. The plot for Na to Cl indicates higher Cl in both seasons, derived from Anthropogenic (human) sources from fertilizer, road salt, human and animal waste, and industrial applications, minor representations of Na also indicates source from weathering of silicate-bearing minerals. The plot for Na/Cl to EC indicates Na released from silicate weathering process which is also supported by higher HCO3 values in both the seasons. Ion exchange process is also activated in the study area which is indicated by shifting to right in plot for Ca + Mg to SO4 + HCO3. The plot of Na-Cl to Ca + Mg-HCO3-SO4 confirms that Ca, Mg and Na concentrations in groundwater are derived from aquifer materials. Thermodynamic plot indicates that groundwater is in equilibrium with kaolinite, muscovite and chlorite minerals. Saturation index of silicate and carbonate minerals indicate oversaturation during pre-monsoon and undersaturation during post-monsoon, conforming dissolution and dilution process. In general, water chemistry is guided by complex weathering process, ion exchange along with influence of Cl ions from anthropogenic impact.

Interim waste storage for the Integral Fast Reactor

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

Benedict, R.W.; Phipps, R.D.; Condiff, D.W.

1991-01-01

The Integral Fast Reactor (IFR), which Argonne National Laboratory is developing, is an innovative liquid metal breeder reactor that uses metallic fuel and has a close coupled fuel recovery process. A pyrochemical process is used to separate the fission products from the actinide elements. These actinides are used to make new fuel for the reactor. As part of the overall IFR development program, Argonne has refurbished an existing Fuel Cycle Facility at ANL-West and is installing new equipment to demonstrate the remote reprocessing and fabrication of fuel for the Experimental Breeder Reactor II (EBR-II). During this demonstration the wastes thatmore » are produced will be treated and packaged to produce waste forms that would be typical of future commercial operations. These future waste forms would, assuming Argonne development goals are fulfilled, be essentially free of long half-life transuranic isotopes. Promising early results indicate that actinide extraction processes can be developed to strip these isotopes from waste stream and return them to the IFR type reactors for fissioning. 1 fig.« less

Attrition reactor system

DOEpatents

Scott, Charles D.; Davison, Brian H.

1993-01-01

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

Non-equilibrium plasma reactors for organic solvent destruction

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

Yang, C.L.; Beltran, M.R.; Kravets, Z.

1997-12-31

Two non-equilibrium plasma reactors were evaluated for their ability to destroy three widely used organic solvents, i.e., 2-butanone, toluene and ethyl acetate. The catalyzed plasma reactor (CPR) with 6 mm glass beads destroys 98% of 50 ppm toluene in air at 24 kV/cm and space velocities of 1,400 v/v/hr. Eighty-five percent of ethyl acetate and 2-butanone are destroyed under the same conditions. The tubular plasma reactor (TPR) has an efficiency of 10% to 20% lower than that of a CPR under the same conditions. The 1,400 v/v/hr in a CPR is equal to a residence time of 2.6 seconds inmore » a TPR. The operating temperatures, corona characteristics, as well as the kinetics of VOC destruction in both TPR and CPR were studied.« less

Tower Shielding Reactor II design and operation report: Vol. 2. Safety Analysis

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

Holland, L. B.; Kolb, J. O.

1970-01-01

Information on the Tower Shielding Reactor II is contained in the TSR-II Design and Operation Report and in the Tower Shielding Facility Manual. The TSR-II Design and Operating Report consists of three volumes. Volume 1 is Descriptions of the Tower Shielding Reactor II and Facility; Volume 2 is Safety analysis of the Tower Shielding Reactor II; and Volume 3 is the Assembly and Testing of the Tower Shielding Reactor II Control Mechanism Housing.

77 FR 8902 - Draft Regulatory Guide: Issuance, Availability Decommissioning of Nuclear Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-15

... Decommissioning of Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide... draft regulatory guide (DG) DG-1271 ``Decommissioning of Nuclear Power Reactors.'' This guide describes... Regulatory Guide 1.184, ``Decommissioning of Nuclear Power Reactors,'' dated July 2000. This proposed...

