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Evolution of Sarcoma 37 in cf1 Rats. Biological Aspects of the Spontaneous Cure of Tumors Studied After Irradiation and Administration of Immune Serum. Evolucao Do Sarcoma 37 Em Muganhos cf1. Aspectos Biologicos Da Cura Espontianea Do Tumor Estudados Apos Irradiacao E Administracao De Soro Imune

Description: Transplants of sarcoma 37 were studied in regard to spontaneous regression of the tumor after irradiation and after the administration of immune serum. The following results were obtained: 1. Of 809 animals, 99% developed the tumor after transplantation; 2. The number of cells transplanted did not influence the growth rate, the percentage of spontaneous regressions, or the life spans of the animals; 3. Reimplantation of the tumor into mice which had previously rejected the tumor resulted in very few successful transplants; 4. The resistance of these mice to further attempts at implantation could not be altered by previous irradiation; 5. Animals transplanted for the first time suffer a faster evolution of the tumor, shortened life span, and a diminution of spontaneous regressions if they have been irradiated prior to transplantation (400r whole body); and, 6. Resistance to tumor growth could not be transferred passively in serum of mice which had rejected the tumor.
Date: December 1, 1965
Creator: Clode, William H.

Reactivity of biogenic manganese oxide for metal sequestration and photochemistry: Computational solid state physics study

Description: Many microbes, including both bacteria and fungi, produce manganese (Mn) oxides by oxidizing soluble Mn(II) to form insoluble Mn(IV) oxide minerals, a kinetically much faster process than abiotic oxidation. These biogenic Mn oxides drive the Mn cycle, coupling it with diverse biogeochemical cycles and determining the bioavailability of environmental contaminants, mainly through strong adsorption and redox reactions. This mini review introduces recent findings based on quantum mechanical density functional theory that reveal the detailed mechanisms of toxic metal adsorption at Mn oxide surfaces and the remarkable role of Mn vacancies in the photochemistry of these minerals.
Date: February 1, 2010
Creator: Kwon, K.D. & Sposito, G.

Assessment of void swelling in austenitic stainless steel PWR core internals.

Description: As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of increasing attention. In this report, the available database on void swelling and density change of austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and microstructural characteristics were carefully examined, and key factors that are important to determine the relevance of the database to PWR conditions were evaluated. Most swelling data were obtained from steels irradiated in fast breeder reactors at temperatures >385 C and at dose rates that are orders of magnitude higher than PWR dose rates. Even for a given irradiation temperature and given steel, the integral effects of dose and dose rate on void swelling should not be separated. It is incorrect to extrapolate swelling data on the basis of 'progressive compounded multiplication' of separate effects of factors such as dose, dose rate, temperature, steel composition, and fabrication procedure. Therefore, the fast reactor data should not be extrapolated to determine credible void swelling behavior for PWR end-of-life (EOL) or life-extension conditions. Although the void swelling data extracted from fast reactor studies is extensive and conclusive, only limited amounts of swelling data and information have been obtained on microstructural characteristics from discharged PWR internals or steels irradiated at temperatures and at dose rates comparable to those of a PWR. Based on this relatively small amount of information, swelling in thin-walled tubes and baffle bolts in a PWR is not considered a concern. As additional data and relevant research becomes available, the newer results should be integrated with existing data, and the worthiness of this conclusion should continue to be scrutinized. PWR baffle reentrant corners are the most likely location to experience high swelling rates, and hence, high ...
Date: January 31, 2006
Creator: Chung, H. M. & Technology, Energy

Well-to-wheels analysis of energy use and greenhouse gas emissions of plug-in hybrid electric vehicles.

