Consolidated fuel reprocessing program

NASA Astrophysics Data System (ADS)

1985-04-01

A survey of electrochemical methods applications in fuel reprocessing was completed. A dummy fuel assembly shroud was cut using the remotely operated laser disassembly equipment. Operations and engineering efforts have continued to correct equipment operating, software, and procedural problems experienced during the previous uranium compaigns. Fuel cycle options were examined for the liquid metal reactor fuel cycle. In high temperature gas cooled reactor spent fuel studies, preconceptual designs were completed for the concrete storage cask and open field drywell storage concept. These and other tasks operating under the consolidated fuel reprocessing program are examined.

Processing of solid solution, mixed uranium/refractory metal carbides for advanced space nuclear power and propulsion systems

NASA Astrophysics Data System (ADS)

Knight, Travis Warren

Nuclear thermal propulsion (NTP) and space nuclear power are two enabling technologies for the manned exploration of space and the development of research outposts in space and on other planets such as Mars. Advanced carbide nuclear fuels have been proposed for application in space nuclear power and propulsion systems. This study examined the processing technologies and optimal parameters necessary to fabricate samples of single phase, solid solution, mixed uranium/refractory metal carbides. In particular, the pseudo-ternary carbide, UC-ZrC-NbC, system was examined with uranium metal mole fractions of 5% and 10% and corresponding uranium densities of 0.8 to 1.8 gU/cc. Efforts were directed to those methods that could produce simple geometry fuel elements or wafers such as those used to fabricate a Square Lattice Honeycomb (SLHC) fuel element and reactor core. Methods of cold uniaxial pressing, sintering by induction heating, and hot pressing by self-resistance heating were investigated. Solid solution, high density (low porosity) samples greater than 95% TD were processed by cold pressing at 150 MPa and sintering above 2600 K for times longer than 90 min. Some impurity oxide phases were noted in some samples attributed to residual gases in the furnace during processing. Also, some samples noted secondary phases of carbon and UC2 due to some hyperstoichiometric powder mixtures having carbon-to-metal ratios greater than one. In all, 33 mixed carbide samples were processed and analyzed with half bearing uranium as ternary carbides of UC-ZrC-NbC. Scanning electron microscopy, x-ray diffraction, and density measurements were used to characterize samples. Samples were processed from powders of the refractory mono-carbides and UC/UC 2 or from powders of uranium hydride (UH3), graphite, and refractory metal carbides to produce hypostoichiometric mixed carbides. Samples processed from the constituent carbide powders and sintered at temperatures above the melting point of UC showed signs of liquid phase sintering and were shown to be largely solid solutions. Pre-compaction of mixed carbide powders prior to sintering was shown to be necessary to achieve high densities. Hypostoichiometric, samples processed at 2500 K exhibited only the initial stage of sintering and solid solution formation. Based on these findings, a suggested processing methodology is proposed for producing high density, solid solution, mixed carbide fuels. Pseudo-binary, refractory carbide samples hot pressed at 3100 K and 6 MPa showed comparable densities (approximately 85% of the theoretical value) to samples processed by cold pressing and sintering at temperatures of 2800 K.

Reprocessing of research reactor fuel the Dounreay option

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

Cartwright, P.

1997-08-01

Reprocessing is a proven process for the treatment of spent U/Al Research Reactor fuel. At Dounreay 12679 elements have been reprocessed during the past 30 years. For reactors converting to LEU fuel the uranium recovered in reprocessing can be blended down to less than 20% U{sub 235}, enrichment and be fabricated into new elements. For reactors already converted to LEU it is technically possible to reprocess spent silicide fuel to reduce the U{sub 235} burden and present to a repository only stable conditioned waste. The main waste stream from reprocessing which contains the Fission products is collected in underground storagemore » tanks where it is kept for a period of at least five years before being converted to a stable solid form for return to the country of origin for subsequent storage/disposal. Discharges to the environment from reprocessing are low and are limited to the radioactive gases contained in the spent fuel and a low level liquid waste steam. Both of these discharges are independently monitored, and controlled within strict discharge limits set by the UK Government`s Scottish Office. Transportation of spent fuel to Dounreay has been undertaken using many routes from mainland Europe and has utilised over the past few years both chartered and scheduled vessel services. Several different transport containers have been handled and are currently licensed in the UK. This paper provides a short history of MTR reprocessing at Dounreay, and provides information to show reprocessing can satisfy the needs of MTR operators, showing that reprocessing is a valuable asset in non-proliferation terms, offers a complete solution and is environmentally acceptable.« less

Process for recovery of palladium from nuclear fuel reprocessing wastes

DOEpatents

Campbell, D.O.; Buxton, S.R.

1980-06-16

Palladium is selectively removed from spent nuclear fuel reprocessing waste by adding sugar to a strong nitric acid solution of the waste to partially denitrate the solution and cause formation of an insoluble palladium compound. The process includes the steps of: (a) adjusting the nitric acid content of the starting solution to about 10 M; (b) adding 50% sucrose solution in an amount sufficient to effect the precipitation of the palladium compound; (c) heating the solution at reflux temperature until precipitation is complete; and (d) centrifuging the solution to separate the precipitated palladium compound from the supernatant liquid.

Process for recovery of palladium from nuclear fuel reprocessing wastes

DOEpatents

Campbell, David O.; Buxton, Samuel R.

1981-01-01

Palladium is selectively removed from spent nuclear fuel reprocessing waste by adding sugar to a strong nitric acid solution of the waste to partially denitrate the solution and cause formation of an insoluble palladium compound. The process includes the steps of: (a) adjusting the nitric acid content of the starting solution to about 10 M, (b) adding 50% sucrose solution in an amount sufficient to effect the precipitation of the palladium compound, (c) heating the solution at reflux temperature until precipitation is complete, and (d) centrifuging the solution to separate the precipitated palladium compound from the supernatant liquid.

A Systematic Theoretical Study of UC6: Structure, Bonding Nature, and Spectroscopy.

PubMed

Du, Jiguang; Jiang, Gang

2017-11-20

The study of uranium carbides has received renewed attention in recent years due to the potential use of these compounds as fuels in new generations of nuclear reactors. The isomers of the UC 6 cluster were determined by DFT and ab initio methods. The structures obtained using SC-RECP for U were generally consistent with those obtained using an all-electron basis set (ZORA-SARC). The CCSD(T) calculations indicated that two isomers had similar energies and may coexist in laser evaporation experiments. The nature of the U-C bonds in the different isomers was examined via a topological analysis of the electron density, and the results indicated that the U-C bonds are predominantly closed-shell (ionic) interactions with a certain degree of covalent character in all cases, particularly in the linear species. The IR and UV-vis spectra of the isomers were theoretically simulated to provide information that can be used to identify the isomers of UC 6 in future experiments.

Alloy 33: A new material for the handling of HNO{sub 3}/HF media in reprocessing of nuclear fuel

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

Koehler, M.; Heubner, U.; Eichenhofer, K.W.

Alloy 33, an austenitic 33Cr-32Fe-31Ni-1.6Mo-0.6Cu-0.4N material shows excellent resistance to corrosion when exposed to highly oxidizing media as e.g. HNO{sub 3} and HNO{sub 3}/HF mixtures which are encountered in reprocessing of nuclear fuel. According to the test results available so far, resistance to corrosion in boiling azeotropic (67%) HNO{sub 3} is about 6 and 2 times superior to AISI 304 L and 310 L. In higher concentrated nitric acid it can be considered corrosion resistant up to 95% HNO{sub 3} at 25 C, up to 90% HNO{sub 3} at 50 C and up to somewhat less than 85% HNO{sub 3}more » at 75 C. In 20% HNO{sub 3}/7% HF at 50 C its resistance to corrosion is superior to AISI 316 Ti and Alloy 28 by factors of about 200 and 2.4. Other media tested with different results include 12% HNO{sub 3} with up to 3.5% HF and 0.4% HF with 32 to 67.5% HNO{sub 3} at 90 C. Alloy 33 is easily fabricated into all product forms required for chemical plants (e.g. plate, sheet, strip, wire, tube and flanges). Components such as dished ends and tube to tube sheet weldments have been successfully fabricated facilitating the use of Alloy 33 for reprocessing of nuclear fuel.« less

Methods for making a porous nuclear fuel element

DOEpatents

Youchison, Dennis L; Williams, Brian E; Benander, Robert E

2014-12-30

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

An assessment of the attractiveness of material associated with thorium/uranium and uranium closed fuel cycles from a safeguards perspective

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

Bathke, Charles Gary; Wallace, Richard K; Hase, Kevin R

2010-01-01

This paper reports the continued evaluation of the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with various proposed nuclear fuel cycles. Specifically, this paper examines two closed fuel cycles. The first fuel cycle examined is a thorium fuel cycle in which a pressurized heavy water reactor (PHWR) is fueled with mixtures of plutonium/thorium and {sup 233}U/thorium. The used fuel is then reprocessed using the THOREX process and the actinides are recycled. The second fuel cycle examined consists of conventional light water reactors (LWR) whose fuel is reprocessed for actinides that are then fed to and recycled untilmore » consumed in fast-spectrum reactors: fast reactors and accelerator driven systems (ADS). As reprocessing of LWR fuel has already been examined, this paper will focus on the reprocessing of the scheme's fast-spectrum reactors' fuel. This study will indicate what is required to render these materials as having low utility for use in nuclear weapons. Nevertheless, the results of this paper suggest that all reprocessing products evaluated so far need to be rigorously safeguarded and provided high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE). The methodology and key findings will be presented.« less

Continental-scale enrichment of atmospheric 14CO2 from the nuclear power industry: potential impact on the estimation of fossil fuel-derived CO2

NASA Astrophysics Data System (ADS)

Graven, H. D.; Gruber, N.

2011-12-01

The 14C-free fossil carbon added to atmospheric CO2 by combustion dilutes the atmospheric 14C/C ratio (Δ14C), potentially providing a means to verify fossil CO2 emissions calculated using economic inventories. However, sources of 14C from nuclear power generation and spent fuel reprocessing can counteract this dilution and may bias 14C/C-based estimates of fossil fuel-derived CO2 if these nuclear influences are not correctly accounted for. Previous studies have examined nuclear influences on local scales, but the potential for continental-scale influences on Δ14C has not yet been explored. We estimate annual 14C emissions from each nuclear site in the world and conduct an Eulerian transport modeling study to investigate the continental-scale, steady-state gradients of Δ14C caused by nuclear activities and fossil fuel combustion. Over large regions of Europe, North America and East Asia, nuclear enrichment may offset at least 20% of the fossil fuel dilution in Δ14C, corresponding to potential biases of more than -0.25 ppm in the CO2 attributed to fossil fuel emissions, larger than the bias from plant and soil respiration in some areas. Model grid cells including high 14C-release reactors or fuel reprocessing sites showed much larger nuclear enrichment, despite the coarse model resolution of 1.8°×1.8°. The recent growth of nuclear 14C emissions increased the potential nuclear bias over 1985-2005, suggesting that changing nuclear activities may complicate the use of Δ14C observations to identify trends in fossil fuel emissions. The magnitude of the potential nuclear bias is largely independent of the choice of reference station in the context of continental-scale Eulerian transport and inversion studies, but could potentially be reduced by an appropriate choice of reference station in the context of local-scale assessments.

Implementation and evaluation of fuel creep using advanced light-water reactor materials in FRAPCON 3.5

NASA Astrophysics Data System (ADS)

Carroll, Spencer

As current reactors approach the end of their operable lifetime, new reactors are needed if nuclear power is to continue being generated in the United States. Some utilities have already began construction on newer, more advanced LWR reactors, which use the same fuel as current reactors and have a similar but updated design. Others are researching next generation (GEN-IV) reactors which have new designs that utilize alternative fuel, coolants and other reactor materials. Many of these alternative fuels are capable of achieving higher burnups and are designed to be more accident tolerant than the currently used UO2 fuel. However, before these new materials can be used, extensive research must be done in order to obtain a detailed understanding of how the new fuels and other materials will interact. New fuels, such as uranium nitride (UN) and uranium carbide (UC) have several advantages over UO2, such as increased burnup capabilities and higher thermal conductivities. However, there are issues with each that prevent UC and UN from being used as direct replacements for UO2. Both UC and UN swell at a significantly higher rate than UO2 and neither fuel reacts favorably when exposed to water. Due to this, UC and UN are being considered more for GEN-IV reactors that use alternative coolant rather than for current LWRs. In an effort to increase accident tolerance, silicon carbide (SiC) is being considered for use as an alternative cladding. The high strength, high melting point and low oxidation of SiC make it an attractive cladding choice, especially in an accident scenario. However, as a ceramic, SiC is not ductile and will not creep outwards upon pellet-clad mechanical interaction (PCMI) which can cause a large build up in interfacial pressure. In order to understand the interaction between the high swelling fuels and unyielding SiC cladding, data on the properties and behaviors of these materials must be gathered and incorporated into FRAPCON. FRAPCON is a fuel performance code developed by PNNL and used by the Nuclear Regulatory Commission (NRC) as a licensing code for US reactors. FRAPCON will give insight into how these new fuel-cladding combinations will affect cladding hoop stress and help determine if the new materials are feasible for use in a reactor. To accurately simulate the interaction between the new materials, a soft pellet model that allows for stresses on the pellet to affect pellet deformation will have to be implemented. Currently, FRAPCON uses a rigid pellet model that does not allow for feedback of the cladding onto the pellet. Since SiC does not creep at the temperatures being considered and is not ductile, any PCMI create a much higher interfacial pressure than is possible with Zircaloy. Because of this, it is necessary to implement a model that allows for pellet creep to alleviate some of these cladding stresses. These results will then be compared to FEMAXI-6, a Japanese fuel performance code that already calculates pellet stress and allows for cladding feedback onto the pellet. This research is intended to be a continuation and verification of previous work done by USC on the analysis of accident tolerant fuels with alternative claddings and is intended to prove that a soft pellet model is necessary to accurately model any fuel with SiC cladding.

Uranium nitride as LWR TRISO fuel: Thermodynamic modeling of U-C-N

NASA Astrophysics Data System (ADS)

Besmann, Theodore M.; Shin, Dongwon; Lindemer, Terrence B.

2012-08-01

TRISO coated particle fuel is envisioned as a next generation replacement for current urania pellet fuel in LWR applications. To obtain adequate fissile loading the kernel of the TRISO particle will likely need to be UN instead of UO2. In support of the necessary development effort for this new fuel system, an assessment of phase regions of interest in the U-C-N system was undertaken as the fuel will be prepared by the carbothermic reduction of the oxide followed by nitriding, will be in equilibrium with carbon within the TRISO particle, and will react with minor actinides and fission products. The phase equilibria and thermochemistry of the U-C-N system is reviewed, including nitrogen pressure measurements above various phase fields. Measurements were used to confirm an ideal solution model of UN and UC adequately represents the UC1-xNx phase. Agreement with the data was significantly improved by effectively adjusting the Gibbs free energy of UN by +12 kJ/mol. This also required adjustment of the value for the sesquinitride by +17 kJ/mol to obtain agreement with phase equilibria. The resultant model together with reported values for other phases in the system was used to generate isothermal sections of the U-C-N phase diagram. Nitrogen partial pressures were also computed for regions of interest.

Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

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

Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

1976-04-01

Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

Determining the minimum required uranium carbide content for HTGR UCO fuel kernels

DOE PAGES

McMurray, Jacob W.; Lindemer, Terrence B.; Brown, Nicholas R.; ...

2017-03-10

There are three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from free O generated when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. Furthermore, in the HTGR UCO kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium in the form of a carbide, UC x. An approach for determining the minimum UC xmore » content to ensure negligible CO formation was developed and demonstrated using CALPHAD models and the Serpent 2 reactor physics and depletion analysis tool. Our results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmutation products on the oxygen distribution as the fuel kernel composition evolves with burnup.« less

New Concepts for Compact Space Reactor Power Systems for Space Based Radar Applications: A Feasibility Study

DTIC Science & Technology

1989-12-01

SPENT FUEL REPROCESSING COULD ALSO BE EMPLOYED IRRADIATION EXPERIENCE - EXTREMELY LIMITED - JOINT US/UK PROGRAM (ONGOING) - TUI/KFK PROGRAM (CANCELED...only the use of off-the-shelf technologies. For example, conventional fuel technology (uranium dioxide), conventional thermionic conversion...advanced fuel (Americium oxide, A1TI2O3) and advanced thermionic conversion. Concept C involves use of an advanced fuel (Americium oxide, Arri203

Thermal Aspects of Using Alternative Nuclear Fuels in Supercritical Water-Cooled Reactors

NASA Astrophysics Data System (ADS)

Grande, Lisa Christine

A SuperCritical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The current research presents a thermal-hydraulic analysis of a single fuel channel within a Pressure Tube (PT)-type SCWR with a single-reheat cycle. Since this reactor is in its early design phase many fuel-channel components are being investigated in various combinations. Analysis inputs are: steam cycle, Axial Heat Flux Profile (AHFP), fuel-bundle geometry, and thermophysical properties of reactor coolant, fuel sheath and fuel. Uniform and non-uniform AHFPs for average channel power were applied to a variety of alternative fuels (mixed oxide, thorium dioxide, uranium dicarbide, uranium nitride and uranium carbide) enclosed in an Inconel-600 43-element bundle. The results depict bulk-fluid, outer-sheath and fuel-centreline temperature profiles together with the Heat Transfer Coefficient (HTC) profiles along the heated length of fuel channel. The objective is to identify the best options in terms of fuel, sheath material and AHFPS in which the outer-sheath and fuel-centreline temperatures will be below the accepted temperature limits of 850°C and 1850°C respectively. The 43-element Inconel-600 fuel bundle is suitable for SCWR use as the sheath-temperature design limit of 850°C was maintained for all analyzed cases at average channel power. Thoria, UC2, UN and UC fuels for all AHFPs are acceptable since the maximum fuel-centreline temperature does not exceed the industry accepted limit of 1850°C. Conversely, the fuel-centreline temperature limit was exceeded for MOX at all AHFPs, and UO2 for both cosine and downstream-skewed cosine AHFPs. Therefore, fuel-bundle modifications are required for UO2 and MOX to be feasible nuclear fuels for SCWRs.

Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pocoima, CA; Benander, Robert E [Pacoima, CA

2010-02-23

Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

2011-03-01

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

2013-09-03

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

76 FR 34007 - Draft Regulatory Basis for a Potential Rulemaking on Spent Nuclear Fuel Reprocessing Facilities

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-10

... processes are more akin to fuel cycle processes. This framework was established in the 1970's to license the... nuclear power globally and close the nuclear fuel cycle through reprocessing spent fuel and deploying fast... Accounting;'' and a Nuclear Energy Institute white [[Page 34009

ARCHITECTURAL SECTIONS A, B, C, D, OF HOT PILOT PLANT ...

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

ARCHITECTURAL SECTIONS A, B, C, D, OF HOT PILOT PLANT (CPP-640). INL DRAWING NUMBER 200-0640-00-279-111681. ALTERNATE ID NUMBER 8952-CPP-640-A-5. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Rare Earth Electrochemical Property Measurements and Phase Diagram Development in a Complex Molten Salt Mixture for Molten Salt Recycle

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

Zhang, Jinsuo; Guo, Shaoqiang

Pyroprocessing is a promising alternative for the reprocessing of used nuclear fuel (UNF) that uses electrochemical methods. Compared to the hydrometallurgical reprocessing method, pyroprocessing has many advantages such as reduced volume of radioactive waste, simple waste processing, ability to treat refractory material, and compatibility with fast reactor fuel recycle. The key steps of the process are the electro-refining of the spent metallic fuel in the LiCl-KCl eutectic salt, which can be integrated with an electrolytic reduction step for the reprocessing of spent oxide fuels.

Laser-enhanced chemical reactions and the liquid state. II. Possible applications to nuclear fuel reprocessing

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

DePoorter, G.L.; Rofer-DePoorter, C.K.

1976-01-01

Laser photochemistry is surveyed as a possible improvement upon the Purex process for reprocessing spent nuclear fuel. Most of the components of spent nuclear fuel are photochemically active, and lasers can be used to selectively excite individual chemical species. The great variety of chemical species present and the degree of separation that must be achieved present difficulties in reprocessing. Lasers may be able to improve the necessary separations by photochemical reaction or effects on rates and equilibria of reactions. (auth)

Carbothermic Synthesis of ~820- m UN Kernels. Investigation of Process Variables

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

Lindemer, Terrence; Silva, Chinthaka M; Henry, Jr, John James

2015-06-01

This report details the continued investigation of process variables involved in converting sol-gel-derived, urainia-carbon microspheres to ~820-μm-dia. UN fuel kernels in flow-through, vertical refractory-metal crucibles at temperatures up to 2123 K. Experiments included calcining of air-dried UO 3-H 2O-C microspheres in Ar and H 2-containing gases, conversion of the resulting UO 2-C kernels to dense UO 2:2UC in the same gases and vacuum, and its conversion in N 2 to in UC 1-xN x. The thermodynamics of the relevant reactions were applied extensively to interpret and control the process variables. Producing the precursor UO 2:2UC kernel of ~96% theoretical densitymore » was required, but its subsequent conversion to UC 1-xN x at 2123 K was not accompanied by sintering and resulted in ~83-86% of theoretical density. Decreasing the UC 1-xN x kernel carbide component via HCN evolution was shown to be quantitatively consistent with present and past experiments and the only useful application of H2 in the entire process.« less

Nuclear Fuel Reprocessing

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

Harold F. McFarlane; Terry Todd

2013-11-01

Reprocessing is essential to closing nuclear fuel cycle. Natural uranium contains only 0.7 percent 235U, the fissile (see glossary for technical terms) isotope that produces most of the fission energy in a nuclear power plant. Prior to being used in commercial nuclear fuel, uranium is typically enriched to 3–5% in 235U. If the enrichment process discards depleted uranium at 0.2 percent 235U, it takes more than seven tonnes of uranium feed to produce one tonne of 4%-enriched uranium. Nuclear fuel discharged at the end of its economic lifetime contains less one percent 235U, but still more than the natural ore.more » Less than one percent of the uranium that enters the fuel cycle is actually used in a single pass through the reactor. The other naturally occurring isotope, 238U, directly contributes in a minor way to power generation. However, its main role is to transmute into plutoniumby neutron capture and subsequent radioactive decay of unstable uraniumand neptuniumisotopes. 239Pu and 241Pu are fissile isotopes that produce more than 40% of the fission energy in commercially deployed reactors. It is recovery of the plutonium (and to a lesser extent the uranium) for use in recycled nuclear fuel that has been the primary focus of commercial reprocessing. Uraniumtargets irradiated in special purpose reactors are also reprocessed to obtain the fission product 99Mo, the parent isotope of technetium, which is widely used inmedical procedures. Among the fission products, recovery of such expensive metals as platinum and rhodium is technically achievable, but not economically viable in current market and regulatory conditions. During the past 60 years, many different techniques for reprocessing used nuclear fuel have been proposed and tested in the laboratory. However, commercial reprocessing has been implemented along a single line of aqueous solvent extraction technology called plutonium uranium reduction extraction process (PUREX). Similarly, hundreds of types of reactor fuels have been irradiated for different purposes, but the vast majority of commercial fuel is uranium oxide clad in zirconium alloy tubing. As a result, commercial reprocessing plants have relatively narrow technical requirements for used nuclear that is accepted for processing.« less

UC Davis Fuel Cell, Hydrogen, and Hybrid Vehicle (FCH2V) GATE Center of Excellence

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

Erickson, Paul

This is the final report of the UC Davis Fuel Cell, Hydrogen, and Hybrid Vehicle (FCH2V) GATE Center of Excellence which spanned from 2005-2012. The U.S. Department of Energy (DOE) established the Graduate Automotive Technology Education (GATE) Program, to provide a new generation of engineers and scientists with knowledge and skills to create advanced automotive technologies. The UC Davis Fuel Cell, Hydrogen, and Hybrid Vehicle (FCH2V) GATE Center of Excellence established in 2005 is focused on research, education, industrial collaboration and outreach within automotive technology. UC Davis has had two independent GATE centers with separate well-defined objectives and research programsmore » from 1998. The Fuel Cell Center, administered by ITS-Davis, has focused on fuel cell technology. The Hybrid-Electric Vehicle Design Center (HEV Center), administered by the Department of Mechanical and Aeronautical Engineering, has focused on the development of plug-in hybrid technology using internal combustion engines. The merger of these two centers in 2005 has broadened the scope of research and lead to higher visibility of the activity. UC Davis's existing GATE centers have become the campus's research focal points on fuel cells and hybrid-electric vehicles, and the home for graduate students who are studying advanced automotive technologies. The centers have been highly successful in attracting, training, and placing top-notch students into fuel cell and hybrid programs in both industry and government.« less

Processes for Removal and Immobilization of 14C, 129I, and 85Kr

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

Strachan, Denis M.; Bryan, Samuel A.; Henager, Charles H.