Growth mechanism of GaAs1-xSbx ternary alloy thin film on MOCVD reactor using TMGa, TDMAAs and TDMASb

NASA Astrophysics Data System (ADS)

Suhandi, A.; Tayubi, Y. R.; Arifin, P.

2016-04-01

Metal Organic Chemical Vapor Deposition (MOCVD) is a method for growing a solid material (in the form of thin films, especially for semiconductor materials) using vapor phase metal organic sources. Studies on the growth mechanism of GaAs1-xSbx ternary alloy thin solid film in the range of miscibility-gap using metal organic sources trimethylgallium (TMGa), trisdimethylaminoarsenic (TDMAAs), and trisdimethylaminoantimony (TDMASb) on MOCVD reactor has been done to understand the physical and chemical processes involved. Knowledge of the processes that occur during alloy formation is very important to determine the couple of growth condition and growth parameters are appropriate for yield high quality GaAs1-xSbx alloy. The mechanism has been studied include decomposition of metal organic sources and chemical reactions that may occur, the incorporation of the alloy elements forming and the contaminants element that are formed in the gown thin film. In this paper presented the results of experimental data on the growth of GaAs1-xSbx alloy using Vertical-MOCVD reactor to demonstrate its potential in growing GaAs1-xSbx alloy in the range of its miscibility gap.

Microfluidic electrochemical reactors

DOEpatents

Nuzzo, Ralph G [Champaign, IL; Mitrovski, Svetlana M [Urbana, IL

2011-03-22

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

REACTOR CONTROL SYSTEM

DOEpatents

MacNeill, J.H.; Estabrook, J.Y.

1960-05-10

A reactor control system including a continuous tape passing through a first coolant passageway, over idler rollers, back through another parallel passageway, and over motor-driven rollers is described. Discrete portions of fuel or poison are carried on two opposed active sections of the tape. Driving the tape in forward or reverse directions causes both active sections to be simultaneously inserted or withdrawn uniformly, tending to maintain a more uniform flux within the reactor. The system is particularly useful in mobile reactors, where reduced inertial resistance to control rod movement is important.

NEUTRONIC REACTOR CONTROL

DOEpatents

Metcalf, H.E.

1958-10-14

Methods of controlling reactors are presented. Specifically, a plurality of neutron absorber members are adjustably disposed in the reactor core at different distances from the center thereof. The absorber members extend into the core from opposite faces thereof and are operated by motive means coupled in a manner to simultaneously withdraw at least one of the absorber members while inserting one of the other absorber members. This feature effects fine control of the neutron reproduction ratio by varying the total volume of the reactor effective in developing the neutronic reaction.

Postassembly Transformation of a Catalytically Active Composite Material, Pt@ZIF-8, via Solvent-Assisted Linker Exchange.

PubMed

Stephenson, Casey J; Hupp, Joseph T; Farha, Omar K

2016-02-15

2-Methylimidazolate linkers of Pt@ZIF-8 are exchanged with imidazolate using solvent-assisted linker exchange (SALE) to expand the apertures of the parent material and create Pt@SALEM-2. Characterization of the material before and after SALE was performed. Both materials are active as catalysts for the hydrogenation of 1-octene, whereas the hydrogenation of cis-cyclohexene occurred only with Pt@SALEM-2, consistent with larger apertures for the daughter material. The largest substrate, β-pinene, proved to be unreactive with H2 when either material was employed as a candidate catalyst, supporting the contention that substrate molecules, for both composites, must traverse the metal-organic framework component in order to reach the catalytic nanoparticles.

Design of megawatt power level heat pipe reactors

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

Mcclure, Patrick Ray; Poston, David Irvin; Dasari, Venkateswara Rao

An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors.more » The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.« less

Enrichment of acetogenic bacteria in high rate anaerobic reactors under mesophilic and thermophilic conditions.