Description: Plug-in hybrid electric vehicles (PHEVs) are being developed for mass production by the automotive industry. PHEVs have been touted for their potential to reduce the US transportation sector's dependence on petroleum and cut greenhouse gas (GHG) emissions by (1) using off-peak excess electric generation capacity and (2) increasing vehicles energy efficiency. A well-to-wheels (WTW) analysis - which examines energy use and emissions from primary energy source through vehicle operation - can help researchers better understand the impact of the upstream mix of electricity generation technologies for PHEV recharging, as well as the powertrain technology and fuel sources for PHEVs. For the WTW analysis, Argonne National Laboratory researchers used the Greenhouse gases, Regulated Emissions, and Energy use in Transportation (GREET) model developed by Argonne to compare the WTW energy use and GHG emissions associated with various transportation technologies to those associated with PHEVs. Argonne researchers estimated the fuel economy and electricity use of PHEVs and alternative fuel/vehicle systems by using the Powertrain System Analysis Toolkit (PSAT) model. They examined two PHEV designs: the power-split configuration and the series configuration. The first is a parallel hybrid configuration in which the engine and the electric motor are connected to a single mechanical transmission that incorporates a power-split device that allows for parallel power paths - mechanical and electrical - from the engine to the wheels, allowing the engine and the electric motor to share the power during acceleration. In the second configuration, the engine powers a generator, which charges a battery that is used by the electric motor to propel the vehicle; thus, the engine never directly powers the vehicle's transmission. The power-split configuration was adopted for PHEVs with a 10- and 20-mile electric range because they require frequent use of the engine for acceleration and to provide energy when the battery is ...
Date: June 14, 2010
Creator: Elgowainy, A.; Han, J.; Poch, L.; Wang, M.; Vyas, A.; Mahalik, M. et al.

Well-to-Wheels analysis of landfill gas-based pathways and their addition to the GREET model.

Description: Today, approximately 300 million standard cubic ft/day (mmscfd) of natural gas and 1600 MW of electricity are produced from the decomposition of organic waste at 519 U.S. landfills (EPA 2010a). Since landfill gas (LFG) is a renewable resource, this energy is considered renewable. When used as a vehicle fuel, compressed natural gas (CNG) produced from LFG consumes up to 185,000 Btu of fossil fuel and generates from 1.5 to 18.4 kg of carbon dioxide-equivalent (CO{sub 2}e) emissions per million Btu of fuel on a 'well-to-wheel' (WTW) basis. This compares with approximately 1.1 million Btu and 78.2 kg of CO{sub 2}e per million Btu for CNG from fossil natural gas and 1.2 million Btu and 97.5 kg of CO{sub 2}e per million Btu for petroleum gasoline. Because of the additional energy required for liquefaction, LFG-based liquefied natural gas (LNG) requires more fossil fuel (222,000-227,000 Btu/million Btu WTW) and generates more GHG emissions (approximately 22 kg CO{sub 2}e /MM Btu WTW) if grid electricity is used for the liquefaction process. However, if some of the LFG is used to generate electricity for gas cleanup and liquefaction (or compression, in the case of CNG), vehicle fuel produced from LFG can have no fossil fuel input and only minimal GHG emissions (1.5-7.7 kg CO{sub 2}e /MM Btu) on a WTW basis. Thus, LFG-based natural gas can be one of the lowest GHG-emitting fuels for light- or heavy-duty vehicles. This report discusses the size and scope of biomethane resources from landfills and the pathways by which those resources can be turned into and utilized as vehicle fuel. It includes characterizations of the LFG stream and the processes used to convert low-Btu LFG into high-Btu renewable natural gas (RNG); documents the conversion efficiencies and losses of those processes, the choice of processes modeled in GREET, and ...
Date: June 30, 2010
Creator: Mintz, M.; Han, J.; Wang, M.; Saricks, C. & Systems, Energy

YALINA-booster subcritical assembly pulsed-neutron e xperiments: detector dead time and apatial corrections.

Description: In almost every detector counting system, a minimal dead time is required to record two successive events as two separated pulses. Due to the random nature of neutron interactions in the subcritical assembly, there is always some probability that a true neutron event will not be recorded because it occurs too close to the preceding event. These losses may become rather severe for counting systems with high counting rates, and should be corrected before any utilization of the experimental data. This report examines the dead time effects for the pulsed neutron experiments of the YALINA-Booster subcritical assembly. The nonparalyzable model is utilized to correct the experimental data due to dead time. Overall, the reactivity values are increased by 0.19$ and 0.32$ after the spatial corrections for the YALINA-Booster 36% and 21% configurations respectively. The differences of the reactivities obtained with He-3 long or short detectors at the same detector channel diminish after the dead time corrections of the experimental data for the 36% YALINA-Booster configuration. In addition, better agreements between reactivities obtained from different experimental data sets are also observed after the dead time corrections for the 21% YALINA-Booster configuration.
Date: October 11, 2010
Creator: Cao, Y.; Gohar, Y. & Division, Nuclear Engineering

YALINA-booster subcritical assembly pulsed-neutron experiments : data processing and spatial corrections.