2009-10-05

This is a white paper covering the results of a literature search and preliminary experiments on materials and methods to remove and immobilize gaseous radionuclided that come from the reprocessing of spent nuclear fuel.

THE ECONOMICS OF REPROCESSING vs DIRECT DISPOSAL OF SPENT NUCLEAR FUEL

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

Matthew Bunn; Steve Fetter; John P. Holdren

This report assesses the economics of reprocessing versus direct disposal of spent nuclear fuel. The breakeven uranium price at which reprocessing spent nuclear fuel from existing light-water reactors (LWRs) and recycling the resulting plutonium and uranium in LWRs would become economic is assessed, using central estimates of the costs of different elements of the nuclear fuel cycle (and other fuel cycle input parameters), for a wide range of range of potential reprocessing prices. Sensitivity analysis is performed, showing that the conclusions reached are robust across a wide range of input parameters. The contribution of direct disposal or reprocessing and recyclingmore » to electricity cost is also assessed. The choice of particular central estimates and ranges for the input parameters of the fuel cycle model is justified through a review of the relevant literature. The impact of different fuel cycle approaches on the volume needed for geologic repositories is briefly discussed, as are the issues surrounding the possibility of performing separations and transmutation on spent nuclear fuel to reduce the need for additional repositories. A similar analysis is then performed of the breakeven uranium price at which deploying fast neutron breeder reactors would become competitive compared with a once-through fuel cycle in LWRs, for a range of possible differences in capital cost between LWRs and fast neutron reactors. Sensitivity analysis is again provided, as are an analysis of the contribution to electricity cost, and a justification of the choices of central estimates and ranges for the input parameters. The equations used in the economic model are derived and explained in an appendix. Another appendix assesses the quantities of uranium likely to be recoverable worldwide in the future at a range of different possible future prices.« less

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

Besmann, Theodore M; Shin, Dongwon

TRISO coated particle fuel is envisioned as a next generation replacement for current urania pellet fuel in LWR applications. To obtain adequate fissile loading the kernel of the TRISO particle will need to be UN. In support of the fuel development effort, an assessment of phase regions of interest in the U-C-N system was undertaken as the fuel will be prepared by the carbothermic reduction of the oxide and it will be in equilibrium with carbon within the TRISO particle. The phase equilibria and thermochemistry of the U-C-N system is reviewed, including nitrogen pressure measurements above various phase fields. Selectedmore » measurements were used to fit a first order model of the UC1-xNx phase, represented by the inter-solution of UN and UC. Fit to the data was significantly improved by also adjusting the heat of formation for UN by ~12 kJ/mol and the phase equilbria was best reproduced by also adjusting the heat for U2N3 by +XXX. The determined interaction parameters yielded a slightly positive deviation from ideality, which agrees with lattice parameter measurements which show positive deviation from Vegard s law. The resultant model together with reported values for other phases in the system were used to generate isothermal sections of the U-C-N phase diagram. Nitrogen partial pressures were also computed for regions of interest.« less

HIGH DENSITY NUCLEAR FUEL COMPOSITION

DOEpatents

Litton, F.B.

1962-07-17

ABS>A nuclear fuel consisting essentially of uranium monocarbide and containing 2.2 to 4.6 wt% carbon, 0.1 to 2.3 wt% oxygen, 0.05 to 2.5 wt% nitrogen, and the balance uranium was developed. The maximum oxygen content was less than one-half the carbon content by weight and the carbon, oxygen, and nitrogen are present as a single phase substituted solid solution of UC, C, O, and N. A method of preparing the fuel composition is described. (AEC)

U.S.-South Korea Relations

DTIC Science & Technology

2013-04-26

reprocessed to make new fuel using a type of reprocessing called pyroprocessing .66 The United States and South Korea are jointly researching pyroprocessing ...solutions to spent fuel disposal. Spent fuel disposal is a key policy issue for South Korean officials, and some see pyroprocessing as a potential solution...proponents of pyroprocessing see it as a way to advance energy independence for South Korea. 66

Evaluation of Ruthenium Capture Methods for Tritium Pretreatment Off-Gas Streams

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

Spencer, Barry B.; Jubin, Robert Thomas; Bruffey, Stephanie H.

2017-07-01

In the reprocessing of used nuclear fuel, radioactive elements are released into various plant off-gas streams. While much research and development has focused on the abatement of the volatile nuclides 3H, 14C, 85Kr, and 129I, the potential release of semivolatile isotopes that could also report to the off-gas streams in a reprocessing facility has been examined. Ruthenium (as 106Ru) has been identified as one of the semivolatile nuclides requiring the greatest degree of abatement prior to discharging the plant off-gas to the environment.

Study of the Optimum Zone of the Independent Variables of an ORGEL Reactor Connected to a 250-MWeb Power Plant. Self Supporting Fuel Elements Made of UC, with Sap Cladding with Four Fuel Rods and Individual Pressure Tubes; STUDIE DER OPTIMALEN ZONE DER UNABHANGIGEN PARAMETER EINES ORGEL- REAKTORS IN EINEM 250-MWe-KRAFTWERK. SELBST-TRAGENDES BRENNELEMENT AUS UC, SAP-UMHUL-LUNG MIT 4 BRENNSTOFFSTABEN UND INDIVIDUELLEN DRUCKROHREN

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

LaFontaine, F.; Tauch, P.

The optimum range of the independent variables of and ORGEL reactor connected to a 250-Mw power plant (4 fuel rods of UC with individual pressure tubes), as well as the geometry of the reactor core and the operation of the plant, is described. (auth)

Multiple recycle of REMIX fuel based on reprocessed uranium and plutonium mixture in thermal reactors

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

Fedorov, Y.S.; Bibichev, B.A.; Zilberman, B.Y.

2013-07-01

REMIX fuel consumption in WWER-1000 is considered. REMIX fuel is fabricated from non-separated mixture of uranium and plutonium obtained during NPP spent fuel reprocessing with further makeup by enriched natural uranium. It makes possible to recycle several times the total amount of uranium and plutonium obtained from spent fuel with 100% loading of the WWER-1000 core. The stored SNF could be also involved in REMIX fuel cycle by enrichment of regenerated uranium. The same approach could be applied to closing the fuel cycle of CANDU reactors. (authors)

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

McMurray, Jacob W.; Lindemer, Terrence B.; Brown, Nicholas R.

There are three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from free O generated when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. Furthermore, in the HTGR UCO kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium in the form of a carbide, UC x. An approach for determining the minimum UC xmore » content to ensure negligible CO formation was developed and demonstrated using CALPHAD models and the Serpent 2 reactor physics and depletion analysis tool. Our results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmutation products on the oxygen distribution as the fuel kernel composition evolves with burnup.« less

Availability analysis of an HTGR fuel recycle facility. Summary report

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

Sharmahd, J.N.

1979-11-01

An availability analysis of reprocessing systems in a high-temperature gas-cooled reactor (HTGR) fuel recycle facility was completed. This report summarizes work done to date to define and determine reprocessing system availability for a previously planned HTGR recycle reference facility (HRRF). Schedules and procedures for further work during reprocessing development and for HRRF design and construction are proposed in this report. Probable failure rates, transfer times, and repair times are estimated for major system components. Unscheduled down times are summarized.

Radial flow nuclear thermal rocket (RFNTR)

DOEpatents

Leyse, Carl F.

1995-11-07

A radial flow nuclear thermal rocket fuel assembly includes a substantially conical fuel element having an inlet side and an outlet side. An annular channel is disposed in the element for receiving a nuclear propellant, and a second, conical, channel is disposed in the element for discharging the propellant. The first channel is located radially outward from the second channel, and separated from the second channel by an annular fuel bed volume. This fuel bed volume can include a packed bed of loose fuel beads confined by a cold porous inlet frit and a hot porous exit frit. The loose fuel beads include ZrC coated ZrC-UC beads. In this manner, nuclear propellant enters the fuel assembly axially into the first channel at the inlet side of the element, flows axially across the fuel bed volume, and is discharged from the assembly by flowing radially outward from the second channel at the outlet side of the element.

Radial flow nuclear thermal rocket (RFNTR)

DOEpatents

Leyse, Carl F.

1995-01-01

A radial flow nuclear thermal rocket fuel assembly includes a substantially conical fuel element having an inlet side and an outlet side. An annular channel is disposed in the element for receiving a nuclear propellant, and a second, conical, channel is disposed in the element for discharging the propellant. The first channel is located radially outward from the second channel, and separated from the second channel by an annular fuel bed volume. This fuel bed volume can include a packed bed of loose fuel beads confined by a cold porous inlet frit and a hot porous exit frit. The loose fuel beads include ZrC coated ZrC-UC beads. In this manner, nuclear propellant enters the fuel assembly axially into the first channel at the inlet side of the element, flows axially across the fuel bed volume, and is discharged from the assembly by flowing radially outward from the second channel at the outlet side of the element.

Closed Fuel Cycle Waste Treatment Strategy

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

Vienna, J. D.; Collins, E. D.; Crum, J. V.

This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significantmore » additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals; Carbon-14 immobilized as a CaCO3 in a cement waste form; Krypton-85 stored as a compressed gas; An aqueous reprocessing high-level waste (HLW) raffinate waste immobilized by the vitrification process; An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel either included in the borosilicate HLW glass or immobilized in the form of a metal alloy or titanate ceramics; Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware super-compacted for disposal or purified for reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations; Electrochemical process salt HLW incorporated into a glass bonded Sodalite waste form; and Electrochemical process UDS and SS cladding hulls melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.« less

Cryochemical and CVD processing of shperical carbide fuels for propulsion reactors

NASA Astrophysics Data System (ADS)

Blair, H. Thomas; Carroll, David W.; Matthews, R. Bruce

1991-01-01

Many of the nuclear propulsion reactor concepts proposed for a manned mission to Mars use a coated spherical particle fuel form similar to that used in the Rover and NERVA propulsion reactors. The formation of uranium dicarbide microspheres using a cryochemical process and the coating of the UC2 spheres with zirconium carbide using chemical vapor deposition are being developed at Los Alamos National Laboratory. The cryochemical process is described with a discussion of the variables affecting the sphere formation and carbothermic reduction to produce UC2 spheres from UO2. Emphasis is placed on minimizing the wastes produced by the process. The ability to coat particles with ZrC was recaptured, and improvements in the process and equipment were developed. Volatile organometallic precursors were investigated as alternatives to the original ZrCl4 precursor.

DESIGN CRITERIA FOR FUEL DISSOLUTION SYSTEMS AND ASSOCIATED SERVICE FACILITIES. PLANT MODIFICATIONS FOR REPROCESSING NON-PRODUCTION REACTOR FUELS. PROJECT CGC-830

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

Bierman, S.R.; Graf, W.A.; Kass, M.

1960-07-29

Design panameters are presented for phases of the facility to reprocess low-enrichment fuels from nonproduction reactors. Included are plant flowsheets and equipment layouts for fuel element dissolution, centrifugation, solution adjustment, and waste handling. Also included are the basic design criteria for the supporting facilities which service these phases and all other facilites located in the vicinity of the selected building (Bldg. 221-U). (J.R.D.)

Apparatus and method for reprocessing and separating spent nuclear fuels. [Patent application

DOEpatents

Krikorian, O.H.; Grens, J.Z.; Parrish, W.H. Sr.

1982-01-19

Spent nuclear fuels, including actinide fuels, volatile and non-volatile fission products, are reprocessed and separated in a molten metal solvent housed in the reaction region of a separation vessel which includes a reflux region positioned above the molten tin solvent. The reflux region minimizes loss of evaporated solvent during the separation of the actinide fuels from the volatile fission products. Additionally, inclusion of the reflux region permits the separation of the more volatile fission products (noncondensable) from the less volatile ones (condensable).

Assessing the effectiveness of safeguards at a medium-sized spent-fuel reprocessing facility

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

Higinbotham, W.; Fishbone, L.G.; Suda, S.

1983-01-01

In order to evaluate carefully and systematically the effectiveness of safeguards at nuclear-fuel-cycle facilities, the International Atomic Energy Agency has adopted a safeguards effectiveness assessment methodology. The methodology has been applied to a well-characterized, medium-sized, spent-fuel reprocessing plant to understand how explicit safeguards inspection procedures would serve to expose conceivable nuclear materials diversion schemes, should such diversion occur.

Dismantling of the 904 Cell at the HAO/Sud Facility - 13466

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

Vaudey, C.E.; Crosnier, S.; Renouf, M.

2013-07-01

La Hague facility, in France, is the spent fuel recycling plant wherein a part of the fuel coming from some of the French, German, Belgian, Swiss, Dutch and Japanese nuclear reactors is reprocessed before being recycled in order to separate certain radioactive elements. The facility has been successively handled by the CEA (1962-1978), Cogema (1978-2006), and AREVA NC (since 2006). La Hague facility is composed of 3 production units: The UP2-400 production unit started to be operated in 1966 for the reprocessing of UNGG metal fuel. In 1976, following the dropout of the graphite-gas technology by EDF, an HAO workshopmore » to reprocess the fuel from the light water reactors is affiliated and then stopped in 2003. - UP2-400 is partially stopped in 2002 and then definitely the 1 January 2004 and is being dismantled - UP2-800, with the same capacity than UP3, started to be operated in 1994 and is still in operation. And UP3 - UP3 was implemented in 1990 with an annual reprocessing capacity of 800 tons of fuel and is still in operation The combined licensed capacity of UP2-800 and UP3 is 1,700 tons of used fuel. (authors)« less

An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

NASA Astrophysics Data System (ADS)

Aufiero, M.; Cammi, A.; Fiorina, C.; Leppänen, J.; Luzzi, L.; Ricotti, M. E.

2013-10-01

In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

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

Mcwilliams, A. J.

2015-09-08

This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniquesmore » through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.« less

Out-of-pile creep behavior of uranium carbide

NASA Technical Reports Server (NTRS)

Wright, T. R.; Seltzer, M. S.

1974-01-01

Compression creep tests were investigated on various UC-based fuel materials having a variation in both density and composition. Specimens were prepared by casting and by hot pressing. Steady-state creep rates were measured under vacuum at 1400 to 1800 C in the stress range 500-4000 psi.

10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

Code of Federal Regulations, 2012 CFR

2012-01-01

... and Related Waste Management Facilities F Appendix F to Part 50 Energy NUCLEAR REGULATORY COMMISSION... Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities 1. Public health... facilities for the temporary storage of highlevel radioactive wastes, may be located on privately owned...

10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

Code of Federal Regulations, 2013 CFR

2013-01-01

... and Related Waste Management Facilities F Appendix F to Part 50 Energy NUCLEAR REGULATORY COMMISSION... Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities 1. Public health... facilities for the temporary storage of highlevel radioactive wastes, may be located on privately owned...

10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

Code of Federal Regulations, 2014 CFR

2014-01-01

... and Related Waste Management Facilities F Appendix F to Part 50 Energy NUCLEAR REGULATORY COMMISSION... Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities 1. Public health... facilities for the temporary storage of highlevel radioactive wastes, may be located on privately owned...

Critical review of analytical techniques for safeguarding the thorium-uranium fuel cycle

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

Hakkila, E.A.

1978-10-01

Conventional analytical methods applicable to the determination of thorium, uranium, and plutonium in feed, product, and waste streams from reprocessing thorium-based nuclear reactor fuels are reviewed. Separations methods of interest for these analyses are discussed. Recommendations concerning the applicability of various techniques to reprocessing samples are included. 15 tables, 218 references.

Next-generation purex flowsheets with acetohydroxamic acid as complexant for FBR and thermal-fuel reprocessing

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

Kumar, Shekhar; Koganti, S.B.

2008-07-01

Acetohydroxamic acid (AHA) is a novel complexant for recycle of nuclear-fuel materials. It can be used in ordinary centrifugal extractors, eliminating the need for electro-redox equipment or complex maintenance requirements in a remotely maintained hot cell. In this work, the effect of AHA on Pu(IV) distribution ratios in 30% TBP system was quantified, modeled, and integrated in SIMPSEX code. Two sets of batch experiments involving macro Pu concentrations (conducted at IGCAR) and one high-Pu flowsheet (literature) were simulated for AHA based U-Pu separation. Based on the simulation and validation results, AHA based next-generation reprocessing flowsheets are proposed for co-processing basedmore » FBR and thermal-fuel reprocessing as well as evaporator-less macro-level Pu concentration process required for MOX fuel fabrication. Utilization of AHA results in significant simplification in plant design and simpler technology implementations with significant cost savings. (authors)« less

NUCLEAR MATERIAL ATTRACTIVENESS: AN ASSESSMENT OF MATERIAL FROM PHWR'S IN A CLOSED THORIUM FUEL CYCLE

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

Sleaford, B W; Collins, B A; Ebbinghaus, B B

2010-04-26

This paper examines the attractiveness of material mixtures containing special nuclear materials (SNM) associated with reprocessing and the thorium-based LWR fuel cycle. This paper expands upon the results from earlier studies that examined the attractiveness of SNM associated with the reprocessing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR. This study shows that {sup 233}U that is produced in thorium-based fuel cycles is very attractive for weapons use. Consistent with other studies, these results also show that all fuel cycles examined to date needmore » to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of 'attractiveness levels' that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities. The methodology and key findings will be presented.« less

Nuclear Material Attractiveness: An Assessment of Material from PHWR's in a Closed Thorium Fuel Cycle

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

Sleaford, Brad W.; Ebbinghaus, B. B.; Bradley, Keith S.

2010-06-11

This paper examines the attractiveness of material mixtures containing special nuclear materials (SNM) associated with reprocessing and the thorium-based LWR fuel cycle. This paper expands upon the results from earlier studies [ , ] that examined the attractiveness of SNM associated with the reprocessing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR. This study shows that 233U that is produced in thorium-based fuel cycles is very attractive for weapons use. Consistent with other studies, these results also show that all fuel cycles examined tomore » date need to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of "attractiveness levels" that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities [ ]. The methodology and key findings will be presented.« less

Disposal of spent fuel

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

Blomeke, J O; Ferguson, D E; Croff, A G

1978-01-01

Based on preliminary analyses, spent fuel assemblies are an acceptable form for waste disposal. The following studies appear necessary to bring our knowledge of spent fuel as a final disposal form to a level comparable with that of the solidified wastes from reprocessing: 1. A complete systems analysis is needed of spent fuel disposition from reactor discharge to final isolation in a repository. 2. Since it appears desirable to encase the spent fuel assembly in a metal canister, candidate materials for this container need to be studied. 3. It is highly likely that some ''filler'' material will be needed betweenmore » the fuel elements and the can. 4. Leachability, stability, and waste-rock interaction studies should be carried out on the fuels. The major disadvantages of spent fuel as a disposal form are the lower maximum heat loading, 60 kW/acre versus 150 kW/acre for high-level waste from a reprocessing plant; the greater long-term potential hazard due to the larger quantities of plutonium and uranium introduced into a repository; and the possibility of criticality in case the repository is breached. The major advantages are the lower cost and increased near-term safety resulting from eliminating reprocessing and the treatment and handling of the wastes therefrom.« less

Using eye movement desensitization and reprocessing to enhance treatment of couples.

PubMed

Protinsky, H; Sparks, J; Flemke, K

2001-04-01

Eye Movement Desensitization and Reprocessing (EMDR) as a clinical technique may enhance treatment effectiveness when applied within a couple therapy approach that is emotionally and experientially oriented. Clinical experience indicates that EMDR-based interventions are useful for accessing, activating, tolerating, and reprocessing the intense emotions that often fuel dysfunctional couple interactions. Using EMDR within conjoint sessions to reprocess negative emotions can amplify intimacy, increase connection, and subsequently lead to a change in problematic relationship patterns.