PubMed

Ryan, P; Forbes, C; McHugh, S; O'Reilly, C; Fleming, G T A; Colleran, E

2010-07-01

The objective of the current study was to expand the knowledge of the role of acetogenic Bacteria in high rate anaerobic digesters. To this end, acetogens were enriched by supplying a variety of acetogenic growth supportive substrates to two laboratory scale high rate upflow anaerobic sludge bed (UASB) reactors operated at 37 degrees C (R1) and 55 degrees C (R2). The reactors were initially fed a glucose/acetate influent. Having achieved high operational performance and granular sludge development and activity, both reactors were changed to homoacetogenic bacterial substrates on day 373 of the trial. The reactors were initially fed with sodium vanillate as a sole substrate. Although % COD removal indicated that the 55 degrees C reactor out performed the 37 degrees C reactor, effluent acetate levels from R2 were generally higher than from R1, reaching values as high as 5023 mg l(-1). Homoacetogenic activity in both reactors was confirmed on day 419 by specific acetogenic activity (SAA) measurement, with higher values obtained for R2 than R1. Sodium formate was introduced as sole substrate to both reactors on day 464. It was found that formate supported acetogenic activity at both temperatures. By the end of the trial, no specific methanogenic activity (SMA) was observed against acetate and propionate indicating that the methane produced was solely by hydrogenotrophic Archaea. Higher SMA and SAA values against H(2)/CO(2) suggested development of a formate utilising acetogenic population growing in syntrophy with hydrogenotrophic methanogens. Throughout the formate trial, the mesophilic reactor performed better overall than the thermophilic reactor. Copyright 2010 Elsevier Ltd. All rights reserved.

Controlled nitric oxide production via O( 1D)+N 2O reactions for use in oxidation flow reactor studies

DOE PAGES

Lambe, Andrew; Massoli, Paola; Zhang, Xuan; ...

2017-06-22

Oxidation flow reactors that use low-pressure mercury lamps to produce hydroxyl (OH) radicals are an emerging technique for studying the oxidative aging of organic aerosols. Here, ozone (O 3) is photolyzed at 254 nm to produce O( 1D) radicals, which react with water vapor to produce OH. However, the need to use parts-per-million levels of O 3 hinders the ability of oxidation flow reactors to simulate NO x-dependent secondary organic aerosol (SOA) formation pathways. Simple addition of nitric oxide (NO) results in fast conversion of NO x (NO+NO 2) to nitric acid (HNO 3), making it impossible to sustain NOmore » x at levels that are sufficient to compete with hydroperoxy (HO 2) radicals as a sink for organic peroxy (RO 2) radicals. We developed a new method that is well suited to the characterization of NO x-dependent SOA formation pathways in oxidation flow reactors. NO and NO 2 are produced via the reaction O( 1D) + N 2O → 2NO, followed by the reaction NO + O 3 → NO 2+O 2. Laboratory measurements coupled with photochemical model simulations suggest that O( 1D) + N 2O reactions can be used to systematically vary the relative branching ratio of RO 2 + NO reactions relative to RO 2 + HO 2 and/or RO 2 + RO 2 reactions over a range of conditions relevant to atmospheric SOA formation. We demonstrate proof of concept using high-resolution time-of-flight chemical ionization mass spectrometer (HR-ToF-CIMS) measurements with nitrate (NO 3 -) reagent ion to detect gas-phase oxidation products of isoprene and α-pinene previously observed in NO x-influenced environments and in laboratory chamber experiments.« less

Controlled nitric oxide production via O( 1D)+N 2O reactions for use in oxidation flow reactor studies

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

Lambe, Andrew; Massoli, Paola; Zhang, Xuan

Oxidation flow reactors that use low-pressure mercury lamps to produce hydroxyl (OH) radicals are an emerging technique for studying the oxidative aging of organic aerosols. Here, ozone (O 3) is photolyzed at 254 nm to produce O( 1D) radicals, which react with water vapor to produce OH. However, the need to use parts-per-million levels of O 3 hinders the ability of oxidation flow reactors to simulate NO x-dependent secondary organic aerosol (SOA) formation pathways. Simple addition of nitric oxide (NO) results in fast conversion of NO x (NO+NO 2) to nitric acid (HNO 3), making it impossible to sustain NOmore » x at levels that are sufficient to compete with hydroperoxy (HO 2) radicals as a sink for organic peroxy (RO 2) radicals. We developed a new method that is well suited to the characterization of NO x-dependent SOA formation pathways in oxidation flow reactors. NO and NO 2 are produced via the reaction O( 1D) + N 2O → 2NO, followed by the reaction NO + O 3 → NO 2+O 2. Laboratory measurements coupled with photochemical model simulations suggest that O( 1D) + N 2O reactions can be used to systematically vary the relative branching ratio of RO 2 + NO reactions relative to RO 2 + HO 2 and/or RO 2 + RO 2 reactions over a range of conditions relevant to atmospheric SOA formation. We demonstrate proof of concept using high-resolution time-of-flight chemical ionization mass spectrometer (HR-ToF-CIMS) measurements with nitrate (NO 3 -) reagent ion to detect gas-phase oxidation products of isoprene and α-pinene previously observed in NO x-influenced environments and in laboratory chamber experiments.« less