Description: The YALINA-Booster experiments and analyses are part of the collaboration between Argonne National Laboratory of USA and the Joint Institute for Power & Nuclear Research - SOSNY of Belarus for studying the physics of accelerator driven systems for nuclear energy applications using low enriched uranium. The YALINA-Booster subcritical assembly is utilized for studying the kinetics of accelerator driven systems with its highly intensive D-T or D-D pulsed neutron source. In particular, the pulsed neutron methods are used to determine the reactivity of the subcritical system. This report examines the pulsed-neutron experiments performed in the YALINA-Booster facility with different configurations for the subcritical assembly. The 1141 configuration with 90% U-235 fuel and the 1185 configuration with 36% or 21% U-235 fuel are examined. The Sjoestrand area-ratio method is utilized to determine the reactivities of the different configurations. The linear regression method is applied to obtain the prompt neutron decay constants from the pulsed-neutron experimental data. The reactivity values obtained from the experimental data are shown to be dependent on the detector locations inside the subcritical assembly and the types of detector used for the measurements. In this report, Bell's spatial correction factors are calculated based on a Monte Carlo model to remove the detector dependences. The large differences between the reactivity values given by the detectors in the fast neutron zone of the YALINA-Booster are reduced after applying the spatial corrections. In addition, the estimated reactivity values after the spatial corrections are much less spatially dependent.
Date: October 11, 2010
Creator: Cao, Y.; Gohar, Y. & Division, Nuclear Engineering

ZPR-3 Assembly 11 : A cylindrical sssembly of highly enriched uranium and depleted uranium with an average {sup 235}U enrichment of 12 atom % and a depleted uranium reflector.

Description: Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 11 (ZPR-3/11) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 12 at.% and a depleted uranium reflector. Approximately 79.7% of the total fissions in this assembly occur above 100 keV, approximately 20.3% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor ...
Date: September 30, 2010
Creator: Lell, R. M.; McKnight, R. D.; Tsiboulia, A.; Rozhikhin, Y.; Security, National & Engineering, Inst. of Physics and Power

ZPR-3 Assembly 12 : A cylindrical assembly of highly enriched uranium, depleted uranium and graphite with an average {sup 235}U enrichment of 21 atom %.

Description: Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 12 (ZPR-3/12) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 21 at.%. Approximately 68.9% of the total fissions in this assembly occur above 100 keV, approximately 31.1% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 9 in the ...
Date: September 30, 2010
Creator: Lell, R. M.; McKnight, R. D.; Perel, R. L.; Wagschal, J. J.; Division, Nuclear Engineering & Physics, Racah Inst. of

ZPR-3 Assembly 6F : A spherical assembly of highly enriched uranium, depleted uranium, aluminum and steel with an average {sup 235}U enrichment of 47 atom %.

Description: Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 6 consisted of six phases, A through F. In each phase a critical configuration was constructed to simulate a very simple shape such as a slab, cylinder or sphere that could be analyzed with the limited analytical tools available in the 1950s. In each case the configuration consisted of a core region of metal plates surrounded by a thick depleted uranium metal reflector. The average ...
Date: September 30, 2010
Creator: Lell, R. M.; McKnight, R. D; Schaefer, R. W. & Division, Nuclear Engineering