LIFE Materials: Thermomechanical Effects Volume 5 - Part I

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

Caro, M; DeMange, P; Marian, J

2009-05-07

Improved fuel performance is a key issue in the current Laser Inertial-Confinement Fusion-Fission Energy (LIFE) engine design. LIFE is a fusion-fission engine composed of a {approx}40-tons fuel blanket surrounding a pulsed fusion neutron source. Fusion neutrons get multiplied and moderated in a Beryllium blanket before penetrating the subcritical fission blanket. The fuel in the blanket is composed of millions of fuel pebbles, and can in principle be burned to over 99% FIMA without refueling or reprocessing. This report contains the following chapters: Chapter A: LIFE Requirements for Materials -- LIFE Fuel; Chapter B: Summary of Existing Knowledge; Chapter C: Identificationmore » of Gaps in Knowledge & Vulnerabilities; and Chapter D: Strategy and Future Work.« less

Evaluation of Tritium Content and Release from Pressurized Water Reactor Fuel Cladding

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

Robinson, Sharon M.; Chattin, Marc Rhea; Giaquinto, Joseph

2015-09-01

It is expected that tritium pretreatment will be required in future reprocessing plants to prevent the release of tritium to the environment (except for long-cooled fuels). To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified. Tritium in light water reactor (LWR) fuel is dispersed between the fuel matrix and the fuel cladding, and some tritium may be in the plenum, probably as tritium labelled water (THO) or T 2O. In a standard processing flowsheet, tritium management would bemore » accomplished by treatment of liquid streams within the plant. Pretreating the fuel prior to dissolution to release the tritium into a single off-gas stream could simplify tritium management, so the removal of tritium in the liquid streams throughout the plant may not be required. The fraction of tritium remaining in the cladding may be reduced as a result of tritium pretreatment. Since Zircaloy® cladding makes up roughly 25% by mass of UNF in the United States, processes are being considered to reduce the volume of reprocessing waste for Zircaloy® clad fuel by recovering the zirconium from the cladding for reuse. These recycle processes could release the tritium in the cladding. For Zircaloy-clad fuels from light water reactors, the tritium produced from ternary fission and other sources is expected to be divided between the fuel, where it is generated, and the cladding. It has been previously documented that a fraction of the tritium produced in uranium oxide fuel from LWRs can migrate and become trapped in the cladding. Estimates of the percentage of tritium in the cladding typically range from 0–96%. There is relatively limited data on how the tritium content of the cladding varies with burnup and fuel history (temperature, power, etc.) and how pretreatment impacts its release. To gain a better understanding of how tritium in cladding will behave during processing, scoping tests are being performed to determine the tritium content in the cladding pre- and post-tritium pretreatment. Samples of Surry-2 and H.B. Robinson pressurized water reactor cladding were heated to 1100–1200°C to oxidize the zirconium and release all of the tritium in the cladding sample. Cladding samples were also heated within the temperature range of 480–600ºC expected for standard air tritium pretreatment systems, and to a slightly higher temperature (700ºC) to determine the impact of tritium pretreatment on tritium release from the cladding. The tritium content of the Surry-2 and H.B. Robinson cladding was measured to be ~234 and ~500 µCi/g, respectively. Heating the Surry-2 cladding at 500°C for 24 h removed ~0.2% of the tritium from the cladding, and heating at 700°C for 24 h removed ~9%. Heating the H.B. Robinson cladding at 700°C for 24 h removed ~11% of the tritium. When samples of the Surry-2 and H.B. Robinson claddings were heated at 700°C for 96 h, essentially all of the tritium in the cladding was removed. However, only ~3% of the tritium was removed when a sample of Surry-2 cladding was heated at 600°C for 96 h. These data indicate that the amount of tritium released from tritium pretreatment systems will be dependent on both the operating temperature and length of time in the system. Under certain conditions, a significant fraction of the tritium could remain bound in the cladding and would need to be considered in operations involving cladding recycle.« less

Estimation of carbon 14 inventory in hull and end-piece wastes from Japanese commercial reprocessing operation

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

Tomofumi Sakuragi; Hiromi Tanabe; Emiko Hirose

2013-07-01

Hull and end-piece wastes generated from reprocessing plant operations are expected to be disposed of in a deep underground repository as Group 2 TRU wastes under the Japanese classification system. The activated metals that compose the spent fuel assemblies such as Zircaloy claddings and stainless steel nozzles are mixed and compressed after fuel dissolution, and then stuffed into stainless steel canisters. Carbon 14 is a typical activated product in the hulls and end-pieces and is mainly generated by the {sup 14}N(n,p){sup 14}C reaction. In the previous safety assessment of the TRU waste in Japan, the radionuclides inventory was calculated bymore » ORIGEN-2 code. Some conservative assumptions and preliminary estimates were used in this calculation. For example, total radionuclides generated from a single type of fuel assembly (45 GWd/tU for a PWR unit), and the thickness of the Zircaloy oxide film on the hulls (80 μm) were both overestimated. The second assumption in particular has a large effect on exposure dose evaluation. Therefore, it is essential to have a realistic source term evaluation regarding such items as the C-14 inventory and its distribution to waste parts. In the present study, a C-14 inventory of the hull and end-piece wastes from the operation of a commercial reprocessing plant in Japan corresponding to 32,000 tU (16,000 tU in each BWR and PWR) was calculated. Analysis using individual irradiation conditions and fuel characteristics was conducted on 6 types of fuel assemblies for BWRs and 12 types for PWRs (4 pile types x 3 burnup limits). The oxide film thickness data for each fuel type cladding were obtained from the published literature. Activation calculations were performed by using ORIGEN-2 code. For the amount of spent assembly and other waste characteristics, representative values were assumed based on the published literature. As a preliminary experiment, C-14 in irradiated BWR claddings was measured and found to be consistent with the calculated activation. The total C-14 inventory was estimated as 4.46x10{sup 14} Bq, consisting of 2.58x10{sup 14} Bq for BWRs and 1.87x10{sup 14} Bq for PWRs, and is consistent with the safety assessment of 4.4x10{sup 14} Bq. However, the distribution of the C-14 inventory to hull oxide, which was estimated under the assumption of instantaneous radionuclide release in the safety assessment, decreased from 5.72x10{sup 13} Bq (13% of the total) in the previous assessment to 1.30x10{sup 13} Bq (2.9% of the total; consisting of 1.48x10{sup 12} for BWRs and 1.15x10{sup 13} for PWRs). In other words, the exposure dose peak is reduced to approximate 25% of its previous value due to the use of detailed oxide film data that the BWR cladding has a thin oxide film. Other instantaneous release components for C-14 such as the fuel residual were negligible. (authors)« less

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

Coble, Jamie; Orton, Christopher; Schwantes, Jon

Abstract—The Multi-Isotope Process (MIP) Monitor provides an efficient approach to monitoring the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of reprocessing streams in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor), initial enrichment, burn up, and cooling time. Simulated gamma spectra were used to develop and test threemore » fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type. Locally weighted PLS models were fitted on-the-fly to estimate continuous fuel characteristics. Burn up was predicted within 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment within approximately 2% RMSPE. This automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters and material diversions.« less

Japan’s Nuclear Future: Policy Debate, Prospects, and U.S. Interests

DTIC Science & Technology

2008-05-09

raised in particular over the construction of an industrial- scale reprocessing facility in Japan,. Additionally, fast breeder reactors also produce more...Nuclear Fuel Cycle Engineering Laboratories. 10 A fast breeder reactor is a fast neutron reactor that produces more plutonium than it consumes, which can...Japan Nuclear Fuel Limited (JNFL) has built and is currently running active testing on a large - scale commercial reprocessing plant at Rokkasho-mura

Method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions

DOEpatents

Horwitz, E. Philip; Delphin, Walter H.

1979-07-24

A method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions containing these and other values by contacting the waste solution with an extractant of tricaprylmethylammonium nitrate in an inert hydrocarbon diluent which extracts the palladium and technetium values from the waste solution. The palladium and technetium values are recovered from the extractant and from any other coextracted values with a strong nitric acid strip solution.

Thermomechanics of candidate coatings for advanced gas reactor fuels

NASA Astrophysics Data System (ADS)

Nosek, A.; Conzen, J.; Doescher, H.; Martin, C.; Blanchard, J.

2007-09-01

Candidate fuel/coating combinations for an advanced, coated-fuel particle for a gas-cooled fast reactor (GFR) have been evaluated. These all-ceramic fuel forms consist of a fuel kernel made of UC or UN, surrounded with two shells (a buffer and a coating) made of TiC, SiC, ZrC, TiN, or ZrN. These carbides and nitrides are analyzed with finite element models to determine the stresses produced in the micro fuel particles from differential thermal expansion, fission gas release, swelling, and creep during particle fabrication and reactor operation. This study will help determine the feasibility of different fuel and coating combinations and identify the critical loads. The analysis shows that differential thermal expansion of the fuel and coating dictate the amount of stress for changing temperatures (such as during fabrication), and that the coating creep is able to mitigate an otherwise overwhelming amount of stress from fuel swelling. Because fracture is a likely mode of failure, a fracture mechanics study is also included to identify the relative likelihood of catastrophic fracture of the coating and resulting gas release. Overall, the analysis predicts that UN/ZrC is the best thermomechanical fuel/coating combination for mitigating the stress within the new fuel particle, but UN/TiN and UN/ZrN could also be strong candidates if their unknown creep rates are sufficiently large.

On the influence of the americium isotopic vector on the cooling time of minor actinides bearing blankets in fast reactors

NASA Astrophysics Data System (ADS)

Kooyman, Timothée; Buiron, Laurent; Rimpault, Gerald

2018-05-01

In the heterogeneous minor actinides transmutation approach, the nuclei to be transmuted are loaded in dedicated targets often located at the core periphery, so that long-lived heavy nuclides are turned into shorter-lived fission products by fission. To compensate for low flux level at the core periphery, the minor actinides content in the targets is set relatively high (around 20 at.%), which has a negative impact on the reprocessing of the targets due to their important decay heat level. After a complete analysis of the main contributors to the heat load of the irradiated targets, it is shown here that the choice of the reprocessing order of the various feeds of americium from the fuel cycle depends on the actual limit for fuel reprocessing. If reprocessing of hot targets is possible, it is more interesting to reprocess first the americium feed with a high 243Am content in order to limit the total cooling time of the targets, while if reprocessing of targets is limited by their decay heat, it is more interesting to wait for an increase in the 241Am content before loading the americium in the core. An optimization of the reprocessing order appears to lead to a decrease of the total cooling time by 15 years compared to a situation where all the americium feeds are mixed together when two feeds from SFR are considered with a high reprocessing limit.

Nuclear fuels - Present and future

NASA Astrophysics Data System (ADS)

Olander, D.

2009-06-01

The important developments in nuclear fuels and their problems are reviewed and compared with the status of present light-water reactor fuels. The limitations of LWR fuels are reviewed with respect to important recent concerns, namely provision of outlet coolant temperatures high enough for use in H 2 production, destruction of plutonium to eliminate proliferation concerns, and burning of the minor actinides to reduce the waste repository heat load and long-term radiation hazard. In addition to current oxide-based fuel rod designs, the hydride fuel with liquid-metal thermal bonding of the fuel-cladding gap is covered. Finally, two of the most promising Generation IV reactor concepts, the very high temperature reactor and the sodium fast reactor, and the accompanying reprocessing technologies, aqueous-based UREX+1a and pyrometallurgical, are summarized. In all of the topics covered, the thermodynamics involved in the fuel's behavior under irradiation and in the reprocessing schemes are emphasized.

Corrosion study of a highly durable electrolyzer based on cold crucible technique for pyrochemical reprocessing of spent nuclear oxide fuel

NASA Astrophysics Data System (ADS)

Takeuchi, M.; Arai, Y.; Kase, T.; Nakajima, Y.

2013-01-01

The application of the cold crucible technique to a pyrochemical electrolyzer used in the oxide-electrowinning method, which is a method for the pyrochemical reprocessing of spent nuclear oxide fuel, is proposed as a means for improving corrosion resistance. The electrolyzer suffers from a severe corrosion environment consisting of molten salt and corrosive gas. In this study, corrosion tests for several metals in molten 2CsCl-NaCl at 923 K with purging chlorine gas were conducted under controlled material temperature conditions. The results revealed that the corrosion rates of several materials were significantly decreased by the material cooling effect. In particular, Hastelloy C-22 showed excellent corrosion resistance with a corrosion rate of just under 0.01 mm/y in both molten salt and vapor phases by controlling the material surface at 473 K. Finally, an engineering-scale crucible composed of Hastelloy C-22 was manufactured to demonstrate the basic function of the cold crucible. The cold crucible induction melting system with the new concept Hastelloy crucible showed good compatibility with respect to its heating and cooling performances.

Variants of Regenerated Fissile Materials Usage in Thermal Reactors as the First Stage of Fuel Cycle Closing

NASA Astrophysics Data System (ADS)

Andrianova, E. A.; Tsibul'skiy, V. F.

2017-12-01

At present, 240 000 t of spent nuclear fuel (SF) has been accumulated in the world. Its long-term storage should meet safety conditions and requires noticeable finances, which grow every year. Obviously, this situation cannot exist for a long time; in the end, it will require a final decision. At present, several variants of solution of the problem of SF management are considered. Since most of the operating reactors and those under construction are thermal reactors, it is reasonable to assume that the structure of the nuclear power industry in the near and medium-term future will be unchanged, and it will be necessary to utilize plutonium in thermal reactors. In this study, different strategies of SF management are compared: open fuel cycle with long-term SF storage, closed fuel cycle with MOX fuel usage in thermal reactors and subsequent long-term storage of SF from MOX fuel, and closed fuel cycle in thermal reactors with heterogeneous fuel arrangement. The concept of heterogeneous fuel arrangement is considered in detail. While in the case of traditional fuel it is necessary to reprocess the whole amount of spent fuel, in the case of heterogeneous arrangement, it is possible to separate plutonium and 238U in different fuel rods. In this case, it is possible to achieve nearly complete burning of fissile isotopes of plutonium in fuel rods loaded with plutonium. These fuel rods with burned plutonium can be buried after cooling without reprocessing. They would contain just several percent of initially loaded plutonium, mainly even isotopes. Fuel rods with 238U alone should be reprocessed in the usual way.

Advantages of Production of New Fissionable Nuclides for the Nuclear Power Industry in Hybrid Fusion-Fission Reactors

NASA Astrophysics Data System (ADS)

Tsibulskiy, V. F.; Andrianova, E. A.; Davidenko, V. D.; Rodionova, E. V.; Tsibulskiy, S. V.

2017-12-01

A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium-plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.

Impact of Reprocessed Uranium Management on the Homogeneous Recycling of Transuranics in PWRs

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

Youinou, Gilles J.

This article presents the results of a neutronics analysis related to the homogeneous recycling of transuranics (TRU) in PWRs with a MOX fuel using enriched uranium instead of depleted uranium. It also addresses an often, if not always, overlooked aspect related to the recycling of TRU in PWRs, namely the use of reprocessed uranium. From a neutronics point of view, it is possible to multi-recycle the entirety of the plutonium with or without neptunium and americium in a PWR fleet using MOX-EU fuel in between one third and two thirds of the fleet. Recycling neptunium and americium with plutonium significantlymore » decreases the decay heat of the waste stream between 100 to 1,000 years compared to those of an open fuel cycle or when only plutonium is recycled. The uranium present in MOX-EU used fuel still contains a significant amount of 235uranium and recycling it makes a major difference on the natural uranium needs. For example, a PWR fleet recycling its plutonium, neptunium and americium in MOXEU needs 28 percent more natural uranium than a reference UO 2 open cycle fleet generating the same energy if the reprocessed uranium is not recycled and 19 percent less if the reprocessed uranium is recycled back in the reactors, i.e. a 47 percent difference.« less

Impact of Reprocessed Uranium Management on the Homogeneous Recycling of Transuranics in PWRs

DOE PAGES

Youinou, Gilles J.

2017-05-04

This article presents the results of a neutronics analysis related to the homogeneous recycling of transuranics (TRU) in PWRs with a MOX fuel using enriched uranium instead of depleted uranium. It also addresses an often, if not always, overlooked aspect related to the recycling of TRU in PWRs, namely the use of reprocessed uranium. From a neutronics point of view, it is possible to multi-recycle the entirety of the plutonium with or without neptunium and americium in a PWR fleet using MOX-EU fuel in between one third and two thirds of the fleet. Recycling neptunium and americium with plutonium significantlymore » decreases the decay heat of the waste stream between 100 to 1,000 years compared to those of an open fuel cycle or when only plutonium is recycled. The uranium present in MOX-EU used fuel still contains a significant amount of 235uranium and recycling it makes a major difference on the natural uranium needs. For example, a PWR fleet recycling its plutonium, neptunium and americium in MOXEU needs 28 percent more natural uranium than a reference UO 2 open cycle fleet generating the same energy if the reprocessed uranium is not recycled and 19 percent less if the reprocessed uranium is recycled back in the reactors, i.e. a 47 percent difference.« less

Metallic conductance at the interface of tri-color titanate superlattices

NASA Astrophysics Data System (ADS)

Kareev, M.; Cao, Yanwei; Liu, Xiaoran; Middey, S.; Meyers, D.; Chakhalian, J.

2013-12-01

Ultra-thin tri-color (tri-layer) titanate superlattices ([3 u.c. LaTiO3/2 u.c. SrTiO3/3 u.c. YTiO3], u.c. = unit cells) were grown in a layer-by-layer way on single crystal TbScO3 (110) substrates by pulsed laser deposition. High sample quality and electronic structure were characterized by the combination of in-situ photoelectron and ex-situ structure and surface morphology probes. Temperature-dependent sheet resistance indicates the presence of metallic interfaces in both [3 u.c. LaTiO3/2 u.c. SrTiO3] bi-layers and all the tri-color structures, whereas a [3 u.c. YTiO3/2 u.c. SrTiO3] bi-layer shows insulating behavior. Considering that in the bulk YTiO3 is ferromagnetic below 30 K, the tri-color titanate superlattices provide an opportunity to induce tunable spin-polarization into the two-dimensional electron gas with Mott carriers.

Evaluation of Non-Oxide Fuel for Fission-based Nuclear Reactors on Spacecraft

DTIC Science & Technology

smaller and potentially lighter core, whichis a significant advantage. The results of this study indicate that use of both UC and UN may result in significant weight savings due tohigher uranium loading density....The goal of this project was to study the performance of atypical uranium-based fuels in a nuclear reactor capable of producing 1 megawattof thermal...UN), or uranium carbide (UC) and compared their performance to uranium oxide (UO2) which is thefuel form used in the vast majority of commercial

Reprocessing and Recycling of Highly Cross-Linked Ion-Conducting Networks through Transalkylation Exchanges of C-N Bonds.

PubMed

Obadia, Mona M; Mudraboyina, Bhanu P; Serghei, Anatoli; Montarnal, Damien; Drockenmuller, Eric

2015-05-13

Exploiting exchangeable covalent bonds as dynamic cross-links recently afforded a new class of polymer materials coined as vitrimers. These permanent networks are insoluble and infusible, but the network topology can be reshuffled at high temperatures, thus enabling glasslike plastic deformation and reprocessing without depolymerization. We disclose herein the development of functional and high-value ion-conducting vitrimers that take inspiration from poly(ionic liquid)s. Tunable networks with high ionic content are obtained by the solvent- and catalyst-free polyaddition of an α-azide-ω-alkyne monomer and simultaneous alkylation of the resulting poly(1,2,3-triazole)s with a series of difunctional cross-linking agents. Temperature-induced transalkylation exchanges of C-N bonds between 1,2,3-triazolium cross-links and halide-functionalized dangling chains enable recycling and reprocessing of these highly cross-linked permanent networks. They can also be recycled by depolymerization with specific solvents able to displace the transalkylation equilibrium, and they display a great potential for applications that require solid electrolytes with excellent mechanical performances and facile processing such as supercapacitors, batteries, fuel cells, and separation membranes.

Mars Mission Analysis Trades Based on Legacy and Future Nuclear Propulsion Options

NASA Astrophysics Data System (ADS)

Joyner, Russell; Lentati, Andrea; Cichon, Jaclyn

2007-01-01

The purpose of this paper is to discuss the results of mission-based system trades when using a nuclear thermal propulsion (NTP) system for Solar System exploration. The results are based on comparing reactor designs that use a ceramic-metallic (CERMET), graphite matrix, graphite composite matrix, or carbide matrix fuel element designs. The composite graphite matrix and CERMET designs have been examined for providing power as well as propulsion. Approaches to the design of the NTP to be discussed will include an examination of graphite, composite, carbide, and CERMET core designs and the attributes of each in regards to performance and power generation capability. The focus is on NTP approaches based on tested fuel materials within a prismatic fuel form per the Argonne National Laboratory testing and the ROVER/NERVA program. NTP concepts have been examined for several years at Pratt & Whitney Rocketdyne for use as the primary propulsion for human missions beyond earth. Recently, an approach was taken to examine the design trades between specific NTP concepts; NERVA-based (UC)C-Graphite, (UC,ZrC)C-Composite, (U,Zr)C-Solid Carbide and UO2-W CERMET. Using Pratt & Whitney Rocketdyne's multidisciplinary design analysis capability, a detailed mission and vehicle model has been used to examine how several of these NTP designs impact a human Mars mission. Trends for the propulsion system mass as a function of power level (i.e. thrust size) for the graphite-carbide and CERMET designs were established and correlated against data created over the past forty years. These were used for the mission trade study. The resulting mission trades presented in this paper used a comprehensive modeling approach that captures the mission, vehicle subsystems, and NTP sizing.

SOUTH ELEVATION OF IRRADIATED FUEL STORAGE FACILITY LOCATED IN FUEL ...

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

SOUTH ELEVATION OF IRRADIATED FUEL STORAGE FACILITY LOCATED IN FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORTH. INL PHOTO NUMBER HD-54-15-2. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

NORTH ELEVATION OF IRRADIATED FUEL STORAGE FACILITY LOCATED IN FUEL ...

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

NORTH ELEVATION OF IRRADIATED FUEL STORAGE FACILITY LOCATED IN FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING SOUTH. INL PHOTO NUMBER HD-54-16-1. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

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

Jolly, Brian C.; Lindemer, Terrence; Terrani, Kurt A.

In support of fully ceramic matrix (FCM) fuel development, coating development work has begun at the Oak Ridge National Laboratory (ORNL) to produce tri-isotropic (TRISO) coated fuel particles with UN kernels. The nitride kernels are used to increase heavy metal density in these SiC-matrix fuel pellets with details described elsewhere. The advanced gas reactor (AGR) program at ORNL used fluidized bed chemical vapor deposition (FBCVD) techniques for TRISO coating of UCO (two phase mixture of UO 2 and UC x) kernels. Similar techniques were employed for coating of the UN kernels, however significant changes in processing conditions were required tomore » maintain acceptable coating properties due to physical property and dimensional differences between the UCO and UN kernels.« less

Evaluation of nuclear fuel reprocessing strategies. 2. LWR fuel storage, recycle economics and plutonium logistics

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

Prince, B.E.; Hadley, S.W.

1983-10-27

This is the second of a two-part report intended as a critical review of certain issues involved with closing the Light Water Reactor (LWR) fuel cycle and establishing the basis for future transition to commercial breeder applications. The report is divided into four main sections consisting of (1) a review of the status of the LWR spent fuel management and storage problem; (2) an analysis of the economic incentives for instituting reprocessing and recycle in LWRs; (3) an analysis of the time-dependent aspects of plutonium economic value particularly as related to the LWR-breeder transition; and (4) an analysis of themore » time-dependent aspects of plutonium requirements and supply relative to this transition.« less

Extending Spent Fuel Storage until Transport for Reprocessing or Disposal

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

Carlsen, Brett; Chiguer, Mustapha; Grahn, Per

Spent fuel (SF) must be stored until an end point such as reprocessing or geologic disposal is imple-mented. Selection and implementation of an end point for SF depends upon future funding, legisla-tion, licensing and other factors that cannot be predicted with certainty. Past presumptions related to the availability of an end point have often been wrong and resulted in missed opportunities for properly informing spent fuel management policies and strategies. For example, dry cask storage systems were originally conceived to free up needed space in reactor spent fuel pools and also to provide SFS of up to 20 years untilmore » reprocessing and/or deep geological disposal became available. Hundreds of dry cask storage systems are now employed throughout the world and will be relied upon well beyond the originally envisioned design life. Given present and projected rates for the use of nuclear power coupled with projections for SF repro-cessing and disposal capacities, one concludes that SF storage will be prolonged, potentially for several decades. The US Nuclear Regulatory Commission has recently considered 300 years of storage to be appropriate for the characterization and prediction of ageing effects and ageing management issues associated with extending SF storage and subsequent transport. This paper encourages addressing the uncertainty associated with the duration of SF storage by de-sign – rather than by default. It suggests ways that this uncertainty may be considered in design, li-censing, policy, and strategy decisions and proposes a framework for safely extending spent fuel storage until SF can be transported for reprocessing or disposal – regardless of how long that may be. The paper however is not intended to either encourage or facilitate needlessly extending spent fuel storage durations. Its intent is to ensure a design and safety basis with sufficient margin to accommodate the full range of potential future scenarios. Although the focus is primarily on storage of SF from commercial operation, the principles described are equally applicable to SF from research and production reactors as well as high-level radioactive waste.« less

Method for reprocessing and separating spent nuclear fuels. [Patent application

DOEpatents

Krikorian, O.H.; Grens, J.Z.; Parrish, W.H. Sr.