Status of French reactors

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

Ballagny, A.

1997-08-01

The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (exceptmore » if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.« less

Startup of reactors for anoxic ammonium oxidation: experiences from the first full-scale anammox reactor in Rotterdam.

PubMed

van der Star, Wouter R L; Abma, Wiebe R; Blommers, Dennis; Mulder, Jan-Willem; Tokutomi, Takaaki; Strous, Marc; Picioreanu, Cristian; van Loosdrecht, Mark C M

2007-10-01

The first full-scale anammox reactor in the world was started in Rotterdam (NL). The reactor was scaled-up directly from laboratory-scale to full-scale and treats up to 750 kg-N/d. In the initial phase of the startup, anammox conversions could not be identified by traditional methods, but quantitative PCR proved to be a reliable indicator for growth of the anammox population, indicating an anammox doubling time of 10-12 days. The experience gained during this first startup in combination with the availability of seed sludge from this reactor, will lead to a faster startup of anammox reactors in the future. The anammox reactor type employed in Rotterdam was compared to other reactor types for the anammox process. Reactors with a high specific surface area like the granular sludge reactor employed in Rotterdam provide the highest volumetric loading rates. Mass transfer of nitrite into the biofilm is limiting the conversion of those reactor types that have a lower specific surface area. Now the first full-scale commercial anammox reactor is in operation, a consistent and descriptive nomenclature is suggested for reactors in which the anammox process is employed.

NUCLEAR REACTOR

DOEpatents

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

1961-06-20

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

Attrition reactor system

DOEpatents

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

1993-09-28

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

REACTOR PHYSICS CONSTANTS

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

None

1963-07-01

This second edition is based on data available on March 15, 1961. Sections on constants necessary for the interpretation of experimental data and on digital computer programs for reactor design and reactor physics have been added. 1344 references. (D.C.W.)

MATERIALS TESTING REACTOR (MTR) BUILDING, TRA603. CONTEXTUAL VIEW OF MTR ...

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

MATERIALS TESTING REACTOR (MTR) BUILDING, TRA-603. CONTEXTUAL VIEW OF MTR BUILDING SHOWING NORTH SIDES OF THE HIGH-BAY REACTOR BUILDING, ITS SECOND/THIRD FLOOR BALCONY LEVEL, AND THE ATTACHED ONE-STORY OFFICE/LABORATORY BUILDING, TRA-604. CAMERA FACING SOUTHEAST. VERTICAL CONCRETE-SHROUDED BEAMS SUPPORT PRECAST CONCRETE PANELS. CONCRETE PROJECTION FORMED AS A BUNKER AT LEFT OF VIEW IS TRA-657, PLUG STORAGE BUILDING. INL NEGATIVE NO. HD46-42-1. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Reactor Dosimetry State of the Art 2008

NASA Astrophysics Data System (ADS)

Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

2009-08-01

Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

Pressurized fluidized bed reactor and a method of operating the same

DOEpatents

Isaksson, J.

1996-02-20

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.

Improved vortex reactor system

DOEpatents

Diebold, James P.; Scahill, John W.

1995-01-01

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

Reactor monitoring using antineutrino detectors

NASA Astrophysics Data System (ADS)

Bowden, N. S.

2011-08-01

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

A brief history of design studies on innovative nuclear reactors

NASA Astrophysics Data System (ADS)

Sekimoto, Hiroshi

2014-09-01

In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970's the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980's the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.