2009 ALCF annual report.

Description: This year the Argonne Leadership Computing Facility (ALCF) delivered nearly 900 million core hours of science. The research conducted at their leadership class facility touched our lives in both minute and massive ways - whether it was studying the catalytic properties of gold nanoparticles, predicting protein structures, or unearthing the secrets of exploding stars. The authors remained true to their vision to act as the forefront computational center in extending science frontiers by solving pressing problems for our nation. Our success in this endeavor was due mainly to the Department of Energy's (DOE) INCITE (Innovative and Novel Computational Impact on Theory and Experiment) program. The program awards significant amounts of computing time to computationally intensive, unclassified research projects that can make high-impact scientific advances. This year, DOE allocated 400 million hours of time to 28 research projects at the ALCF. Scientists from around the world conducted the research, representing such esteemed institutions as the Princeton Plasma Physics Laboratory, National Institute of Standards and Technology, and European Center for Research and Advanced Training in Scientific Computation. Argonne also provided Director's Discretionary allocations for research challenges, addressing such issues as reducing aerodynamic noise, critical for next-generation 'green' energy systems. Intrepid - the ALCF's 557-teraflops IBM Blue/Gene P supercomputer - enabled astounding scientific solutions and discoveries. Intrepid went into full production five months ahead of schedule. As a result, the ALCF nearly doubled the days of production computing available to the DOE Office of Science, INCITE awardees, and Argonne projects. One of the fastest supercomputers in the world for open science, the energy-efficient system uses about one-third as much electricity as a machine of comparable size built with more conventional parts. In October 2009, President Barack Obama recognized the excellence of the entire Blue Gene series by awarding it to the National Medal ...
Date: November 23, 2010
Creator: Beckman, P.; Martin, D. & Drugan, C.

Advances in coupled safety modeling using systems analysis and high-fidelity methods.

Description: The potential for a sodium-cooled fast reactor to survive severe accident initiators with no damage has been demonstrated through whole-plant testing in EBR-II and FFTF. Analysis of the observed natural protective mechanisms suggests that they would be characteristic of a broad range of sodium-cooled fast reactors utilizing metal fuel. However, in order to demonstrate the degree to which new, advanced sodium-cooled fast reactor designs will possess these desired safety features, accurate, high-fidelity, whole-plant dynamics safety simulations will be required. One of the objectives of the advanced safety-modeling component of the Reactor IPSC is to develop a science-based advanced safety simulation capability by utilizing existing safety simulation tools coupled with emerging high-fidelity modeling capabilities in a multi-resolution approach. As part of this integration, an existing whole-plant systems analysis code has been coupled with a high-fidelity computational fluid dynamics code to assess the impact of high-fidelity simulations on safety-related performance. With the coupled capabilities, it is possible to identify critical safety-related phenomenon in advanced reactor designs that cannot be resolved with existing tools. In this report, the impact of coupling is demonstrated by evaluating the conditions of outlet plenum thermal stratification during a protected loss of flow transient. Outlet plenum stratification was anticipated to alter core temperatures and flows predicted during natural circulation conditions. This effect was observed during the simulations. What was not anticipated, however, is the far-reaching impact that resolving thermal stratification has on the whole plant. The high temperatures predicted at the IHX inlet due to thermal stratification in the outlet plenum forces heat into the intermediate system to the point that it eventually becomes a source of heat for the primary system. The results also suggest that flow stagnation in the intermediate system is possible, raising questions about the effectiveness of the intermediate decay heat removal systems in ...
Date: May 31, 2010
Creator: Fanning, T. H.; Thomas, J. W. & Division, Nuclear Engineering

Analysis of failed and nickel-coated 3093 beam clamp components at the East Tennessee Technology Park (ETTP).

Description: The U.S. Department of Energy and its contractor, Bechtel Jacobs Company (BJC), are undertaking a major effort to clean up the former gaseous diffusion facility (K-25) located in Oak Ridge, TN. The decontamination and decommissioning activities require systematic removal of contaminated equipment and machinery followed by demolition of the buildings. As part of the cleanup activities, a beam clamp, used for horizontal life lines (HLLs) for fall protection, was discovered to be fractured during routine inspection. The beam clamp (yoke and D-ring) was a component in the HLL system purchased from Reliance Industries LLC. Specifically, the U-shaped stainless steel yoke of the beam clamp failed in a brittle mode at under less than 10% of the rated design capacity of 14,500 lb. The beam clamp had been in service for approximately 16 months. Bechtel Jacobs approached Argonne National Laboratory to assist in identifying the root cause of the failure of the beam clamp. The objectives of this study were to (1) review the prior reports and documents on the subject, (2) understand the possible failure mechanism(s) that resulted in the failed beam clamp components, (3) recommend approaches to mitigate the failure mechanism(s), and (4) evaluate the modified beam clamp assemblies. Energy dispersive x-ray analysis and chemical analysis of the corrosion products on the failed yoke and white residue on an in-service yoke indicated the presence of zinc, sulfur, and calcium. Analysis of rainwater in the complex, as conducted by BJC, indicated the presence of sulfur and calcium. It was concluded that, as a result of galvanic corrosion, zinc from the galvanized components of the beam clamp assembly (D-ring) migrated to the corroded region in the presence of the rainwater. Under mechanical stress, the corrosion process would have accelerated, resulting in the catastrophic failure of the yoke. As suggested by Bechtel ...
Date: October 11, 2010
Creator: Singh, D.; Pappacena, K.; Gaviria, J.; Burtsteva, T. & Division, Nuclear Engineering