1982-01-19

Spent nuclear fuels, including actinide fuels, volatile and nonvolatile fission products, are reprocessed and separated in a molten metal solvent housed in a separation vessel made of a carbon-containing material. A first catalyst, which promotes the solubility and permeability of carbon in the metal solvent, is included. By increasing the solubility and permeability of the carbon in the solvent, the rate at which actinide oxides are reduced (carbothermic reduction) is greatly increased. A second catalyst, included to increase the affinity for nitrogen in the metal solvent, is added to increase the rate at which actinide nitrides form after carbothermic reduction is complete.

Container for reprocessing and permanent storage of spent nuclear fuel assemblies

DOEpatents

Forsberg, Charles W.

1992-01-01

A single canister process container for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining their integrity at temperature necessary to oxide the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container.

Iodine Pathways and Off-Gas Stream Characteristics for Aqueous Reprocessing Plants – A Literature Survey and Assessment

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

R. T. Jubin; D. M. Strachan; N. R. Soelberg

2013-09-01

Used nuclear fuel is currently being reprocessed in only a few countries, notably France, England, Japan, and Russia. The need to control emissions of the gaseous radionuclides to the air during nuclear fuel reprocessing has already been reported for the entire plant. But since the gaseous radionuclides can partition to various different reprocessing off-gas streams, for example, from the head end, dissolver, vessel, cell, and melter, an understanding of each of these streams is critical. These off-gas streams have different flow rates and compositions and could have different gaseous radionuclide control requirements, depending on how the gaseous radionuclides partition. Thismore » report reviews the available literature to summarize specific engineering data on the flow rates, forms of the volatile radionuclides in off-gas streams, distributions of these radionuclides in these streams, and temperatures of these streams. This document contains an extensive bibliography of the information contained in the open literature.« less

EAST ELEVATION OF HIGH BAY ADDITION OF FUEL STORAGE BUILDING ...

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

EAST ELEVATION OF HIGH BAY ADDITION OF FUEL STORAGE BUILDING (CPP-603). INL DRAWING NUMBER 200-0603-00-706-051286. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

The evaluation of the pyrochemistry for the treatment of Gen IV nuclear fuels Inert matrix chlorination studies in the gas phase or molten chloride salts

NASA Astrophysics Data System (ADS)

Bourg, S.; Péron, F.; Lacquement, J.

2007-01-01

The structure of the fuels for the future Gen IV nuclear reactors will be totally different from those of PWR, especially for the GFR concept including a closed cycle. In these reactors, fissile materials (carbides or nitrides of actinides) should be surrounded by an inert matrix. In order to build a reprocessing process scheme, the behavior of the potential inert matrices (silicon carbide, titanium nitride, and zirconium carbide and nitride) was studied by hydro- and pyrometallurgy. This paper deals with the chlorination results at high temperature by pyrometallurgy. For the first time, the reactivity of the matrix towards chlorine gas was assessed in the gas phase. TiN, ZrN and ZrC are very reactive from 400 °C whereas it is necessary to be over 900 °C for SiC to be as fast. In molten chloride melts, the bubbling of chlorine gas is less efficient than in gas phase but it is possible to attack the matrices. Electrochemical methods were also used to dissolve the refractory materials, leading to promising results with TiN, ZrN and ZrC. The massive SiC samples used were not conductive enough to be studied and in this case specific SiC-coated carbon electrodes were used. The key point of these studies was to find a method to separate the matrix compounds from the fissile material in order to link the head to the core of the process (electrochemical separation or liquid-liquid reductive extraction in the case of a pyrochemical reprocessing).

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Picklesimer, M.L.; Thurber, W.C.

1961-01-01

A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

CONSTRUCTION PROGRESS PHOTO SHOWING WEST STORAGE BASIN AT FUEL STORAGE ...

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

CONSTRUCTION PROGRESS PHOTO SHOWING WEST STORAGE BASIN AT FUEL STORAGE BUILDING (CPP-603). INL PHOTO NUMBER NRTS-51-689. Unknown Photographer, 1950 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

The benefits of a fast reactor closed fuel cycle in the UK

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

Gregg, R.; Hesketh, K.

2013-07-01

The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size,more » so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since the fission product will primarily be a function of nuclear energy generated). However, by reprocessing spent fuel, it is possible to immobilise the fission product in a more suitable waste form that has far more superior in-repository performance. (authors)« less

FACILITY LAYOUT OF FUEL STORAGE BUILDING (CPP603) SHOWING STORAGE BASINS, ...

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

FACILITY LAYOUT OF FUEL STORAGE BUILDING (CPP-603) SHOWING STORAGE BASINS, FUEL ELEMENT CUTTING FACILITY, AND DRY GRAPHITE STORAGE FACILITY. INL DRAWING NUMBER 200-0603-00-030-056329. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

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

Bruffey, S. H.; Spencer, B. B.; Strachan, D. M.

Four radionuclides have been identified as being sufficiently volatile in the reprocessing of nuclear fuel that their gaseous release needs to be controlled to meet regulatory requirements (Jubin et al. 2011, 2012). These radionuclides are 3H, 14C, 85Kr, and 129I. Of these, 129I has the longest half-life and potentially high biological impact. Accordingly, control of the release of 129I is most critical with respect to the regulations for the release of radioactive material in stack emissions. It is estimated that current EPA regulations (EPA 2010) would require any reprocessing plant in the United States to limit 129I release to lessmore » than 0.05 Ci/MTIHM for a typical fuel burnup of 55 gigawatt days per metric tonne (GWd/t) (Jubin 2011). The study of inorganic iodide in off-gas systems has been almost exclusively limited to I2 and the focus of organic iodide studies has been CH3I. In this document, we provide the results of an examination of publically available literature that is relevant to the presence and sources of both inorganic and organic iodine-bearing species in reprocessing plants. We especially focus on those that have the potential to be poorly sequestered with traditional capture methodologies. Based on the results of the literature survey and some limited thermodynamic modeling, the inorganic iodine species hypoiodous acid (HOI) and iodine monochloride (ICl) were identified as potentially low-sorbing iodine species that could present in off-gas systems. Organic species of interest included both short chain alkyl iodides such as methyl iodide (CH3I) and longer alkyl iodides up to iodododecane (C10H21I). It was found that fuel dissolution may provide conditions conducive to HOI formation and has been shown to result in volatile long-chain alkyl iodides, though these may not volatilize until later in the reprocessing sequence. Solvent extraction processes were found to be significant sources of various organic iodine-bearing species; formation of these was facilitated by the presence of radiolytic decomposition products resulting from radiolysis of tri-n-butyl phosphate and dodecane. Primarily inorganic iodine compounds were expected from waste management processes, including chlorinated species such as ICl. Critical knowledge gaps that must still be addressed include confirmation of the existence and quantification of low-sorbing species in the off-gas of reprocessing facilities. The contributions from penetrating forms of iodine to the plant DF are largely unknown and highly dependent on the magnitude of their presence. These species are likely to be more difficult to remove and it is likely that their sequestration could be improved through the use of different sorbents, through design modifications of the off-gas capture system, or through chemical conversion prior to iodine abatement that would produce more easily captured forms.« less

75 FR 45167 - Notice of Public Workshop on a Potential Rulemaking for Spent Nuclear Fuel Reprocessing Facilities

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-02

... more sophisticated reprocessing technology. During the Bush Administration, the Global Nuclear Energy... Associated with the Global Nuclear Energy Partnership,'' dated June 27, 2007 (ADAMS ML071800084), directed... on some Global Nuclear Energy Partnership (GNEP) initiatives had waned and it appeared appropriate to...

MicroRaman measurements for nuclear fuel reprocessing applications

DOE PAGES

Casella, Amanda; Lines, Amanda; Nelson, Gilbert; ...

2016-12-01

Treatment and reuse of used nuclear fuel is a key component in closing the nuclear fuel cycle. Solvent extraction reprocessing methods that have been developed contain various steps tailored to the separation of specific radionuclides, which are highly dependent upon solution properties. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. Our group has been investigating the use of optical spectroscopy for the on-line monitoring of actinides, lanthanides, and acid strength within fuel reprocessing streams. This paper willmore » focus on the development and application of a new MicroRaman probe for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid in solutions relevant to used nuclear fuel reprocessing. Previous research has successfully demonstrated the applicability on the macroscopic scale, using sample probes requiring larger solution volumes. In an effort to minimize waste and reduce dose to personnel, we have modified this technique to allow measurement at the microfluidic scale using a Raman microprobe. Under the current sampling environment, Raman samples typically require upwards of 10 mL and larger. Using the new sampling system, we can sample volumes at 10 μL or less, which is a scale reduction of over 1,000 fold in sample size. Finally, this paper will summarize our current work in this area including: comparisons between the macroscopic and microscopic probes for detection limits, optimized channel focusing, and application in a flow cell with varying levels of HNO 3, and UO 2(NO 3) 2.« less

Apparatus and method for reprocessing and separating spent nuclear fuels. [Patent application

DOEpatents

Krikorian, O.H.; Grens, J.Z.; Parrish, W.H. Sr.; Coops, M.S.

1982-01-19

A method and apparatus for separating and reprocessing spent nuclear fuels includes a separation vessel housing a molten metal solvent in a reaction region, a reflux region positioned above and adjacent to the reaction region, and a porous filter member defining the bottom of the separation vessel in a supporting relationship with the metal solvent. Spent fuels are added to the metal solvent. A nonoxidizing nitrogen-containing gas is introduced into the separation vessel, forming solid actinide nitrides in the metal solvent from actinide fuels, while leaving other fission products in solution. A pressure of about 1.1 to 1.2 atm is applied in the reflux region, forcing the molten metal solvent and soluble fission products out of the vessel, while leaving the solid actinide nitrides in the separation vessel.

Conceptual designs of NDA instruments for the NRTA system at the Rokkasho Reprocessing Plant

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

Li, T.K.; Klosterbuer, S.F.; Menlove, H.O.

The authors are studying conceptual designs of selected nondestructive assay (NDA) instruments for the near-real-time accounting system at the rokkasho Reprocessing Plant (RRP) of Japan Nuclear Fuel Limited (JNFL). The JNFL RRP is a large-scale commercial reprocessing facility for spent fuel from boiling-water and pressurized-water reactors. The facility comprises two major components: the main process area to separate and produce purified plutonium nitrate and uranyl nitrate from irradiated reactor spent fuels, and the co-denitration process area to combine and convert the plutonium nitrate and uranyl nitrate into mixed oxide (MOX). The selected NDA instruments for conceptual design studies are themore » MOX-product canister counter, holdup measurement systems for calcination and reduction furnaces and for blenders in the co-denitration process, the isotope dilution gamma-ray spectrometer for the spent fuel dissolver solution, and unattended verification systems. For more effective and practical safeguards and material control and accounting at RRP, the authors are also studying the conceptual design for the UO{sub 3} large-barrel counter. This paper discusses the state-of-the-art NDA conceptual design and research and development activities for the above instruments.« less

[Expression and clinical significance of 5hmC in bladder urothelial carcinoma].

PubMed

Li, Jie; Xu, Yuqiao; Zhang, Zhiwen; Zhang, Ming; Zhang, Zhekai; Zhang, Feng; Li, Qing

2016-02-01

To investigate the expression of 5-hydroxymethylcytosine (5hmC) in bladder urothelial carcinoma (UC) and its clinical significance. The expression of 5hmC in 21 cases of UC tissues and pericarcinous urinary tract epithelium was detected by immunohistochemical staining. Then the expression of 5hmC in the surgical resection of UC tissues in 92 cases was also surveyed. Non parametric U Mann-Whitney test was used to analyze the correlation between 5hmC expression and clinical data. Single factor survival analysis was performed by Kaplan-Meier test. The expression of 5hmC in normal urinary tract epithelium and UC tissues was significantly different, but there was no significant difference in the expression of 5hmC between low and high grades of UC tissues as well as between different TNM grades. Kaplan-Meier single factor survival analysis showed that there was no significant correlation between the 5hmC expression level and the survival rate or the recurrence-free survival of UC patients. The expression level of 5hmC in UC tissues is significantly lower than that in pericarcinous urinary tract epithelium. There is no correlation between the 5hmC expression and the progression, prognosis and recurrence of UC.

CONSTRUCTION VIEW FUEL STORAGE BUILDING (CPP603) LOOKING EAST SHOWING ASBESTOS ...

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

CONSTRUCTION VIEW FUEL STORAGE BUILDING (CPP-603) LOOKING EAST SHOWING ASBESTOS SIDING. INL PHOTO NUMBER NRTS-51-1543. Unknown Photographer, 2/28/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO SHOWING FUEL STORAGE BUILDING (CPP603) LOOKING NORTHWEST. ...

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

CONSTRUCTION PROGRESS PHOTO SHOWING FUEL STORAGE BUILDING (CPP-603) LOOKING NORTHWEST. INL PHOTO NUMBER NRTS-50-895. Unknown Photographer, 10/30/1950 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

BUILDING PLANS OF FUEL STORAGE BUILDING (CPP603). INL DRAWING NUMBER ...

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

BUILDING PLANS OF FUEL STORAGE BUILDING (CPP-603). INL DRAWING NUMBER 200-0603-61-299-103029. ALTERNATE ID NUMBER 542-31-B-21. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Container for reprocessing and permanent storage of spent nuclear fuel assemblies

DOEpatents

Forsberg, C.W.

1992-03-24

A single canister process container is described for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining its integrity at a temperature necessary to oxidize the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container. 10 figs.

Diethylene-triamine-penta-acetate administration protocol for radiological emergency medicine in nuclear fuel reprocessing plants.

PubMed

Jin, Yutaka

2008-01-01

Inhalation therapy of diethylene-triamine-penta-acetate (DTPA) should be initiated immediately to workers who have significant incorporation of plutonium, americium or curium in the nuclear fuel reprocessing plant. A newly designed electric mesh nebulizer is a small battery-operated passive vibrating mesh device, in which vibrations in an ultrasonic horn are used to force drug solution through a mesh of micron-sized holes. This nebulizer enables DTPA administration at an early stage in the event of a radiation emergency from contamination from the above radioactive metals.

Survey of simulation methods for modeling pulsed sieve-plate extraction columns

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

Burkhart, L.

1979-03-01

The report first considers briefly the use of liquid-liquid extraction in nuclear fuel reprocessing and then describes the operation of the pulse column. Currently available simulation models of the column are reviewed, and followed by an analysis of the information presently available from which the necessary parameters can be obtained for use in a model of the column. Finally, overall conclusions are given regarding the information needed to develop an accurate model of the column for materials accountability in fuel reprocessing plants. 156 references.

Multivariate analysis of gamma spectra to characterize used nuclear fuel

DOE PAGES

Coble, Jamie; Orton, Christopher; Schwantes, Jon

2017-01-17

The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

Multivariate analysis of gamma spectra to characterize used nuclear fuel

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

Coble, Jamie; Orton, Christopher; Schwantes, Jon

The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

Assessment for advanced fuel cycle options in CANDU

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

Morreale, A.C.; Luxat, J.C.; Friedlander, Y.

2013-07-01

The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a drivermore » fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.« less

Stability and adaptability of popcorn genotypes in the State of Rio de Janeiro, Brazil.

PubMed

Pena, G F; do Amaral, A T; Gonçalves, L S A; Candido, L S; Vittorazzi, C; Ribeiro, R M; Freitas, S P

2012-08-31

This study aimed to obtain estimates of stability and adaptability of phase launched materials and materials recommended in the country, for the northern and northwestern regions of Rio de Janeiro State, Brazil, and made a comparative analysis of different methods to evaluate stability and adaptability of grain yield and popping expansion. To this end, 10 genotypes were evaluated (UNB2U-C3, UNB2U-C4, BRS Angela, Viçosa, Beija-Flor, IAC 112, IAC 125, Zélia, Jade, and UFVM2 Barão de Viçosa) in five environments. The Yates and Cochran method revealed that genotypes UFV2M Barão de Viçosa, BRS Angela and UNB2U-C3 were the most stable for grain yield. This method also indicated superiority of genotypes UNB2U-C3, UNB2U-C4, BRS Angela, Viçosa, IAC 125, and Zélia for popping expansion. The Plaisted and Peterson and Wricke methods demonstrated that genotypes Zélia and UNB2U-C4 were the most productive and stable. These methods indicated genotypes UNB2U-C3 and BRS Angela as the most stable for popping expansion. The Kang and Phan ranking system uses methods based on analysis of variance and classified population UNB2U-C4 as the genotype with the highest stability of grain production and confirmed cultivar BRS Angela as the most stable for popping expansion. Genotypes IAC 112 and UNB2U-C4 were the most stable and adapted for grain yield, according to the Lin and Binns method. The P(i) statistics also ranked UNB2U-C3 and UNB2U-C4 as the genotypes with the best predictability and capacity for popping expansion.

MISCELLANEOUS SECTIONS AND DETAILS OF FUEL STORAGE BUILDING (CPP603). INL ...

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

MISCELLANEOUS SECTIONS AND DETAILS OF FUEL STORAGE BUILDING (CPP-603). INL DRAWING NUMBER 200-0603-61-299-103032. ALTERNATE ID NUMBER 542-31-B-24. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

WEST ELEVATION OF FUEL STORAGE BUILDING (CPP603). PHOTO TAKEN LOOKING ...

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

WEST ELEVATION OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORTHEAST. INL PHOTO NUMBER HD-54-20-1. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

WEST ELEVATIONS AND SECTIONS OF FUEL STORAGE BUILDING (CPP603). INL ...

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

WEST ELEVATIONS AND SECTIONS OF FUEL STORAGE BUILDING (CPP-603). INL DRAWING NUMBER 200-063-61-299-103031. ALTERNATE ID NUMBER 542-31-B-23. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO SHOWING EMPLACEMENT STEEL BEAMS FUEL STORAGE BUILDING ...

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

CONSTRUCTION PROGRESS PHOTO SHOWING EMPLACEMENT STEEL BEAMS FUEL STORAGE BUILDING (CPP-603) LOOKING EAST. INL PHOTO NUMBER NRTS-51-1371. Unknown Photographer, 1/31/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

SIDING AND ROOF DETAILS OF FUEL STORAGE BUILDING (CPP603). INL ...

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

SIDING AND ROOF DETAILS OF FUEL STORAGE BUILDING (CPP-603). INL DRAWING NUMBER 200-0603-61-299-103033. ALTERNATE ID NUMBER 542-31-B-25. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

INTERIOR VIEW OF FUEL STORAGE BUILDING (CPP603) SHOWING CRANE ASSEMBLY ...

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

INTERIOR VIEW OF FUEL STORAGE BUILDING (CPP-603) SHOWING CRANE ASSEMBLY FOR TRANSFER PIT. INL PHOTO NUMBER NRTS-51-2404. Unknown Photographer, 5/31/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

PLAN VIEW OF FUEL STORAGE BUILDING (CPP603) SHOWING STORAGE BASINS. ...

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

PLAN VIEW OF FUEL STORAGE BUILDING (CPP-603) SHOWING STORAGE BASINS. INL DRAWING NUMBER 200-0603-00-706-051285. ALTERNATE ID NUMBER CPP-D-1285. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

WEST ELEVATION OF FUEL STORAGE BUILDING (CPP603). PHOTO TAKEN LOOKING ...

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

WEST ELEVATION OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORTHEAST. INL PHOTO NUMBER HD-54-20-3. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

SOUTH, EAST, NORTH ELEVATIONS AND SECTIONS OF FUEL STORAGE BUILDING ...

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

SOUTH, EAST, NORTH ELEVATIONS AND SECTIONS OF FUEL STORAGE BUILDING (CPP-603). INL DRAWING NUMBER 200-0603-61-299-103030. ALTERNATE ID NUMBER 542-31-B-22. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

INTERIOR OF SECOND FLOOR CONTROL ROOM OF FUEL STORAGE BUILDING ...

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

INTERIOR OF SECOND FLOOR CONTROL ROOM OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING SOUTHWEST. INL PHOTO NUMBER HD-54-19-2. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

NORTHERN PORTION OF WEST ELEVATION OF FUEL STORAGE BUILDING (CPP603). ...

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

NORTHERN PORTION OF WEST ELEVATION OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORTHEAST. INL PHOTO NUMBER HD-54-20-4. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

OBLIQUE PHOTO OF NORTHWEST CORNER OF FUEL STORAGE BUILDING (CPP603). ...

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

OBLIQUE PHOTO OF NORTHWEST CORNER OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING SOUTHEAST. INL PHOTO NUMBER HD-54-14-4. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

NORTHERN PORTION OF WEST ELEVATION OF FUEL STORAGE BUILDING (CPP603). ...

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

NORTHERN PORTION OF WEST ELEVATION OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING SOUTHEAST. INL PHOTO NUMBER HD-54-20-2. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

VIEW OF FECF HOT CELL OF FUEL STORAGE BUILDING (CPP603). ...

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

VIEW OF FECF HOT CELL OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORHTWEST. INL PHOTO NUMBER HD-54-18-3. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

VIEW OF TRANSFER BASIN CORRIDOR OF FUEL STORAGE BUILDING (CPP603). ...

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

VIEW OF TRANSFER BASIN CORRIDOR OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORTH. INL PHOTO NUMBER HD-54-17-2. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

OBLIQUE PHOTO OF NORTH ELEVATION OF FUEL STORAGE BUILDING (CPP603). ...

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

OBLIQUE PHOTO OF NORTH ELEVATION OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING SOUTH. INL PHOTO NUMBER HD-54-14-3. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

NASA Astrophysics Data System (ADS)

Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; Pierce, David A.; Ebert, William L.; Williams, Benjamin D.; Snyder, Michelle M. V.; Frank, Steven M.; George, Jaime L.; Kruska, Karen

2017-11-01

This paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersive X-ray spectroscopy. These glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability.

Novel Sorbent Development and Evaluation for the Capture of Krypton and Xenon from Nuclear Fuel Reprocessing Off-Gas Streams

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

Troy G. Garn; Mitchell R. Greenhalgh; Jack D. Law

2013-10-01

The release of volatile radionuclides generated during Used Nuclear Fuel reprocessing in the US will most certainly need to be controlled to meet US regulatory emission limits. A US DOE sponsored Off-Gas Sigma Team has been tasked with a multi-lab collaborative research and development effort to investigate and evaluate emissions and immobilization control technologies for the volatile radioactive species generated from commercial Used Nuclear Fuel (UNF) Reprocessing. Physical Adsorption technology is a simpler and potential economical alternative to cryogenic distillation processes that can be used for the capture of krypton and xenon and has resulted in a novel composite sorbentmore » development procedure using synthesized mordenite as the active material. Utilizing the sorbent development procedure, INL sigma team members have developed two composite sorbents that have been evaluated for krypton and xenon capacities at ambient and 191 K temperature using numerous test gas compositions. Adsorption isotherms have been generated to predict equilibration and maximum capacities enabling modeling to support process equipment scale-up.« less

Novel Sorbent Development and Evaluation for the Capture of Krypton and Xenon from Nuclear Fuel Reprocessing Off-Gas Streams

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

Troy G. Garn; Mitchell R. Greenhalgh; Jack D. Law

2013-09-01

The release of volatile radionuclides generated during Used Nuclear Fuel reprocessing in the US will most certainly need to be controlled to meet US regulatory emission limits. A US DOE sponsored Off-Gas Sigma Team has been tasked with a multi-lab collaborative research and development effort to investigate and evaluate emissions and immobilization control technologies for the volatile radioactive species generated from commercial Used Nuclear Fuel (UNF) Reprocessing. Physical Adsorption technology is a simpler and potential economical alternative to cryogenic distillation processes that can be used for the capture of krypton and xenon and has resulted in a novel composite sorbentmore » development procedure using synthesized mordenite as the active material. Utilizing the sorbent development procedure, INL sigma team members have developed two composite sorbents that have been evaluated for krypton and xenon capacities at ambient and 191 K temperature using numerous test gas compositions. Adsorption isotherms have been generated to predict equilibration and maximum capacities enabling modeling to support process equipment scale-up.« less

Molten tin reprocessing of spent nuclear fuel elements

DOEpatents

Heckman, Richard A.

1983-01-01

A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support the liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

Nuclear Energy Policy

DTIC Science & Technology

2007-07-12

Nuclear Waste Storage Act of 2007. Requires commercial nuclear power plants to transfer spent fuel from pools to dry storage ...enrichment, spent fuel recycling (also called reprocessing), and other fuel cycle facilities that could be used to produce nuclear weapons materials...that had used the leased fuel , along with supplies of fresh nuclear fuel , according to the GNEP concept; see [http://www.gnep.energy.gov].

VIEW OF CRANE LOADING AND UNLOADING AREA OF FUEL STORAGE ...

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

VIEW OF CRANE LOADING AND UNLOADING AREA OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORTH. INL PHOTO NUMBER HD-54-17-4. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

0BLIQUE PHOTO OF EAST ELEVATION OF FUEL STORAGE BUILDING (CPP603). ...

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

0BLIQUE PHOTO OF EAST ELEVATION OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING WEST. INL PHOTO NUMBER HD-54-15-1. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

VIEW OF CRANE LOADING AND UNLOADING AREA OF FUEL STORAGE ...

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

VIEW OF CRANE LOADING AND UNLOADING AREA OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING SOUTH. INL PHOTO NUMBER HD-54-17-1. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

VIEW OF SOUTH STORAGE BASIN NUMBER 1 OF FUEL STORAGE ...

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

VIEW OF SOUTH STORAGE BASIN NUMBER 1 OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORTHEAST. INL PHOTO NUMBER HD-54-18-4. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

VIEW OF MIDDLE STORAGE BASIN NUMBER 2 OF FUEL STORAGE ...

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

VIEW OF MIDDLE STORAGE BASIN NUMBER 2 OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORTHEAST. INL PHOTO NUMBER HD-54-17-3. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Interns reflect: the effect of formative assessment with feedback during pre-internship.

PubMed

McKenzie, Susan; Burgess, Annette; Mellis, Craig

2017-01-01

It is widely known that the opportunity for medical students to be observed and to receive feedback on their procedural skills performance is variable in the senior years. To address this problem, we provided our Pre-Intern (PrInt) students with "one-to-one" formative feedback on their ability to perform urethral catheterization (U/C) and hypothesized that their future practice of U/C as interns would benefit. This study sought to evaluate the performance and practice of interns in U/C 4-5 months after having received feedback on their performance of U/C as PrInt students. Between 2013 and 2014, two cohorts of interns, (total n=66) who had received recent formative feedback on their U/C performance as PrInt students at Central Clinical School, were invited to complete an anonymous survey. The survey contained nine closed unvalidated questions and one open-ended question, designed to allow interns to report on their current practice of U/C. Forty-one out of 66 interns (62%) completed the survey. Thirty-five out of 41 respondents (85%) reported that the assessment with feedback during their PrInt term was beneficial to their practice. Thirty of 41 (73%) reported being confident to perform U/C independently. Eleven out of 41 respondents (27%) reported that they had received additional training at intern orientation. Nine of the 11 interns (82%) reported that they had a small, but a significant, increase in confidence to perform U/C when compared with the 30 of the 41 respondents (73%) who had not ( p =0.03). Our results substantiate our hypothesis that further education by assessment with feedback in U/C during PrInt was of benefit to interns' performance. Additional educational reinforcement in U/C during intern orientation further improved intern confidence. Our results indicate that extra pre- and post-graduation procedural skills training, with feedback, should be universal.

Interns reflect: the effect of formative assessment with feedback during pre-internship

PubMed Central

McKenzie, Susan; Burgess, Annette; Mellis, Craig

2017-01-01

Background It is widely known that the opportunity for medical students to be observed and to receive feedback on their procedural skills performance is variable in the senior years. To address this problem, we provided our Pre-Intern (PrInt) students with “one-to-one” formative feedback on their ability to perform urethral catheterization (U/C) and hypothesized that their future practice of U/C as interns would benefit. This study sought to evaluate the performance and practice of interns in U/C 4–5 months after having received feedback on their performance of U/C as PrInt students. Methods Between 2013 and 2014, two cohorts of interns, (total n=66) who had received recent formative feedback on their U/C performance as PrInt students at Central Clinical School, were invited to complete an anonymous survey. The survey contained nine closed unvalidated questions and one open-ended question, designed to allow interns to report on their current practice of U/C. Results Forty-one out of 66 interns (62%) completed the survey. Thirty-five out of 41 respondents (85%) reported that the assessment with feedback during their PrInt term was beneficial to their practice. Thirty of 41 (73%) reported being confident to perform U/C independently. Eleven out of 41 respondents (27%) reported that they had received additional training at intern orientation. Nine of the 11 interns (82%) reported that they had a small, but a significant, increase in confidence to perform U/C when compared with the 30 of the 41 respondents (73%) who had not (p=0.03). Conclusion Our results substantiate our hypothesis that further education by assessment with feedback in U/C during PrInt was of benefit to interns’ performance. Additional educational reinforcement in U/C during intern orientation further improved intern confidence. Our results indicate that extra pre- and post-graduation procedural skills training, with feedback, should be universal. PMID:28138270

Carbon foam anode modified by urea and its higher electrochemical performance in marine benthic microbial fuel cell

NASA Astrophysics Data System (ADS)

Fu, Yubin; Lu, Zhikai; Zai, Xuerong; Wang, Jian

2015-08-01

Electrode materials have an important effect on the property of microbial fuel cell (MFC). Carbon foam is utilized as an anode and further modified by urea to improve its performance in marine benthic microbial fuel cell (BMFC) with higher voltage and output power. The electrochemical properties of plain carbon foam (PC) and urea-modified carbon foam (UC) are measured respectively. Results show that the UC obtains better wettability after its modification and higher anti-polarization ability than the PC. A novel phenomenon has been found that the electrical potential of the modified UC anode is nearly 100 mV lower than that of the PC, reaching -570 ±10 mV ( vs. SCE), and that it also has a much higher electron transfer kinetic activity, reaching 9399.4 mW m-2, which is 566.2-fold higher than that from plain graphite anode (PG). The fuel cell containing the UC anode has the maximum power density (256.0 mW m-2) among the three different BMFCs. Urea would enhance the bacteria biofilm formation with a more diverse microbial community and maintain more electrons, leading to a lower anodic redox potential and higher power output. The paper primarily analyzes why the electrical potential of the modified anode becomes much lower than that of others after urea modification. These results can be utilized to construct a novel BMFC with higher output power and to design the conditioner of voltage booster with a higher conversion ratio. Finally, the carbon foam with a bigger pore size would be a potential anodic material in conventional MFC.

Comparisons of sodium void and Doppler reactivities in large oxide and carbide LMFBRs

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

Su, S F

Sodium void and Doppler reactivities in two full scale (3000 MWth) LMFBRs are analyzed; one is fueled with UO/sub 2/ - PuO/sub 2/ and the other is fueled with UC - PuC. These two reactors are analyzed for beginning of life as well as for beginning and end of equilibrium cycle conditions, and the variations of these two safety parameters with burnup are explained. A series of comperative analyses of these two and several hypothetical reactors are carried out to determine how differences in fuel type, sodium content, and heavy metal concentration between an oxide and a carbide reactor affectmore » their sodium void and Doppler reactivities. The effect of the presence of conrol poison on sodium void reactivity is also addressed.« less

EAST/WEST TRUCK BAY AREA OF TRANSFER BASIN CORRIDOR OF FUEL ...

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

EAST/WEST TRUCK BAY AREA OF TRANSFER BASIN CORRIDOR OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORTHWEST. INL PHOTO NUMBER HD-54-19-1. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Life Cycle Sustainment of Commercial Off-the-Shelf (COTS) Support Equipment

DTIC Science & Technology

2012-05-01

CO E WS50AA 1 N9C422319 BHL10326 UC09AQ Jun-10 AR HOUSTON TX 808 VERTICAL CONST CO E WS50AA 1 N9C422321 BHL10328 UC09AS Jun-10 AR HILO HI 871...VERTICAL CONST CO E WQX8AA 1 N9C422528 BHL10535 UC09RP 23-Apr-10 AR HILO HI 871 VERTICAL CONST CO E WQX8AA 1 N9C422534 BHL10541 UC09RV 23-Apr-10 AR... HILO HI 871 VERTICAL CONST CO E WQX8AA 1 N9C422561 BHL10568 UC09SN 23-Apr-10 AR FT LEONARD WOOD MO 955 HORIZONTAL CONST CO E WRZUAA 1 N9C422375

Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

DOE PAGES

Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; ...

2017-08-30

Here, this paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li 2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl 2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersivemore » X-ray spectroscopy. These glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability.« less

Potential radiological impact of tornadoes on the safety of Nuclear Fuel Services' West Valley Fuel Reprocessing Plant. 2. Reentrainment and discharge of radioactive materials

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

Davis, W Jr

1981-07-01

This report describes results of a parametric study of quantities of radioactive materials that might be discharged by a tornado-generated depressurization on contaminated process cells within the presently inoperative Nuclear Fuel Services' (NFS) fuel reprocessing facility near West Valley, New York. The study involved the following tasks: determining approximate quantities of radioactive materials in the cells and characterizing particle-size distribution; estimating the degree of mass reentrainment from particle-size distribution and from air speed data presented in Part 1; and estimating the quantities of radioactive material (source term) released from the cells to the atmosphere. The study has shown that improperlymore » sealed manipulator ports in the Process Mechanical Cell (PMC) present the most likely pathway for release of substantial quantities of radioactive material in the atmosphere under tornado accident conditions at the facility.« less

Fragments of Citrullinated and MMP-degraded Vimentin and MMP-degraded Type III Collagen Are Novel Serological Biomarkers to Differentiate Crohn's Disease from Ulcerative Colitis.

PubMed

Mortensen, Joachim Høg; Godskesen, Line Elbjerg; Jensen, Michael Dam; Van Haaften, Wouter Tobias; Klinge, Lone Gabriels; Olinga, Peter; Dijkstra, Gerard; Kjeldsen, Jens; Karsdal, Morten Asser; Bay-Jensen, Anne-Christine; Krag, Aleksander

2015-10-01

A hallmark of inflammatory bowel disease [IBD] is chronic inflammation, which leads to excessive extracellular matrix [ECM] remodelling and release of specific protein fragments, called neoepitopes. We speculated that the biomarker profile panel for ulcerative colitis [UC] and Crohn's disease [CD] represent a heterogeneous expression pattern, and may be applied as a tool to aid in the differentiation between UC and CD. Serum biomarkers of degraded collagens I, III-IV [C1M, C3M, and C4M], collagen type 1 and IV formation [P1NP, P4NP], and citrullinated and MMP-degraded vimentin [VICM] were studied with a competitive ELISA assay system in a cohort including 164 subjects [CD n = 72, UC n = 60, and non-IBD controls n = 32] and a validation cohort of 61 subjects [CD n = 46, and UC n = 15]. Receiver operating characteristic curve analysis and logistic regression modelling were carried out to evaluate the discriminative power of the biomarkers. All biomarkers were corrected for confounding factors. VICM and C3M demonstrated the highest diagnostic power, alone, to differentiate CD from UC with an area under the curve [AUC] of 0.77 and 0.69, respectively. Furthermore, the biomarkers C1M [AUC = 0.81], C3M [AUC = 0.83], VICM [AUC = 0.83], and P1NP [AUC = 0.77] were best to differentiate UC from non-IBD. The best combinations of biomarkers to differentiate CD from UC and UC from non-IBD were VICM, C3M, C4M [AUC = 0.90] and VICM, C3M [AUC = 0.98] respectively. Specific extracellular matrix degradation markers are elevated in IBD and can discriminate CD from UC and UC from non-IBD controls with a high diagnostic accuracy. Copyright © 2015 European Crohn’s and Colitis Organisation (ECCO). Published by Oxford University Press. All rights reserved. For permissions, please email: journals.permissions@oup.com.

The association between three promoter polymorphisms of IL-10 and inflammatory bowel diseases (IBD): a meta-analysis.

PubMed

Zou, Liwei; Wang, Longsheng; Gong, Xijun; Zhao, Hong; Jiang, Anhong; Zheng, Suisheng

2014-02-01

To assess the relationship of the Interleukin-10 (IL-10) -1082G/A (rs1800896), -819C/T (rs1800871) and -592C/A (rs1800872) polymorphism with inflammatory bowel disease (IBD) by means of meta-analysis. Published data addressing the association between polymorphism of the IL-10 with Crohn's disease (CD) and Ulcerative colitis (UC) were selected from electronic databases. A total of 17 studies including 4132 cases and 5109 controls were included in this meta-analysis which detected whether -1082G/A, -819C/T and -592C/A polymorphism were associated with CD or UC susceptibility. The IL-10 -819C/T and -519C/A variant allele observed a significant association with UC (OR 1.16, 95%CI 1.03-1.31 and OR 1.19, 95%CI 1.03-1.38) not CD while there is no significant association between -1082G/A and UC or CD. The IL-10 -819C/T and -592C/A polymorphisms contribute to susceptibility to UC, but IL-10 -1082G/A polymorphism neither associated with CD nor UC.

Fuel supply of nuclear power industry with the introduction of fast reactors

NASA Astrophysics Data System (ADS)

Muraviev, E. V.

2014-12-01

The results of studies conducted for the validation of the updated development strategy for nuclear power industry in Russia in the 21st century are presented. Scenarios with different options for the reprocessing of spent fuel of thermal reactors and large-scale growth of nuclear power industry based on fast reactors of inherent safety with a breeding ratio of ˜1 in a closed nuclear fuel cycle are considered. The possibility of enhanced fuel breeding in fast reactors is also taken into account in the analysis. The potential to establish a large-scale nuclear power industry that covers 100% of the increase in electric power requirements in Russia is demonstrated. This power industry may be built by the end of the century through the introduction of fast reactors (replacing thermal ones) with a gross uranium consumption of up to ˜1 million t and the termination of uranium mining even if the reprocessing of spent fuel of thermal reactors is stopped or suffers a long-term delay.

Molten tin reprocessing of spent nuclear fuel elements. [Patent application; continuous process

DOEpatents

Heckman, R.A.

1980-12-19

A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support te liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

Fabrication and life testing of thermionic converters

NASA Technical Reports Server (NTRS)

Yang, L.; Bruce, R.

1973-01-01

An unfueled converter containing a chloride-fluoride duplex tungsten emitter of 4.78 eV vacuum work function was tested for 46,647 hours at an emitter temperature of 1973 K and an electrode power output of about 8 watts/sq cm. The test demonstrated the superior and stable performance of the (110) oriented tungsten emitter at high temperatures. Three 90 UC-10 ZrC(C/U = 1.04, tungsten additive = 4 wt %) fueled converters were fabricated and tested at an emitter temperature of 1873 K. Converter containing chloride-arc-cast duplex tungsten cladding showed temperature thermionic performance and slower rate of performance drop than converter containing chloride-fluoride duplex tungsten cladding. This is believed to be due to the superior fuel component diffusion resistance of the arc-cast tungsten substrate used in the fuel cladding. It was shown that a converter containing a carbide fueled chloride-arc-cast duplex tungsten emitter with an initial electrode power output of 6.80 watts/sq cm could still deliver an electrode power output of 6.16 watts/sq cm after 18,632 hours of operation at an emitter temperature of 1873 K.

Immobilization of Fast Reactor First Cycle Raffinate

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

Langley, K. F.; Partridge, B. A.; Wise, M.

This paper describes the results of work to bring forward the timing for the immobilization of first cycle raffinate from reprocessing fuel from the Dounreay Prototype Fast Reactor (PFR). First cycle raffinate is the liquor which contains > 99% of the fission products separated from spent fuel during reprocessing. Approximately 203 m3 of raffinate from the reprocessing of PFR fuel is held in four tanks at the UKAEA's site at Dounreay, Scotland. Two methods of immobilization of this high level waste (HLW) have been considered: vitrification and cementation. Vitrification is the standard industry practice for the immobilization of first cyclemore » raffinate, and many papers have been presented on this technique elsewhere. However, cementation is potentially feasible for immobilizing first cycle raffinate because the heat output is an order of magnitude lower than typical HLW from commercial reprocessing operations such as that at the Sellafield site in Cumbria, England. In fact, it falls within the upper end of the UK definition of intermediate level waste (ILW). Although the decision on which immobilization technique will be employed has yet to be made, initial development work has been undertaken to identify a suitable cementation formulation using inactive simulant of the raffinate. An approach has been made to the waste disposal company Nirex to consider the disposability of the cemented product material. The paper concentrates on the process development work that is being undertaken on cementation to inform the decision making process for selection of the immobilization method.« less

Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

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

Ragusa, Jean; Vierow, Karen

2011-09-01

The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzedmore » advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.« less

Rotational Rehybridization and the High Temperature Phase of UC2

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

Wen, Xiaodong; Rudin, Sven P.; Batista, Enrique R.

2012-12-03

The screened hybrid approximation (HSE) of density functional theory (DFT) is used to examine the structural, optical, and electronic properties of the high temperature phase, cubic UC(2). This phase contains C(2) units with a computed C-C distance of 1.443 Å which is in the range of a CC double bond; U is formally 4+, C(2) 4-. The closed shell paramagnetic state (NM) was found to lie lowest. Cubic UC(2) is found to be a semiconductor with a narrow gap, 0.4 eV. Interestingly, the C(2) units connecting two uranium sites can rotate freely up to an angle of 30°, indicating amore » hindered rotational solid. Ab-initio molecular dynamic simulations (HSE) show that the rotation of C(2) units in the low temperature phase (tetragonal UC(2)) occurs above 2000 K, in good agreement with experiment. The computed energy barrier for the phase transition from tetragonal UC(2) to cubic UC(2) is around 1.30 eV per UC(2). What is fascinating about this system is that at high temperature, the phase transformation to the cubic phase is associated with a rehybridization of the C atoms from sp to sp(3).« less

AERIAL SHOWING COMPLETED REMOTE ANALYTICAL FACILITY (CPP627) ADJOINING FUEL PROCESSING ...

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

AERIAL SHOWING COMPLETED REMOTE ANALYTICAL FACILITY (CPP-627) ADJOINING FUEL PROCESSING BUILDING AND EXCAVATION FOR HOT PILOT PLANT TO RIGHT (CPP-640). INL PHOTO NUMBER NRTS-60-1221. J. Anderson, Photographer, 3/22/1960 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Supply of enriched uranium for research reactors

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

Mueller, H.

1997-08-01

Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel onmore » December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.« less

Flowsheet Analysis of U-Pu Co-Crystallization Process as a New Reprocessing System

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

Shunji Homma; Jun-ichi Ishii; Jiro Koga

2006-07-01

A new fuel reprocessing system by U-Pu co-crystallization process is proposed and examined by flowsheet analysis. This reprocessing system is based on the fact that hexavalent plutonium in nitric acid solution is co-crystallized with uranyl nitrate, whereas it is not crystallized when uranyl nitrate does not exist in the solution. The system consists of five steps: dissolution of spent fuel, plutonium oxidation, U-Pu co-crystallization as a co-decontamination, re-dissolution of the crystals, and U re-crystallization as a U-Pu separation. The system requires a recycling of the mother liquor from the U-Pu co-crystallization step and the appropriate recycle ratio is determined bymore » flowsheet analysis such that the satisfactory decontamination is achieved. Further flowsheet study using four different compositions of LWR spent fuels demonstrates that the constant ratio of plutonium to uranium in mother liquor from the re-crystallization step is achieved for every composition by controlling the temperature. It is also demonstrated by comparing to the Purex process that the size of the plant based on the proposed system is significantly reduced. (authors)« less

Records of wells, test borings, and some measured geologic sections near the Western New York Nuclear Service Center, Cattaraugus County, New York

USGS Publications Warehouse

Bergeron, M.P.

1985-01-01

The Western New York Nuclear Service Center (WNYNSC) is a 3 ,336-acre tract of land in northern Cattaraugus County, NY, about 30 mi south of Buffalo. In 1963, 247 acres within the WNYNSC was developed for a nuclear-fuel reprocessing plant and ancillary facilities, including (1) a receiving and storage facility to store fuel prior to reprocessing, (2) underground storage tanks for liquid high-level radioactive wastes from fuel reprocessing, (3) a low-level wastewater treatment plant, and (4) two burial grounds for shallow burial of solid radioactive waste. A series of geologic and hydrologic investigations was done as part of the initial development and construction of the facilities by numerous agencies during 1960-62; these produced a large quantity of well data, some of which are difficult to locate or obtain. This report is a compilation of well and boring data collected during this period. The data include records of 236 wells, geologic logs of 145 wells and 167 test borings, and descriptions of 20 measured geologic sections. Two oversized maps show locations of the reported data. (USGS)

Production of Low Enriched Uranium Nitride Kernels for TRISO Particle Irradiation Testing

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

McMurray, J. W.; Silva, C. M.; Helmreich, G. W.