Annual report of monitoring at Morrill, Kansas, in 2009 .

Description: In September 2005, the Commodity Credit Corporation of the U.S. Department of Agriculture (CCC/USDA) initiated periodic sampling of groundwater in the vicinity of a grain storage facility formerly operated by the CCC/USDA at Morrill, Kansas. The sampling at Morrill is being performed on behalf of the CCC/USDA by Argonne National Laboratory, in accord with a monitoring program approved by the Kansas Department of Health and Environment (KDHE 2005), to monitor levels of carbon tetrachloride contamination identified in the groundwater at this site (Argonne 2004, 2005a). This report provides results for monitoring events in April and September 2009. Under the KDHE-approved monitoring plan (Argonne 2005b), groundwater was initially sampled twice yearly for a period of two years (in fall 2005, in spring and fall 2006, and in spring and fall 2007). The samples were analyzed for volatile organic compounds (VOCs), as well as for selected geochemical parameters to aid in the evaluation of possible natural contaminant degradation (reductive dechlorination) processes in the subsurface environment. The analytical results for groundwater sampling events at Morrill from September 2005 to October 2008 were documented previously (Argonne 2006a,b, 2007, 2008a,b, 2009). Those results consistently demonstrated the presence of carbon tetrachloride contamination, at levels exceeding the KDHE Tier 2 risk-based screening level of 5.0 {micro}g/L for this compound, in a groundwater plume extending generally south-southeastward from the former CCC/USDA facility, toward Terrapin Creek at the south edge of the town. Low levels ({le} 1.3 {micro}g/L) of carbon tetrachloride were persistently detected at monitoring well MW8S, on the bank of an intermittent tributary to Terrapin Creek. This observation suggested a possible risk of contamination of the surface waters of the creek. That concern is the regulatory driver for ongoing monitoring. In light of the early findings, in 2006 the CCC/USDA recommended expansion of the approved monitoring program ...
Date: August 5, 2010
Creator: LaFreniere, L. M. & Division, Environmental Science

APS Science 2009.

Description: It is my pleasure to introduce the 2009 annual report of the Advanced Photon Source. This was a very good year for us. We operated with high reliability and availability, despite growing problems with obsolete systems, and our users produced a record output of publications. The number of user experiments increased by 14% from 2008 to more than 3600. We congratulate the recipients of the 2009 Nobel Prize in Chemistry-Venkatraman Ramakrishnan (Cambridge Institute for Medical Research), Thomas Steitz (Yale University), and Ada Yonath (Weizmann Institute) - who did a substantial amount of this work at APS beamlines. Thanks to the efforts of our users and staff, and the ongoing counsel of the APS Scientific Advisory Committee, we made major progress in advancing our planning for the upgrade of the APS (APS-U), producing a proposal that was positively reviewed. We hope to get formal approval in 2010 to begin the upgrade. With advocacy from our users and the support of our sponsor, the Office of Basic Energy Sciences in the Department of Energy (DOE) Office of Science, our operating budgets have grown to the level needed to more adequately staff our beamlines. We were also extremely fortunate to have received $7.9 M in American Recovery and Reinvestment Act ('stimulus') funding to acquire new detectors and improve several of our beamlines. The success of the new Linac Coherent Light Source at Stanford, the world's first x-ray free-electron laser, made us particularly proud since the undulators were designed and built by the APS. Among other highlights, we note that more than one-quarter of the 46 Energy Frontier Research Centers, funded competitively across the U.S. in 2009 by the DOE, included the Advanced Photon Source in their proposed work, which shows that synchrotron radiation, and the APS in particular, are central to energy research. ...
Date: May 1, 2010
Creator: Gibson, J. M; Mills, D. M. & Gerig, R.