2016-06-01

A large batch of UN microspheres to be used as kernels for TRISO particle fuel was produced using carbothermic reduction and nitriding of a sol-gel feedstock bearing tailored amounts of low-enriched uranium (LEU) oxide and carbon. The process parameters, established in a previous study, produced phasepure NaCl structure UN with dissolved C on the N sublattice. The composition, calculated by refinement of the lattice parameter from X-ray diffraction, was determined to be UC 0.27N 0.73. The final accepted product weighed 197.4 g. The microspheres had an average diameter of 797±1.35 μm and a composite mean theoretical density of 89.9±0.5% formore » a solid solution of UC and UN with the same atomic ratio; both values are reported with their corresponding calculated standard error.« less

The effects of muscle contraction and recombinant osteocalcin on insulin sensitivity ex vivo.

PubMed

Levinger, I; Lin, X; Zhang, X; Brennan-Speranza, T C; Volpato, B; Hayes, A; Jerums, G; Seeman, E; McConell, G

2016-02-01

We tested whether GPRC6A, the putative receptor of undercarboxylated osteocalcin (ucOC), is present in mouse muscle and whether ucOC increases insulin sensitivity following ex vivo muscle contraction. GPPRC6A is expressed in mouse muscle and in the mouse myotubes from a cell line. ucOC potentiated the effect of ex vivo contraction on insulin sensitivity. Acute exercise increases skeletal muscle insulin sensitivity. In humans, exercise increases circulating ucOC, a hormone that increases insulin sensitivity in rodents. We tested whether GPRC6A, the putative receptor of ucOC, is present in mouse muscle and whether recombinant ucOC increases insulin sensitivity in both C2C12 myotubes and whole mouse muscle following ex vivo muscle contraction. Glucose uptake was examined in C2C12 myotubes that express GPRC6A following treatment with insulin alone or with insulin and increasing ucOC concentrations (0.3, 3, 10 and 30 ng/ml). In addition, glucose uptake, phosphorylated (p-)AKT and p-AS160 were examined ex vivo in extensor digitorum longus (EDL) dissected from C57BL/6J wild-type mice, at rest, following insulin alone, after muscle contraction followed by insulin and after muscle contraction followed by recombinant ucOC then insulin exposure. We observed protein expression of the likely receptor for ucOC, GPRC6A, in whole muscle sections and differentiated mouse myotubes. We observed reduced GPRC6A expression following siRNA transfection. ucOC significantly increased insulin-stimulated glucose uptake dose-dependently up to 10 ng/ml, in differentiated mouse C2C12 myotubes. Insulin increased EDL glucose uptake (∼30 %, p < 0.05) and p-AKT and p-AKT/AKT compared with rest (all p < 0.05). Contraction prior to insulin increased muscle glucose uptake (∼25 %, p < 0.05), p-AKT, p-AKT/AKT, p-AS160 and p-AS160/AS160 compared with contraction alone (all p < 0.05). ucOC after contraction increased insulin-stimulated muscle glucose uptake (∼12 % p < 0.05) and p-AS160 (<0.05) more than contraction plus insulin alone but without effect on p-AKT. In the absence of insulin and/or of contraction, ucOC had no significant effect on muscle glucose uptake. GPRC6A, the likely receptor of osteocalcin (OC), is expressed in mouse muscle. ucOC treatment augments insulin-stimulated skeletal muscle glucose uptake in C2C12 myotubes and following ex vivo muscle contraction. ucOC may partly account for the insulin sensitizing effect of exercise.

Electro-Magnetic Actuated Valve for MEMS Fuel Metering System

DTIC Science & Technology

2007-09-01

This model is utilized material properties of Silicon (Si), Copper (Cu), Nickel Iron ( NiFe ), and air. C11 Air NiSe Figure 5. Design of a simplified a... NiFe are defined and shown table 4. It is assumed that the properties of materials are independent of orientation (i.e. isotropic materials). Relative...dry filn resist. This process enables an integrated NiFe armature with a hole-in-the-wall within the main flow channel. UC Berkeley, Pisano - 2007

EXFILE: A program for compiling irradiation data on UN and UC fuel pins

NASA Technical Reports Server (NTRS)

Mayer, J. T.; Smith, R. L.; Weinstein, M. B.; Davison, H. W.

1973-01-01

A FORTRAN-4 computer program for handling fuel pin data is described. Its main features include standardized output, easy access for data manipulation, and tabulation of important material property data. An additional feature allows simplified preparation of input decks for a fuel swelling computer code (CYGRO-2). Data from over 300 high temperature nitride and carbide based fuel pin irradiations are listed.

Nuclear Fuel Reprocessing: U.S. Policy Development

DTIC Science & Technology

2006-11-29

to the chemical separation of fissionable uranium and plutonium from irradiated nuclear fuel. The World War II-era Manhattan Project developed...created the Atomic Energy Commission (AEC) and transferred production and control of fissionable materials from the Manhattan Project . As the exclusive

Status of the nuclear measurement stations for the process control of spent fuel reprocessing at AREVA NC/La Hague

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

Eleon, Cyrille; Passard, Christian; Hupont, Nicolas

2015-07-01

Nuclear measurements are used at AREVA NC/La Hague for the monitoring of spent fuel reprocessing. The process control is based on gamma-ray spectroscopy, passive neutron counting and active neutron interrogation, and gamma transmission measurements. The main objectives are criticality and safety, online process monitoring, and the determination of the residual fissile mass and activities in the metallic waste remained after fuel shearing and dissolution (empty hulls, grids, end pieces), which are put in radioactive waste drums before compaction. The whole monitoring system is composed of eight measurement stations which will be described in this paper. The main measurement stations no.more » 1, 3 and 7 are needed for criticality control. Before fuel element shearing for dissolution, station no. 1 allows determining the burn-up of the irradiated fuel by gamma-ray spectroscopy with HP Ge (high purity germanium) detectors. The burn-up is correlated to the {sup 137}Cs and {sup 134}Cs gamma emission rates. The fuel maximal mass which can be loaded in one bucket of the dissolver is estimated from the lowest burn-up fraction of the fuel element. Station no. 3 is dedicated to the control of the correct fuel dissolution, which is performed with a {sup 137}Cs gamma ray measurement with a HP Ge detector. Station no. 7 allows estimating the residual fissile mass in the drums filled with the metallic residues, especially in the hulls, from passive neutron counting (spontaneous fission and alpha-n reactions) and active interrogation (fission prompt neutrons induced by a pulsed neutron generator) with proportional {sup 3}He detectors. The measurement stations have been validated for the reprocessing of Uranium Oxide (UOX) fuels with a burn-up rate up to 60 GWd/t. This paper presents a brief overview of the current status of the nuclear measurement stations. (authors)« less

A Brief User's Guide to the Excel ® -Based DF Calculator

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

Jubin, Robert T.

2016-06-01

To understand the importance of capturing penetrating forms of iodine as well as the other volatile radionuclides, a calculation tool was developed in the form of an Excel ® spreadsheet to estimate the overall plant decontamination factor (DF). The tool requires the user to estimate splits of the volatile radionuclides within the major portions of the reprocessing plant, speciation of iodine and individual DFs for each off-gas stream within the Used Nuclear Fuel reprocessing plant. The Impact to the overall plant DF for each volatile radionuclide is then calculated by the tool based on the specific user choices. The Excelmore » ® spreadsheet tracks both elemental and penetrating forms of iodine separately and allows changes in the speciation of iodine at each processing step. It also tracks 3H, 14C and 85Kr. This document provides a basic user's guide to the manipulation of this tool.« less

Experimental study of UC polycrystals in the prospect of improving the as-fabricated sample purity

NASA Astrophysics Data System (ADS)

Raveu, Gaëlle; Martin, Guillaume; Fiquet, Olivier; Garcia, Philippe; Carlot, Gaëlle; Palancher, Hervé; Bonnin, Anne; Khodja, Hicham; Raepsaet, Caroline; Sauvage, Thierry; Barthe, Marie-France

2014-12-01

Uranium and plutonium carbides are candidate fuels for Generation IV nuclear reactors. This study is focused on the characterization of uranium monocarbide samples. The successive fabrication steps were carried out under atmospheres containing low oxygen and moisture concentrations (typically less than 100 ppm) but sample transfers occurred in air. Six samples were sliced from four pellets elaborated by carbothermic reaction under vacuum. Little presence of UC2 is expected in these samples. The α-UC2 phase was indeed detected within one of these UC samples during an XRD experiment performed with synchrotron radiation. Moreover, oxygen content at the surface of these samples was depth profiled using a recently developed nuclear reaction analysis method. Large oxygen concentrations were measured in the first micron below the sample surface and particularly in the first 100-150 nm. UC2 inclusions were found to be more oxidized than the surrounding matrix. This work points out to the fact that more care must be given at each step of UC fabrication since the material readily reacts with oxygen and moisture. A new glovebox facility using a highly purified atmosphere is currently being built in order to obtain single phase UC samples of better purity.

Alternative Fuels Data Center: UC Davis Pioneers Research for Plug-In

Science.gov Websites

gas vehicle District of Columbia's Government Fleet Uses a Wide Variety of Alternative Fuels Dec. 5 . Maryland County Fleet Uses Wide Variety of Alternative Fuels Jan. 17, 2015 Photo of a school bus Diego Feb. 2, 2013 Photo of neighborhood electric vehicle Mammoth Cave National Park Uses Only

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2008-10-17

safeguards-irrelevant.” The following facilities and activities were not on the separation list: ! 8 indigenous Indian power reactors ! Fast Breeder ...test Reactor (FTBR) and Prototype Fast Breeder Reactors (PFBR) under construction ! Enrichment facilities ! Spent fuel reprocessing facilities (except...potential use in a bomb. In addition, safeguards on enrichment, reprocessing plants, and breeder reactors would support the 2002 U.S. National Strategy to

Nuclear Energy Policy

DTIC Science & Technology

2008-01-28

2007. Requires commercial nuclear power plants to transfer spent fuel from pools to dry storage casks and then convey title to the Secretary of Energy...far more economical options for reducing fossil fuel use .15 (For more on federal incentives and the economics of nuclear power, see CRS Report RL33442...uranium enrichment, spent fuel recycling (also called reprocessing), and other fuel cycle facilities that could be used to produce nuclear weapons

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

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

Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.

This paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersive X-ray spectroscopy. Thesemore » glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability. (C) 2017 Published by Elsevier B.V.« less

INTERIOR VIEW OF FUEL STORAGE BUILDING (CPP603) LOOKING SOUTHWEST SHOWING ...

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

INTERIOR VIEW OF FUEL STORAGE BUILDING (CPP-603) LOOKING SOUTHWEST SHOWING STORAGE BASIN IN FOREGROUND, TRANSFER CRANE AND UNLOADER TO LEFT OF NORTH SIDE OF HOT CELL. INL PHOTO NUMBER NRTS-58-157. J. Anderson, Photographer, 1/15/1958 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

IMPACTS ON HUMAN HEALTH FROM THE COAL AND NUCLEAR FUEL CYCLES AND OTHER TECHNOLOGIES ASSOCIATED WITH ELECTRIC POWER GENERATION AND TRANSMISSION

EPA Science Inventory

The report evaluates major public health impacts of electric power generation and transmission associated with the nuclear fuel cycle and with coal use. Only existing technology is evaluated. For the nuclear cycle, effects of future use of fuel reprocessing and long-term radioact...

Silica-based waste form for immobilization of iodine from reprocessing plant off-gas streams

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

Matyáš, Josef; Canfield, Nathan; Sulaiman, Sannoh

A high selectivity and sorption capacity for iodine and a feasible consolidation to a durable SiO2-based waste form makes silver-functionalized silica aerogel (Ag0-aerogel) an attractive choice for the removal and sequestration of iodine compounds from the off-gas of a nuclear fuel reprocessing plant. Hot uniaxial pressing of iodine-loaded Ag0-aerogel (20.2 mass% iodine) at 1200°C for 30 min under 29 MPa pressure provided a partially sintered product with residual open porosity of 16.9% that retained ~93% of sorbed iodine. Highly iodine-loaded Ag0-aerogel was successfully consolidated by hot isostatic pressing at 1200°C with a 30-min hold and under 207 MPa. The fullymore » densified waste form had a bulk density of 3.3 g/cm3 and contained ~39 mass% iodine. The iodine was retained in the form of nano- and micro-particles of AgI that were uniformly distributed inside and along boundaries of fused silica grains.« less

Waste Estimates for a Future Recycling Plant in the US Based Upon AREVA Operating Experience - 13206

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

Foare, Genevieve; Meze, Florian; Bader, Sven

2013-07-01

Estimates of process and secondary wastes produced by a recycling plant built in the U.S., which is composed of a used nuclear fuel (UNF) reprocessing facility and a mixed oxide (MOX) fuel fabrication facility, are performed as part of a U.S. Department of Energy (DOE) sponsored study [1]. In this study, a set of common inputs, assumptions, and constraints were identified to allow for comparison of these wastes between different industrial teams. AREVA produced a model of a reprocessing facility, an associated fuel fabrication facility, and waste treatment facilities to develop the results for this study. These facilities were dividedmore » into a number of discrete functional areas for which inlet and outlet flow streams were clearly identified to allow for an accurate determination of the radionuclide balance throughout the facility and the waste streams. AREVA relied primarily on its decades of experience and feedback from its La Hague (reprocessing) and MELOX (MOX fuel fabrication) commercial operating facilities in France to support this assessment. However, to perform these estimates for a U.S. facility with different regulatory requirements and to take advantage of some technological advancements, such as in the potential treatment of off-gases, some deviations from this experience were necessary. A summary of AREVA's approach and results for the recycling of 800 metric tonnes of initial heavy metal (MTIHM) of LWR UNF per year into MOX fuel under the assumptions and constraints identified for this DOE study are presented. (authors)« less

10 CFR 50.54 - Conditions of licenses.

Code of Federal Regulations, 2012 CFR

2012-01-01

...)(1) Each nuclear power plant or fuel reprocessing plant licensee subject to the quality assurance... irradiated fuel. (ff) For licensees of nuclear power plants that have implemented the earthquake engineering... of rated thermal power only if the Commission finds that the state of onsite emergency preparedness...

10 CFR 50.54 - Conditions of licenses.

Code of Federal Regulations, 2013 CFR

2013-01-01

...)(1) Each nuclear power plant or fuel reprocessing plant licensee subject to the quality assurance... irradiated fuel. (ff) For licensees of nuclear power plants that have implemented the earthquake engineering... of rated thermal power only if the Commission finds that the state of onsite emergency preparedness...

Strength Measurements of Archive K Basin Sludge Using a Soil Penetrometer

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

Delegard, Calvin H.; Schmidt, Andrew J.; Chenault, Jeffrey W.

2011-12-06

Spent fuel radioactive sludge present in the K East and K West spent nuclear fuel storage basins now resides in the KW Basin in six large underwater engineered containers. The sludge will be dispositioned in two phases under the Sludge Treatment Project: (1) hydraulic retrieval into sludge transport and storage containers (STSCs) and transport to interim storage in Central Plateau and (2) retrieval from the STSCs, treatment, and packaging for shipment to the Waste Isolation Pilot Plant. In the years the STSCs are stored, sludge strength is expected to increase through chemical reaction, intergrowth of sludge crystals, and compaction andmore » dewatering by settling. Increased sludge strength can impact the type and operation of the retrieval equipment needed prior to final sludge treatment and packaging. It is important to determine whether water jetting, planned for sludge retrieval from STSCs, will be effective. Shear strength is a property known to correlate with the effectiveness of water jetting. Accordingly, the unconfined compressive strengths (UCS) of archive K Basin sludge samples and sludge blends were measured using a pocket penetrometer modified for hot cell use. Based on known correlations, UCS values can be converted to shear strengths. Twenty-six sludge samples, stored in hot cells for a number of years since last being disturbed, were identified as potential candidates for UCS measurement and valid UCS measurements were made for twelve, each of which was found as moist or water-immersed solids at least 1/2-inch deep. Ten of the twelve samples were relatively weak, having consistencies described as 'very soft' to 'soft'. Two of the twelve samples, KE Pit and KC-4 P250, were strong with 'very stiff' and 'stiff' consistencies described, respectively, as 'can be indented by a thumb nail' or 'can be indented by thumb'. Both of these sludge samples are composites collected from KE Basin floor and Weasel Pit locations. Despite both strong sludges having relatively high iron concentrations, attribution of their high strengths to this factor could not be made with confidence as other measured sludge samples, also from the KE Basin floor and of high iron concentration, were relatively weak. The observed UCS and shear strengths for the two strong sludges were greater than observed in any prior testing of K Basin sludge except for sludge processed at 185 C under hydrothermal conditions.« less

Aspects of remote maintenance in an FRG reprocessing plant from the manufacturer's viewpoint

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

Zeitzchel, G.; Tennie, M.; Saal, G.

In April 1986 a consortium led by Kraftwerk Union AG was commissioned by the German society for nuclear fuel reprocessing (DWK) to build the first West German commercial reprocessing plant for spent fuel assemblies. The main result of the planning efforts regarding remote maintenance operations inside the main process building was the introduction of FEMO technology (FEMO is an acronym based on German for remote handling modular technique). According to this technology the two cells in which the actual reprocessing (which is based on the PUREX technique) takes place are provided with frames to accommodate the process components (tanks, pumps,more » agitators, etc.), each frame together with the components which it supports forming one module. The two cells are inaccessible and windowless. For handling operations each cell is equipped with an overhead crane and a crane-like manipulator carrier system (MTS) with power manipulator. Viewing of the operations from outside the cells is made possible by television (TV) cameras installed at the crane, the MTS, and the manipulator. This paper addresses some examples of problems that still need to be solved in connection with FEMO handling. In particular, the need for close cooperation between the equipment operator, the component designer, the process engineer, the planning engineer, and the licensing authorities will be demonstrated.« less

Receptivity of a Cryogenic Coaxial Liquid Jet to Acoustic Disturbances

DTIC Science & Technology

2014-05-21

i iio thc UUU DISTRIBUTION STATEMENT A. Approved for public release; distribution unlimited. PA Clearance 14208 12Place Proper DISTRIBUTION STATEMENT...dynamic pressures are approximately equal. Uc Uo – Uc Uc – Ui (Uo > Ui) 2/12/1 2/12/1 io iioo c UUU St = Uc fnatD If St, D, Uc are held constant then

CESAR: A Code for Nuclear Fuel and Waste Characterisation

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

Vidal, J.M.; Grouiller, J.P.; Launay, A.

2006-07-01

CESAR (Simplified Evolution Code Applied to Reprocessing) is a depletion code developed through a joint program between CEA and COGEMA. In the late 1980's, the first use of this code dealt with nuclear measurement at the Laboratories of the La Hague reprocessing plant. The use of CESAR was then extended to characterizations of all entrance materials and for characterisation, via tracer, of all produced waste. The code can distinguish more than 100 heavy nuclides, 200 fission products and 100 activation products, and it can characterise both the fuel and the structural material of the fuel. CESAR can also make depletionmore » calculations from 3 months to 1 million years of cooling time. Between 2003-2005, the 5. version of the code was developed. The modifications were related to the harmonisation of the code's nuclear data with the JEF2.2 nuclear data file. This paper describes the code and explains the extensive use of this code at the La Hague reprocessing plant and also for prospective studies. The second part focuses on the modifications of the latest version, and describes the application field and the qualification of the code. Many companies and the IAEA use CESAR today. CESAR offers a Graphical User Interface, which is very user-friendly. (authors)« less

Metal–organic framework with optimally selective xenon adsorption and separation

DOE PAGES

Banerjee, Debasis; Simon, Cory M.; Plonka, Anna M.; ...

2016-06-13

Nuclear energy is considered among the most viable alternatives to our current fossil fuel based energy economy.1 The mass-deployment of nuclear energy as an emissions-free source requires the reprocessing of used nuclear fuel to mitigate the waste.2 One of the major concerns with reprocessing used nuclear fuel is the release of volatile radionuclides such as Xe and Kr. The most mature process for removing these radionuclides is energy- and capital-intensive cryogenic distillation. Alternatively, porous materials such as metal-organic frameworks (MOFs) have demonstrated the ability to selectively adsorb Xe and Kr at ambient conditions.3-8 High-throughput computational screening of large databases ofmore » porous materials has identified a calcium-based nanoporous MOF, SBMOF-1, as the most selective for Xe over Kr.9,10 Here, we affirm this prediction and report that SBMOF-1 exhibits by far the highest Xe adsorption capacity and a remarkable Xe/Kr selectivity under relevant nuclear reprocessing conditions. The exceptional selectivity of SBMOF-1 is attributed to its pore size tailored to Xe and its dense wall of atoms that constructs a binding site with a high affinity for Xe, as evident by single crystal X-ray diffraction and molecular simulation.« less

10 CFR 50.54 - Conditions of licenses.

Code of Federal Regulations, 2014 CFR

2014-01-01

... chapter. (a)(1) Each nuclear power plant or fuel reprocessing plant licensee subject to the quality... irradiated fuel. (ff) For licensees of nuclear power plants that have implemented the earthquake engineering... of rated thermal power only if the Commission finds that the state of onsite emergency preparedness...

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

Spencer, Barry B.; Bruffey, Stephanie H.; Jordan, Jacob A.

US regulations will require the removal of iodine and tritium, along with other volatile and semi-volatile radionuclides, from the off-gas streams of nuclear fuel reprocessing plants. Advanced tritium pretreatment (TPT) is an additional head-end operation that could be incorporated within nuclear fuel reprocessing plants. It utilizes nitrogen dioxide (NOR2R) as an oxidant to convert UOR2R to UR3ROR8R prior to traditional aqueous dissolution. Advanced TPT can result in the quantitative volatilization of both tritium and iodine. Up-front removal of iodine is of significant advantage because otherwise it distributes to several unit operations and the associated off-gas streams. The off-gas streams willmore » then require treatment to comply with US regulations. Advanced TPT is currently under development at Oak Ridge National Laboratory, and a kilogram-scale hot cell demonstration with used nuclear fuel (UNF) is planned for fiscal year (FY) 2018.« less

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

Green, D.W.; Heinrich, R.R.; Graczyk, D.G.

The ACL activities covered IFR fuel reprocessing, corium-concrete interactions, environmental samples, wastes, WIPP support, Advanced Photon Source, H-Tc superconductors, EBWR vessel, soils, illegal drug detection, quality control, etc.

Thermally and chemically responsive nanoporous materials for efficient capture of fission product gases.

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

Stroeve, Pieter; Faller, Roland

The objective of this project was to develop robust, high-efficiency materials for capture of fission product gases such as He, Xe and Kr in scenarios relevant for both reactor fuels and reprocessing operations. The relevant environments are extremely harsh, encompassing temperatures up to 1500 °C, high levels of radiation, as well as potential exposures to highly-reactive chemicals such as nitric acid and organic solvents such as kerosene. The requirement for nanostructured capture materials is driven in part by the very short (few micron) diffusion distances for product gases in nuclear fuel. We achieved synthesis, characterization and detailed modeling of themore » materials. Although not all materials reviewed in this report will be feasible for the ultimate goal of integration in nuclear fuel, nevertheless each material studied has particular properties which will enable an optimized material to be efficiently developed and characterized.« less

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

Reiche, Helmut Matthias; Vogel, Sven C.

New in situ data for the U-C system are presented, with the goal of improving knowledge of the phase diagram to enable production of new ceramic fuels. The none quenchable, cubic, δ-phase, which in turn is fundamental to computational methods, was identified. Rich datasets of the formation synthesis of uranium carbide yield kinetics data which allow the benchmarking of modeling, thermodynamic parameters etc. The order-disorder transition (carbon sublattice melting) was observed due to equal sensitivity of neutrons to both elements. This dynamic has not been accurately described in some recent simulation-based publications.

Critical review of carbon monoxide pressure measurements in the uranium carbon oxygen ternary system

NASA Astrophysics Data System (ADS)

Gossé, S.; Guéneau, C.; Chatillon, C.; Chatain, S.

2006-06-01

For high temperature reactors (HTR), the high level of fuel operating temperature in normal and accidental conditions requires to predict the possible chemical interactions between the fuel components. Among the concerns of the TRISO fuel particle thermomechanical behavior, it is necessary to better understand the carbon monoxide formation due to chemical interactions at the UO2 kernel and graphite buffer's interface. In a first step, the thermodynamic properties of the U-C-O system have to be assessed. The experimental data from literature on the equilibrium CO gas pressure measurements in the UO2-UC2-C ternary section of the U-C-O system are critically reviewed. Discrepancies between the different determinations can be explained - (i) by the different gaseous flow regimes in the experiments and - (ii) by the location of the measuring pressure gauge away from the reaction site. Experimental values are corrected - (i) from the gaseous flow type (molecular, transition or viscous) defined by the Knudsen number and - (ii) from the thermomolecular effect due to the temperature gradient inside the experimental vessels. Taking account of the selected and corrected values improves greatly the consistency of the original set of measurements.

Characterization and simulation of soft gamma-ray mirrors for their use with spent fuel rods at reprocessing facilities

DOE PAGES

Ruz, J.; Descalle, M. A.; Alameda, J. B.; ...

2016-05-24

The use of a grazing incidence optic to selectively reflect K-shell fluorescence emission and isotope-specific lines from special nuclear materials is a highly desirable nondestructive analysis method for use in reprocessing fuel environments. Preliminary measurements have been performed, and a simulation suite has been developed to give insight into the design of the x ray optics system as a function of the source emission, multilayer coating characteristics, and general experimental configurations. As a result, the experimental results are compared to the predictions from our simulation toolkit to illustrate the ray-tracing capability and explore the effect of modified optics in futuremore » measurement campaigns.« less

Technetium-99 and strontium-90: Abundance determination at ultratrace sensitivity by AMS as signatures of undeclared nuclear reprocessing activity

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

McAninch, J.E.; Proctor, I.D.

1995-03-01

The purpose of this White Paper is to examine the use of the ultratrace technique Accelerator Mass Spectrometry (AMS) to lower detection limits for {sup 99}Tc and {sup 90}Sr, and to examine the utility of these isotopes as signatures of a convert reprocessing facility. The International Atomic Energy Agency (IAEA) has committed to improving the effectiveness of the IAEA Safeguards System. This is in some degree a result of the discovery in 1991 of an undeclared Iraqi EMIS program. Recommendations from the March 1993 Consultants Group Meeting have resulted in several studies and follow on field trials to identify environmentalmore » signatures from covert nuclear fuel reprocessing activity. In particular, the April, 1993 reports of the Standing Advisory Group on Safeguards Implementation (SAGSI) identified the long-lived radioisotopes Technetium-99 and strontium-90 as two reliable signatures of fuel reprocessing activity. This report also suggested pathways in the chemical processing of irradiated fuel where these elements would be volatilized and potentially released in amounts detectable with ultratrace sensitivity techniques. Based on measured {sup 99}Tc background levels compiled from a variety of sources, it is estimated that AMS can provide 10% measurements of environmental levels of {sup 99}Tc in a few minutes using modestly sized samples: a few grams for soils, plants, or animal tissues; one to several liters for rain or seawater samples; and tens to hundreds of cubic meters for air sampling. Small sample sizes and high sample throughput result in significant increases in feasibility, cost effectiveness, and quality of data for a regional monitoring program. Similar results are expected for {sup 90}Sr.« less

Receptivity of a Cryogenic Coaxial Gas-Liquid Jet to Acoustic Disturbances (Briefing Charts)

DTIC Science & Technology

2014-03-01

meas = Δt Δs Uc,meas 2/12/1 0 2/12/1 0 , i iio thc UUU      DISTRIBUTION STATEMENT A. Approved for public release; distribution unlimited...this point, dynamic pressures are approximately equal. Uc Uo – Uc Uc – Ui (Uo > Ui) 2/12/1 2/12/1 io iioo c UUU      St = Uc fnatD If St, D, Uc

Reconstituted asbestos matrix for fuel cells

NASA Technical Reports Server (NTRS)

Mcbryar, H.

1975-01-01

Method is described for reprocessing commercially available asbestos matrix stock to yield greater porosity and bubble pressure (due to increased surface tension), improved homogeneity, and greater uniformity.

10 CFR 110.40 - Commission review.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Canada. (5) An export involving assistance to end uses related to isotope separation, chemical reprocessing, heavy water production, advanced reactors, or the fabrication of nuclear fuel containing...

First-principles study of uranium carbide: Accommodation of point defects and of helium, xenon, and oxygen impurities

NASA Astrophysics Data System (ADS)

Freyss, Michel

2010-01-01

Point defects and volatile impurities (helium, xenon, oxygen) in uranium monocarbide UC are studied by first-principles calculations. Preliminarily, bulk properties of UC and of two other uranium carbide phases, UC2 and U2C3 , are calculated in order to compare them to experimental data and to get confidence in the use of the generalized gradient approximation for this class of compounds. The subsequent study of different types of point defects shows that the carbon sublattice best accommodates the defects. The perturbation of the crystal structure induced by the defects is weak and the interaction between defects is found short range. Interstitial carbon dumbbells possibly play an important role in the diffusion of carbon atoms. The most favorable location of diluted helium, xenon, and oxygen impurities in the UC crystal lattice is then determined. The rare-gas atoms occupy preferably a uranium substitution site or a uranium site in a U-C bivacancy. But their incorporation in UC is, however, not energetically favorable, especially for xenon, suggesting their propensity to diffuse in the material and/or form bubbles. On the other hand, oxygen atoms are very favorably incorporated as diluted atoms in the UC lattice, confirming the easy oxidation of UC. The oxygen atoms preferably occupy a carbon substitution site or the carbon site of a U-C bivacancy. Our results are compared to available experimental data on UC and to similar studies by first-principles calculations for other carbides and nitrides with the rock-salt structure.

Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides

DOEpatents

Lloyd, M.H.

1981-01-09

Method for direct coprocessing of nuclear fuels derived from a product stream of fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides

DOEpatents

Lloyd, Milton H.

1983-01-01

Method for direct coprocessing of nuclear fuels derived from a product stream of a fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

Sensitivity Analysis of Fuel Centerline Temperatures in SuperCritical Water-cooled Reactors (SCWRs)

NASA Astrophysics Data System (ADS)

Abdalla, Ayman

SuperCritical Water-cooled Reactors (SCWRs) are one of the six nuclear-reactor concepts currently being developed under the Generation-IV International Forum (GIF). A main advantage of SCW Nuclear Power Plants (NPPs) is that they offer higher thermal efficiencies compared to those of current conventional NPPs. Unlike today's conventional NPPs, which have thermal efficiencies between 30 - 35%, SCW NPPs will have thermal efficiencies within a range of 45 - 50%, owing to high operating temperatures and pressures (i.e., coolant temperatures as high as 625°C at 25 MPa pressure). The use of current fuel bundles with UO2 fuel at the high operating parameters of SCWRs may cause high fuel centerline temperatures, which could lead to fuel failure and fission gas release. Studies have shown that when the Variant-20 (43-element) fuel bundle was examined at SCW conditions, the fuel centerline temperature industry limit of 1850°C for UO2 and the sheath temperature design limit of 850°C might be exceeded. Therefore, new fuel-bundle designs, which comply with the design requirements, are required for future use in SCWRs. The main objective of this study to conduct a sensitivity analysis in order to identify the main factors that leads to fuel centerline temperature reduction. Therefore, a 54-element fuel bundle with smaller diameter of fuel elements compared to that of the 43-element bundle was designed and various nuclear fuels are examined for future use in a generic Pressure Tube (PT) SCWR. The 54-element bundle consists of 53 heated fuel elements with an outer diameter of 9.5 mm and one central unheated element of 20-mm outer diameter which contains burnable poison. The 54-element fuel bundle has an outer diameter of 103.45 mm, which is the same as the outer diameter of the 43-element fuel bundle. After developing the 54-element fuel bundle, one-dimensional heat-transfer analysis was conducted using MATLAB and NIST REFPROP programs. As a result, the Heat Transfer Coefficient (HTC), bulk-fluid, sheath and fuel centerline temperature profiles were generated along the heated length of 5.772 m for a generic fuel channel. The fuel centerline and sheath temperature profiles have been determined at four Axial Heat Flux Profiles (AHFPs) using an average thermal power per channel of 8.5 MWth. The four examined AHFPs are the uniform, cosine, upstream-skewed and downstream-skewed profiles. Additionally, this study focuses on investigating a possibility of using low, enhanced and high thermal-conductivity fuels. The low thermal-conductivity fuels, which have been examined in this study, are uranium dioxide (UO 2), Mixed Oxide (MOX) and Thoria (ThO2) fuels. The examined enhanced thermal-conductivity fuels are uranium dioxide - silicon carbide (UO2 - SiC) and uranium dioxide - beryllium oxide (UO2 - BeO). Lastly, uranium carbide (UC), uranium dicarbide (UC2) and uranium nitride (UN) are the selected high thermal-conductivity fuels, which have been proposed for use in SCWRs. A comparison has been made between the low, enhanced and high thermal-conductivity fuels in order to identify the fuel centerline temperature behaviour when different nuclear fuels are used. Also, in the process of conducting the sensitivity analysis, the HTC was calculated using the Mokry et al. correlation, which is the most accurate supercritical water heat-transfer correlation so far. The sheath and the fuel centerline temperature profiles were determined for two cases. In Case 1, the HTC was calculated based on the Mokry et al. correlation, while in Case 2, the HTC values calculated for Case 1 were multiplied by a factor of 2. This factor was used in order to identify the amount of decrease in temperatures if the heat transfer is enhanced with appendages. Results of this analysis indicate that the use of the newly developed 54-element fuel bundle along with the proposed fuels is promising when compared with the Variant-20 (43-element) fuel bundle. Overall, the fuel centerline and sheath temperatures were below the industry and design limits when most of the proposed fuels were examined in the 54-element fuel bundle, however, the fuel centerline temperature limit was exceeded while MOX fuel was examined. Keywords: SCWRs, Fuel Centerline Temperature, Sheath Temperature, High Thermal Conductivity Fuels, Low Thermal Conductivity Fuels, HTC.

10 CFR Appendix I to Part 110 - Illustrative List of Reprocessing Plant Components Under NRC Export Licensing Authority

Code of Federal Regulations, 2010 CFR

2010-01-01

... transuranic elements. Different technical processes can accomplish this separation. However, over the years Purex has become the most commonly used and accepted process. Purex involves the dissolution of... facilities have process functions similar to each other, including: irradiated fuel element chopping, fuel...

Clostridium difficile infection worsens the prognosis of ulcerative colitis

PubMed Central

Negrón, María E; Barkema, Herman W; Rioux, Kevin; De Buck, Jeroen; Checkley, Sylvia; Proulx, Marie-Claude; Frolkis, Alexandra; Beck, Paul L; Dieleman, Levinus A; Panaccione, Remo; Ghosh, Subrata; Kaplan, Gilaad G

2014-01-01

BACKGROUND: The impact of Clostridium difficile infections among ulcerative colitis (UC) patients is well characterized. However, there is little knowledge regarding the association between C difficile infections and postoperative complications among UC patients. OBJECTIVE: To determine whether C difficile infection is associated with undergoing an emergent colectomy and experiencing postoperative complications. METHODS: The present population-based case-control study identified UC patients admitted to Calgary Health Zone hospitals for a flare between 2000 and 2009. C difficile toxin tests ordered in hospital or 90 days before hospital admission were provided by Calgary Laboratory Services (Calgary, Alberta). Hospital records were reviewed to confirm diagnoses and to extract clinical data. Multivariate logistic regression analyses were performed among individuals tested for C difficile to examine the association between C difficile infection and emergent colectomy and diagnosis of any postoperative complications and, secondarily, an infectious postoperative complication. Estimates were presented as adjusted ORs with 95% CIs. RESULTS: C difficile was tested in 278 (58%) UC patients and 6.1% were positive. C difficile infection was associated with an increased risk for emergent colectomy (adjusted OR 3.39 [95% CI 1.02 to 11.23]). Additionally, a preoperative diagnosis of C difficile was significantly associated with the development of postoperative infectious complications (OR 4.76 [95% CI 1.10 to 20.63]). CONCLUSION: C difficile diagnosis worsened the prognosis of UC by increasing the risk of colectomy and postoperative infectious complications following colectomy. Future studies are needed to explore whether early detection and aggressive management of C difficile infection will improve UC outcomes. PMID:25157528

Prognostic value of the urea:creatinine ratio in decompensated heart failure and its relationship with acute kidney damage.

PubMed

Josa-Laorden, C; Sola, A; Giménez-López, I; Rubio-Gracia, J; Garcés-Horna, V; Pérez-Calvo, J I

Worsening renal function is associated with an adverse prognosis for patients with acute heart failure (AHF). Urea-creatinine ratio (U:C ratio) might be useful for measuring renal function and could help stratify patients with AHF. An observational and prospective study was conducted to analyse the prognostic value of the U:C ratio, measured during the first 24-28 hours of admission, for patients hospitalised for decompensated Heart failure, and its relationship with estimated glomerular filtration rate (eGFR) and acute kidney injury (AKI). The study included 204 patients, with a mean age of 79.3 years, and a median eGFR of 55 mL/min/1.73m 2 . In the multivariate analysis, an U:C ratio above the median (50) was related to the development of AKI (36.5% vs. 21.9%) and to increased mortality, both overall (OR 2.75) and by HF (OR 3.50) in long term. In combination with eGFR, the U:C ratio showed prognostic value in patients with normal eGFR (mortality of 4.4% for an U:C ratio ≤ 50 vs. 22% for U:C ratio > 50; p=0.01), as well as a better predictive capacity for AKI than each of them separately (AUC, 0.718; 95% CI 0.643-0.793; p>.000). An U:C ratio > 50 is a predictor of increased long-term mortality for patients hospitalised for decompensated HF and with normal eGFR. Given the simplicity of this biomarker, its use in clinical practice should be more systematic. Copyright © 2018 Elsevier España, S.L.U. and Sociedad Española de Medicina Interna (SEMI). All rights reserved.

Approaches to improve the stability of the antiviral agent UC781 in aqueous solutions.

PubMed

Damian, Festo; Fabian, Judit; Friend, David R; Kiser, Patrick F

2010-08-30

In this work, we evaluated the chemical stability profiles of UC781 based solutions to identify excipients that stabilize the microbicidal agent UC781. When different antioxidants were added to UC781 in sulfobutylether-beta-cyclodextrin (SBE-beta-CD) solutions and subjected to a 50 degrees C stability study, it was observed that EDTA was a better stabilizing agent than sodium metabisulfite, glutathione or ascorbic acid. Some antioxidants accelerated the degradation of UC781, suggesting metal-catalyzed degradation of UC781. Furthermore, we observed substantial degradation of UC781 when stored in 1% Tween 80 and 1% DMSO solutions alone or in those with 10mM EDTA. On the other hand, improved stability of UC781 in the presence of 100 and 200mM of EDTA was observed in these solutions. The addition of both EDTA and citric acid in the stock solutions resulted in recovery of more than 60% of UC781 after 12 weeks. Generally, 10% SBE-beta-CD in the presence of EDTA and citric acid stabilized UC781 solutions: the amount of UC781 recovered approaching 95% after 12 weeks of storage at 40 degrees C. We also showed that the desulfuration reaction of the UC781 thioamide involves oxygen by running solution stability studies in deoxygenated media. Improved stability of UC781 in the present study indicates that the incorporation of EDTA, citric acid and SBE-beta-CD and the removal of oxygen in formulations of this drug will aid in increasing the stability of UC781 where solutions of the drug are required. Published by Elsevier B.V.

National Policy Implications of Storing Nuclear Waste in the Pacific Region,

DTIC Science & Technology

1981-01-01

US Congress, Senate, Committee on Energy and Natural Resources, Pacific Spent Nuclear Fuel Storage , Hearing...selected. 17 One type of shipping cask which has been used to transport spent fuel assemblies to the Nevada Test Site is a leakproof steel cask that can...discussion the following conclusions on the nuclear waste storage issue appear valid. The Reagan decision to reprocess spent fuel has not changed US

Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, task 17208: Final report

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

Amoroso, J. W.; Marra, J. C.

2015-08-26

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics)more » over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).« less

Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208: Final report

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

Amoroso, J. W.; Marra, J. C.

2015-08-26

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics)more » over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).« less

Associations between NRAMP1 Polymorphisms and Susceptibility to Ulcerative Colitis/Crohn's Disease: A Meta-Analysis.

PubMed

Sun, Manyi; Zhang, Li; Shi, Songli

2016-01-01

Multiple environmental and genetic factors contribute to the risks of ulcerative colitis (UC) and Crohn's disease (CD). Several allelic variants have been identified in natural resistance associated macrophage protein 1 (NRAMP1) gene; however, their association with UC/CD remains conflicting. The purpose of this study was to evaluate whether NRAMP1 polymorphisms are associated with the susceptibility to UC/CD. A meta-analysis on the association between the NRAMP1 polymorphisms and susceptibility to UC/CD was performed. Relevant studies were retrieved from the databases. After eligible data were extracted, Mantel-Haenszel statistics and random/fixed effects model were applied to calculate the pooled odds radio (OR) and 95% confidence interval (95% CI). Seven articles containing 536 UC cases, 997 CD cases, and 1361 controls were collected. No significant association between allele 2 frequency of NRAMP1 and susceptibility to UC/CD was detected in overall population (all p > 0.05). However, increased UC/CD risk for allele 3 was observed in Caucasian population (OR = 1.27, 95% CI = 1.08~1.50, p = 0.04), whereas decreased UC/CD risk was detected in non-Caucasian population (OR = 0.72, 95% CI = 0.60~0.87, p < 0.001), under "allele 3 vs. other alleles" model. Moreover, a significant increase in CD risk for T carrier frequency of -237 C/T (OR = 0.44, 95% CI, 0.26~0.75, p = 0.003) was detected, but not 274 C/T and 1729+55del4 (TGTG) +/del. The polymorphism of -237 C/T is related to the risk of CD; and the association of allele 3 with UC/CD risk differs in Caucasian and non-Caucasian population, which might be the potential biomarkers for clinical diagnosis of UC/CD.

Development of Crystallizer for Advanced Aqueous Reprocessing Process

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

Tadahiro Washiya; Atsuhiro Shibata; Toshiaki Kikuchi

2006-07-01

Crystallization is one of the remarkable technologies for future fuel reprocessing process that has safety and economical advantages. Japan Atomic Energy Agency (JAEA) (former Japan Nuclear Cycle Development Institute), Mitsubishi Material Corporation and Saitama University have been developing the crystallization process. In previous study, we carried out experimental studies with uranium, MOX and spent fuel conditions, and flowsheet analysis was considered. In association with these studies, an innovative continuous crystallizer and its system was developed to ensure high process performance. From the design study, an annular type continuous crystallizer was selected as the most promising design, and performance was confirmedmore » by small-scale test and engineering scale demonstration at uranium crystallization conditions. In this paper, the design study and the demonstration test results are described. (authors)« less

Corrosion property of 9Cr-ODS steel in nitric acid solution for spent nuclear fuel reprocessing

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

Takeuchi, M.; Koizumi, T.; Inoue, M.

2013-07-01

Corrosion tests of oxide dispersion strengthened with 9% Cr (9Cr-ODS) steel, which is one of the desirable materials for cladding tube of sodium-cooled fast reactors, in pure nitric acid solution, spent FBR fuel solution, and its simulated solution were performed to understand the corrosion behavior in a spent nuclear fuel reprocessing. In this study, the 9Cr-ODS steel with lower effective chromium content was evaluated to understand the corrosion behavior conservatively. As results, the tube-type specimens of the 9Cr-ODS steels suffered severe weight loss owing to active dissolution at the beginning of the immersion test in pure nitric acid solution inmore » the range from 1 to 3.5 M. In contrast, the weight loss was decreased and they showed a stable corrosion in the higher nitric acid concentration, the dissolved FBR fuel solution, and its simulated solution by passivation. The corrosion rates of the 9Cr-ODS steel in the dissolved FBR fuel solution and its simulated solution were 1-2 mm/y and showed good agreement with each other. The passivation was caused by the shift of corrosion potential to noble side owing to increase in nitric acid concentration or oxidative ions in the dissolved FBR fuel solution and the simulated spent fuel solution. (authors)« less

Between-cycle laser system for depressurization and resealing of modified design nuclear fuel assemblies

DOEpatents

Bradley, John G.

1982-01-01

A laser beam is used to puncture fuel cladding for release of contained pressurized fission gas from plenum sections or irradiated fuel pins. Exhausted fission gases are collected and trapped for safe disposal. The laser beam, adjusted to welding mode, is subsequently used to reseal the puncture holes. The fuel assembly is returned to additional irradiation or, if at end of reactivity lifetime, is routed to reprocess. The fuel assembly design provides graded cladding lengths, by rows or arrays, such that the cladding of each component fuel element of the assembly is accessible to laser beam reception.

Insufficiency of pro-heparin-binding epidermal growth factor-like growth factor shedding enhances hypoxic cell death in H9c2 cardiomyoblasts via the activation of caspase-3 and c-Jun N-terminal kinase.

PubMed

Uetani, Teruyoshi; Nakayama, Hironao; Okayama, Hideki; Okura, Takafumi; Higaki, Jitsuo; Inoue, Hirofumi; Higashiyama, Shigeki

2009-05-01

Heparin-binding epidermal growth factor-like growth factor (HB-EGF) is a cardiogenic and cardiohypertrophic growth factor. ProHB-EGF, a product of the Hb-egf gene and the precursor of HB-EGF, is anchored to the plasma membrane. Its ectodomain region is shed by a disintegrin and metalloproteases (ADAMs) when activated by various stimulations. It has been reported that an uncleavable mutant of Hb-egf, uc-Hb-egf, produces uc-proHB-EGF, which is not cleaved by ADAMs and causes dilation of the heart in knock-in mice. This suggests that the shedding of proHB-EGF is essential for the development and survival of cardiomyocytes: however, the molecular mechanism involved has remained unclear. In this study, we investigated the relationship between uc-proHB-EGF expression and cardiomyocyte survival. Human uc-proHB-EGF was adenovirally introduced into the rat cardiomyoblast cell line H9c2, and the cells were cultured under normoxic and hypoxic conditions. Uc-proHB-EGF-expressing H9c2 cells underwent apoptosis under normoxic conditions, which distinctly increased under hypoxic conditions. Furthermore, we observed an increased Caspase-3 activity, reactive oxygen species accumulation, and an increased c-Jun N-terminal kinase (JNK) activity in the uc-proHB-EGF-expressing H9c2 cells. Treatment of the uc-proHB-EGF transfectants with inhibitors of Caspase-3, reactive oxygen species, and JNK, namely, Z-VAD-fmk, N-acetylcysteine, and SP600125, respectively, significantly reduced hypoxic cell death. These data indicate that insufficiency of proHB-EGF shedding under hypoxic stress leads to cardiomyocyte apoptosis via Caspase-3- and JNK-dependent pathways.