Argonne National Laboratory Site Environmental report for calendar year 2009.

Description: This report discusses the status and the accomplishments of the environmental protection program at Argonne National Laboratory for calendar year 2009. The status of Argonne environmental protection activities with respect to compliance with the various laws and regulations is discussed, along with the progress of environmental corrective actions and restoration projects. To evaluate the effects of Argonne operations on the environment, samples of environmental media collected on the site, at the site boundary, and off the Argonne site were analyzed and compared with applicable guidelines and standards. A variety of radionuclides were measured in air, surface water, on-site groundwater, and bottom sediment samples. In addition, chemical constituents in surface water, groundwater, and Argonne effluent water were analyzed. External penetrating radiation doses were measured, and the potential for radiation exposure to off-site population groups was estimated. Results are interpreted in terms of the origin of the radioactive and chemical substances (i.e., natural, Argonne, and other) and are compared with applicable environmental quality standards. A U.S. Department of Energy (DOE) dose calculation methodology, based on International Commission on Radiological Protection recommendations and the U.S. Environmental Protection Agency's (EPA) CAP-88 Version 3 (Clean Air Act Assessment Package-1988) computer code, was used in preparing this report.
Date: August 4, 2010
Creator: Golchert, N. W.; Davis, T. M. & Moos, L. P.

Atmospheric Radiation Measurement Program Climate Research Facility Operation quarterly report July 1 - September 30, 2010.

Description: Individual raw datastreams from instrumentation at the Atmospheric Radiation Measurement (ARM) Climate Research Facility fixed and mobile sites are collected and sent to the Data Management Facility (DMF) at Pacific Northwest National Laboratory (PNNL) for processing in near real-time. Raw and processed data are then sent approximately daily to the ARM Archive, where they are made available to users. For each instrument, we calculate the ratio of the actual number of data records received daily at the Archive to the expected number of data records. The results are tabulated by (1) individual datastream, site, and month for the current year and (2) site and fiscal year (FY) dating back to 1998. The U.S. Department of Energy (DOE) requires national user facilities to report time-based operating data. The requirements concern the actual hours of operation (ACTUAL); the estimated maximum operation or uptime goal (OPSMAX), which accounts for planned downtime; and the VARIANCE [1-(ACTUAL/OPSMAX)], which accounts for unplanned downtime. The OPSMAX time for the fourth quarter of FY2010 for the Southern Great Plains (SGP) site is 2097.60 hours (0.95 2208 hours this quarter). The OPSMAX for the North Slope of Alaska (NSA) locale is 1987.20 hours (0.90 2208) and for the Tropical Western Pacific (TWP) locale is 1876.80 hours (0.85 2208). The first ARM Mobile Facility (AMF1) deployment in Graciosa Island, the Azores, Portugal, continues, so the OPSMAX time this quarter is 2097.60 hours (0.95 x 2208). The differences in OPSMAX performance reflect the complexity of local logistics and the frequency of extreme weather events. It is impractical to measure OPSMAX for each instrument or datastream. Data availability reported here refers to the average of the individual, continuous datastreams that have been received by the Archive. Data not at the Archive are caused by downtime (scheduled or unplanned) of the individual instruments. ...
Date: October 26, 2010
Creator: Sisterson, D. L.

Benchmark specifications and data requirements for initial modeling of the China experimental fast reactor.

Description: A specification is proposed for an initial transient benchmark analysis of the China Experimental Fast Reactor design based on the analysis capabilities of the SAS4A/SASSYS-1 code. For the initial benchmark, a single-channel protected transient overpower accident is defined. Reactivity feedback coefficients will not be required and simplified material properties are recommended. This report also describes the data required for developing the modeling input. This data includes assembly geometry, reactor power distributions, kinetics and decay heat data, and material properties. Comparisons of benchmark results will take place at a future SAS4A/SASSYS-1 training meeting planned to occur at Argonne National Laboratory. Future benchmark specifications will be planned to expand upon this initial model to include more complex reactivity feedback models, material properties, additional assembly geometry, and primary and intermediate coolant systems.
Date: June 4, 2010
Creator: Fanning, T. H. & Division, Nuclear Engineering