HLW Return from France to Germany - 15 Years of Experience in Public Acceptance and Technical Aspects - 12149

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

Graf, Wilhelm

Since in 1984 the national reprocessing concept was abandoned the reprocessing abroad was the only existing disposal route until 1994. With the amendment of the Atomic Energy Act in 2001 spent fuel management changed completely since from 1 June 2005 any delivery of spent fuel to reprocessing plants was prohibited and the direct disposal of spent fuel became mandatory. Until 2005 the total amount of spent fuel to be reprocessed abroad added up to 6080 t HM, 5309 t HM thereof in France. The waste generated from reprocessing - alternatively an equivalent amount of radioactive material - has to bemore » returned to the country of origin according to the commercial contracts signed between the German utilities and COGEMA, now AREVA NC, in France and BNFL, now INS in UK. In addition the German and the French government exchanged notes with the obligation of both sides to enable and support the return of reprocessing residues or equivalents to Germany. The return of high active vitrified waste from La Hague to the interim storage facility at Gorleben was demanding from the technical view i. e. the cask design and the transport. Unfortunately the Gorleben area served as a target for nuclear opponents from the first transport in 1996 to the latest one in 2011. The protection against sabotage of the railway lines and mass protests needed highly improved security measures. In France and Germany special working forces and projects have been set up to cope with this extraordinary situation. A complex transport organization was established to involve all parties in line with the German and French requirements during transport. The last transport of vitrified residues from France has been completed successfully so far thus confirming the efficiency of the applied measures. Over 15 years there was and still is worldwide no comparable situation it is still unique. Summing up, the exceptional project handling challenge that resulted from the continuous anti-nuclear civil disobedience in Germany over the whole 15-year long project running time could be faced efficiently. It has to be concluded that despite of all problems the anti-nuclear activities have caused so far, all transports of vitrified HLW have always been completed successfully by adapting the commonly established safety, security and public acceptance measures to the special conditions and needs in Germany and coordinating the activities of all parties involved but at the expense of high costs for industry and government and a challenging operational complexity. Apart from an anticipatory project planning a good communication between all involved industrial parties and the French and the German government was the key to the effective management of such shipments and to minimize the radiological, economic, environmental, public and political impact. The future will show how efficiently the gained experience can be used for further return projects which are to be realized since no reprocessed waste has yet been returned from UK and neither the medium-level nor the low-level radioactive waste has been transferred from France to Germany. (author)« less

Evaluation of various carbon blacks and dispersing agents for use in the preparation of uranium microspheres with carbon

NASA Astrophysics Data System (ADS)

Hunt, R. D.; Johnson, J. A.; Collins, J. L.; McMurray, J. W.; Reif, T. J.; Brown, D. R.

2018-01-01

A comparison study on carbon blacks and dispersing agents was performed to determine their impacts on the final properties of uranium fuel kernels with carbon. The main target compositions in this internal gelation study were 10 and 20 mol % uranium dicarbide (UC2), which is UC1.86, with the balance uranium dioxide. After heat treatment at 1900 K in flowing carbon monoxide in argon for 12 h, the density of the kernels produced using a X-energy proprietary carbon suspension, which is commercially available, ranged from 96% to 100% of theoretical density (TD), with full conversion of UC to UC2 at both carbon concentrations. However, higher carbon concentrations such as a 2.5 mol ratio of carbon to uranium in the feed solutions failed to produce gel spheres with the proprietary carbon suspension. The kernels using our former baseline of Mogul L carbon black and Tamol SN were 90-92% of TD with full conversion of UC to UC2 at a variety of carbon levels. Raven 5000 carbon black and Tamol SN were used to produce 10 mol % UC2 kernels with 95% of TD. However, an increase in the Raven 5000 concentration led to a kernel density below 90% of TD. Raven 3500 carbon black and Tamol SN were used to make very dense kernels without complete conversion to UC2. The selection of the carbon black and dispersing agent is highly dependent on the desired final properties of the target kernels.

The Best Defense: Making Maximum Sense of Minimum Deterrence

DTIC Science & Technology

2011-06-01

uranium fuel cycles and has unmatched experience in the thorium fuel cycle.25 Published sources claim India produces between 20 and 40kg of plutonium...nuclear energy was moderate at best. Pakistan‘s first reactor , which it received from the United States, did not become operational until 1965.4...In 1974 Pakistan signed an agreement with France to supply a reprocessing plant for extracting plutonium from spent fuel from power reactors

A quasi-hexagonal prism-shaped carbon nitride for photoreduction of carbon dioxide under visible light.

PubMed

He, Zhiqiao; Wang, Danfen; Tang, Juntao; Song, Shuang; Chen, Jianmeng; Tao, Xinyong

2017-03-01

A quasi-hexagonal prism-shaped carbon nitride (H-C 3 N 4 ) was synthesized from urea-derived C 3 N 4 (U-C 3 N 4 ) using an alkaline hydrothermal process. U-C 3 N 4 decomposition followed by hydrogen bond rearrangement of hydrolyzed products leads to the formation of a quasi-hexagonal prism-shaped structure. The H-C 3 N 4 catalysts displayed superior activity in the photoreduction of CO 2 with H 2 O compared to U-C 3 N 4 . The enhanced photocatalytic activities can be attributed to the promotion of incompletely coordinated nitrogen atom formation in the C 3 N 4 molecules. Graphical abstract ᅟ.

Expensive Enrichment

ERIC Educational Resources Information Center

Resnikoff, Marvin

1975-01-01

This article presents an economic analysis of the nuclear fuel reprocessing industry. It indicates that while environmental safety devices have improved the working conditions, they have also added ever-increasing costs to this necessary process. (MA)

Spent Nuclear Fuel Disposition

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

Wagner, John C.

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNFmore » is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).« less

Spent Nuclear Fuel Disposition

DOE PAGES

Wagner, John C.

2016-05-22

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNFmore » is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).« less

Usefulness of rectally administering [1-(13)C]-butyrate for breath test in patients with active and quiescent ulcerative colitis.

PubMed

Kato, Kimitoshi; Ishii, Yukimoto; Mizuno, Shigeaki; Sugitani, Masahiko; Asai, Satoshi; Kohno, Tadashi; Takahashi, Katsuyuki; Komuro, Sachiko; Iwamoto, Maho; Miyamoto, Shunpachi; Takayama, Tadatoshi; Arakawa, Yasuyuki

2007-02-01

Impaired butyrate metabolism plays a part in ulcerative colitis (UC). To assess the usefulness of measuring butyrate metabolism as an indication of inflammatory activity, we investigated the rate of butyrate metabolism by breath test after administering [1-(13)C]-butyrate rectally to patients with UC. Thirty-eight UC patients (22 active, 16 quiescent) and 15 healthy controls were given [1-(13)C]-butyrate enemas. The (13)CO2 production rate was measured by breath test using an infrared spectrometric analyzer. The quantity of expired (13)CO2 was significantly lower in the active than in the quiescent UC and control groups. Cumulative (13)CO2 production at 240 min showed significant negative correlations with the clinical activity index (r=-0.65, p<0.0001), endoscopic activity index (r=-0.63, p=0.0001) and histology (r=-0.71, p<0.0001) in the active UC group. The (13)CO2 production rate was significantly increased in the quiescent stage as compared with the active stage in six UC patients, in whom clinical remission was achieved, in accordance with improvements in the clinical activity index, the endoscopic activity index, histology and fecal butyrate concentrations. Significant inverse correlations between the cumulative (13)CO2 production rate and these three parameters were seen in these six UC patients assessed in both the active and quiescent stages. Measurement of expired (13)CO2 after rectally administering [1-(13)C]-butyrate in active and quiescent UC appears to be a promising and reliable method for evaluating disease activity and metabolic changes associated with amelioration of inflammation.

Gaseous fuel reactors for power systems

NASA Technical Reports Server (NTRS)

Kendall, J. S.; Rodgers, R. J.

1977-01-01

Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

Experimental study on thermal hazard of tributyl phosphate-nitric acid mixtures using micro calorimeter technique.

PubMed

Sun, Qi; Jiang, Lin; Gong, Liang; Sun, Jin-Hua

2016-08-15

During PUREX spent nuclear fuel reprocessing, mixture of tributyl phosphate (TBP) and hydrocarbon solvent are employed as organic solvent to extract uranium in consideration of radiation contaminated safety and resource recycling, meanwhile nitric acid is utilized to dissolve the spent fuel into small pieces. However, once TBP contacts with nitric acid or nitrates above 130°C, a heavy "red oil" layer would occur accompanied by thermal runaway reactions, even caused several nuclear safety accident. Considering nitric acid volatility and weak exothermic detection, C80micro calorimeter technique was used in this study to investigate thermal decomposition of TBP mixed with nitric acid. Results show that the concentration of nitric acid greatly influences thermal hazard of the system by direct reactions. Even with a low heating rate, if the concentration of nitric acid increases due to evaporation of water or improper operations, thermal runaway in the closed system could start at a low temperature. Copyright © 2016 Elsevier B.V. All rights reserved.

Converting Maturing Nuclear Sites to Integrated Power Production Islands

DOE PAGES

Solbrig, Charles W.

2011-01-01

Nuclear islands, which are integrated power production sites, could effectively sequester and safeguard the US stockpile of plutonium. A nuclear island, an evolution of the integral fast reactor, utilizes all the Transuranics (Pu plus minor actinides) produced in power production, and it eliminates all spent fuel shipments to and from the site. This latter attribute requires that fuel reprocessing occur on each site and that fast reactors be built on-site to utilize the TRU. All commercial spent fuel shipments could be eliminated by converting all LWR nuclear power sites to nuclear islands. Existing LWR sites have the added advantage ofmore » already possessing a license to produce nuclear power. Each could contribute to an increase in the nuclear power production by adding one or more fast reactors. Both the TRU and the depleted uranium obtained in reprocessing would be used on-site for fast fuel manufacture. Only fission products would be shipped to a repository for storage. The nuclear island concept could be used to alleviate the strain of LWR plant sites currently approaching or exceeding their spent fuel pool storage capacity. Fast reactor breeding ratio could be designed to convert existing sites to all fast reactors, or keep the majority thermal.« less

Back-end of the fuel cycle - Indian scenario

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

Wattal, P.K.

Nuclear power has a key role in meeting the energy demands of India. This can be sustained by ensuring robust technology for the back end of the fuel cycle. Considering the modest indigenous resources of U and a huge Th reserve, India has adopted a three stage Nuclear Power Programme (NPP) based on 'closed fuel cycle' approach. This option on 'Recovery and Recycle' serves twin objectives of ensuring adequate supply of nuclear fuel and also reducing the long term radio-toxicity of the wastes. Reprocessing of the spent fuel by Purex process is currently employed. High Level Liquid Waste (HLW) generatedmore » during reprocessing is vitrified and undergoes interim storage. Back-end technologies are constantly modified to address waste volume minimization and radio-toxicity reduction. Long-term management of HLW in Indian context would involve partitioning of long lived minor actinides and recovery of valuable fission products specifically cesium. Recovery of minor actinides from HLW and its recycle is highly desirable for the sustained growth of India's NPPs. In this context, programme for developing and deploying partitioning technologies on industrial scale is pursued. The partitioned elements could be either transmuted in Fast Reactors (FRs)/Accelerated Driven Systems (ADS) as an integral part of sustainable Indian NPP. (authors)« less

129I in the oceans: origins and applications.

PubMed

Raisbeck, G M; Yiou, F

1999-09-30

The quantity of the long lived (half-life 15.7 million years) radioactive isotope 129I in the pre-nuclear age ocean was approximately 100 kg. Various nuclear related activities, including weapons testing, nuclear fuel reprocessing, Chernobyl and other authorized or non-authorized dumping of radioactive waste have increased the ocean inventory of 129I by more than one order of magnitude. The most important of these sources are the direct marine discharges from the commercial reprocessing facilities at La Hague (France) and Sellafield (UK) which have discharged approximately 1640 kg in the English Channel, and approximately 720 kg in the Irish Sea, respectively. We discuss how this 129I can be used as both a 'pathway' and 'transit time' tracer in the North Atlantic and Arctic oceans, as well as a parameter for distinguishing between reprocessed and non-reprocessed nuclear waste in the ocean, and as a proxy for the transport and dilution of other soluble pollutants input to the North Sea.

Modelling of radiation field around spent fuel container.

PubMed

Kryuchkov, E F; Opalovsky, V A; Tikhomirov, G V

2005-01-01

Operation of nuclear reactors leads to the production of spent nuclear fuel (SNF). There are two basic strategies of SNF management: ultimate disposal of SNF in geological formations and recycle or repeated utilisation of reprocessed SNF. In both options, there is an urgent necessity to study radiation properties of SNF. Information about SNF radiation properties is required at all stages of SNF management. In order to reach more effective utilisation of nuclear materials, new fuel cycles are under development based on uranium-plutonium, uranium-thorium and some other types of nuclear fuel. These promising types of nuclear fuel are characterised by quite different radiation properties at all the stages of nuclear fuel cycle (NFC) listed above. So, comparative analysis is required for radiation properties of different nuclear fuel types at different NFC stages. The results presented here were obtained from the numerical analysis of the radiation field around transport containers of different SNF types and in SNF storage. The calculations are carried out with the application of the computer code packages SCALE-4.3 and MCNP-4C. Comparison of the dose parameters obtained for different models of the transport container with experimental data allowed us to make certain conclusions about the errors of numerical results caused by the approximate geometrical description of the transport container.

Promises and Challenges of Thorium Implementation for Transuranic Transmutation - 13550

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

Franceschini, F.; Lahoda, E.; Wenner, M.

2013-07-01

This paper focuses on the challenges of implementing a thorium fuel cycle for recycle and transmutation of long-lived actinide components from used nuclear fuel. A multi-stage reactor system is proposed; the first stage consists of current UO{sub 2} once-through LWRs supplying transuranic isotopes that are continuously recycled and burned in second stage reactors in either a uranium (U) or thorium (Th) carrier. The second stage reactors considered for the analysis are Reduced Moderation Pressurized Water Reactors (RMPWRs), reconfigured from current PWR core designs, and Fast Reactors (FRs) with a burner core design. While both RMPWRs and FRs can in principlemore » be employed, each reactor and associated technology has pros and cons. FRs have unmatched flexibility and transmutation efficiency. RMPWRs have higher fuel manufacturing and reprocessing requirements, but may represent a cheaper solution and the opportunity for a shorter time to licensing and deployment. All options require substantial developments in manufacturing, due to the high radiation field, and reprocessing, due to the very high actinide recovery ratio to elicit the claimed radiotoxicity reduction. Th reduces the number of transmutation reactors, and is required to enable a viable RMPWR design, but presents additional challenges on manufacturing and reprocessing. The tradeoff between the various options does not make the choice obvious. Moreover, without an overarching supporting policy in place, the costly and challenging technologies required inherently discourage industrialization of any transmutation scheme, regardless of the adoption of U or Th. (authors)« less

Ketonization of levulinic acid and γ-valerolactone to hydrocarbon fuel precursors

DOE PAGES

Lilga, Michael A.; Padmaperuma, Asanga B.; Auberry, Deanna L.; ...

2017-06-21

We studied a new process for direct conversion of either levulinic acid (LA) or γ-valerolactone (GVL) to hydrocarbon fuel precursors. The process involves passing an aqueous solution of LA or GVL containing a reducing agent, such as ethylene glycol or formic acid, over a ketonization catalyst at 380–400 °C and atmospheric pressure to form a biphasic liquid product. The organic phase is significantly oligomerized and deoxygenated and comprises a complex mixture of open-chain alkanes and olefins, aromatics, and low concentrations of ketones, alcohols, ethers, and carboxylates or lactones. Carbon content in the aqueous phase decreases with decreasing feed rate; themore » aqueous phase can be reprocessed through the same catalyst to form additional organic oils to improve carbon yield. Catalysts are readily regenerated to restore initial activity. Furthermore, the process might be valuable in converting cellulosics to biorenewable gasoline, jet, and diesel fuels as a means to decrease petroleum use and decrease greenhouse gas emissions.« less

ORNL experience and perspectives related to processing of thorium and 233U for nuclear fuel

DOE PAGES

Croff, Allen G.; Collins, Emory D.; Del Cul, G. D.; ...

2016-05-01

Thorium-based nuclear fuel cycles have received renewed attention in both research and public circles since about the year 2000. Much of the attention has been focused on nuclear fission energy production that utilizes thorium as a fertile element for producing fissionable 233U for recycle in thermal reactors, fast reactors, or externally driven systems. Here, lesser attention has been paid to other fuel cycle operations that are necessary for implementation of a sustainable thorium-based fuel cycle such as reprocessing and fabrication of recycle fuels containing 233U.

Spent fuel data base: commercial light water reactors. [PWR; BWR

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

Hauf, M.J.; Kniazewycz, B.G.

1979-12-01

As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

ADVANCED COURSE ON FUEL ELEMENTS FOR WATER COOLED POWER REACTORS, ORGANIZED BY THE NETHERLANDS'-NORWEGIAN REACTOR SCHOOL AT INSTITUTT FOR ATOMENERGI, KJELLER, NORWAY, 22nd AUGUST-3rd SEPTEMBER,1960. VOLUME III

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

Aas, S.; Barendregt, T.J.; Chesne, A.

1960-07-01

A series of lectures on fuel elements for water-cooled power reactors are presented. Topics covered include fabrication, properties, cladding, radiation damage, design, cycling, storage and transpont, and reprocessing. Separate records have been prepared for each section.

Aqueous and pyrochemical reprocessing of actinide fuels

NASA Astrophysics Data System (ADS)

Toth, L. Mac; Bond, Walter D.; Avens, Larry R.

1993-02-01

Processing of the nuclear fuel actinides has developed in two independent directions—aqueous processing and pyroprocessing. Similarities in the two processes, their goals, and restraints are indicated in brief parallel descriptions along with distinguishing advantages and areas of future development. It is suggested that from a technical viewpoint, the ultimate process might be a hybrid which incorporates the best steps of each process.

Concentration of 129I in aquatic biota collected from a lake adjacent to the spent nuclear fuel reprocessing plant in Rokkasho, Japan.

PubMed

Ueda, Shinji; Kakiuchi, Hideki; Hasegawa, Hidenao; Kawamura, Hidehisa; Hisamatsu, Shun'ichi

2015-11-01

The spent nuclear fuel reprocessing plant in Rokkasho, Japan, has been undergoing final testing since March 2006. During April 2006-October 2008, that spent fuel was cut and chemically processed, the plant discharged (129)I into the atmosphere and coastal waters. To study (129)I behaviour in brackish Lake Obuchi, which is adjacent to the plant, (129)I concentrations in aquatic biota were measured by accelerator mass spectrometry. Owing to (129)I discharge from the plant, the (129)I concentration in the biota started to rise from the background concentration in 2006 and was high during 2007-08. The (129)I concentration has been rapidly decreasing after the fuel cutting and chemically processing were finished. The (129)I concentration factors in the biota were higher than those reported by IAEA for marine organisms and similar to those reported for freshwater biota. The estimated annual committed effective dose due to ingestion of foods with the maximum (129)I concentration in the biota samples was 2.8 nSv y(-1). © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Potential radiological impact of tornadoes on the safety of Nuclear Fuel Services' West Valley Fuel Reprocessing Plant. Volume I. Tornado effects on head-end cell airflow

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

Holloway, L.J.; Andrae, R.W.

1981-09-01

This report describes results of a parametric study of the impacts of a tornado-generated depressurization on airflow in the contaminated process cells within the presently inoperative Nuclear Fuel Services fuel reprocessing facility near West Valley, NY. The study involved the following tasks: (1) mathematical modeling of installed ventilation and abnormal exhaust pathways from the cells and prediction of tornado-induced airflows in these pathways; (2) mathematical modeling of individual cell flow characteristics and prediction of in-cell velocities induced by flows from step 1; and (3) evaluation of the results of steps 1 and 2 to determine whether any of the pathwaysmore » investigated have the potential for releasing quantities of radioactively contaminated air from the main process cells. The study has concluded that in the event of a tornado strike, certain pathways from the cells have the potential to release radioactive materials of the atmosphere. Determination of the quantities of radioactive material released from the cells through pathways identified in step 3 is presented in Part II of this report.« less

Direct disposal of spent fuel: developing solutions tailored to Japan

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

Kawamura, Hideki; McKinley, Ian G

2013-07-01

With the past Government policy of 100% reprocessing in Japan now open to discussion, options for direct disposal of spent fuel (SF) are now being considered in Japan. The need to move rapidly ahead in developing spent fuel management concepts is closely related to the ongoing debate on the future of nuclear power in Japan and the desire to understand the true costs of the entire life cycle of different options. Different scenarios for future nuclear power - and associated decisions on extent of reprocessing - will give rise to quite different inventories of SF with different disposal challenges. Althoughmore » much work has been carried out spent fuel disposal within other national programmes, the potential for mining the international knowledge base is limited by the boundary conditions for disposal in Japan. Indeed, with a volunteer approach to siting, no major salt deposits and few undisturbed sediments, high tectonic activity, relatively corrosive groundwater and no deserts, it is evident that a tailored solution is needed. Nevertheless, valuable lessons can be learned from projects carried out worldwide, if focus is placed on basic principles rather than implementation details. (authors)« less

Estimation of 85Kr dispersion from the spent nuclear fuel reprocessing plant in Rokkasho, Japan, using an atmospheric dispersion model.

PubMed

Abe, K; Iyogi, T; Kawabata, H; Chiang, J H; Suwa, H; Hisamatsu, S

2015-11-01

The spent nuclear fuel reprocessing plant of Japan Nuclear Fuel Limited (JNFL) located in Rokkasho, Japan, discharged small amounts of (85)Kr into the atmosphere during final tests of the plant with actual spent fuel from 31 March 2006 to October 2008. During this period, the gamma-ray dose rates due to discharged (85)Kr were higher than the background rates measured at the Institute for Environmental Sciences and at seven monitoring stations of the Aomori prefectural government and JNFL. The dispersion of (85)Kr was simulated by means of the fifth-generation Penn State/NCAR Mesoscale Model and the CG-MATHEW/ADPIC models (ver. 5.0) with a vertical terrain-following height coordinate. Although the simulated gamma-ray dose rates due to discharged (85)Kr agreed fairly well with measured rates, the agreement between the estimated monthly mean (85)Kr concentrations and the observed concentrations was poor. Improvement of the vertical flow of air may lead to better estimation of (85)Kr dispersion. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

[Infection frequency in patients with chronic idiopathic ulcerative colitis].

PubMed

Yamamoto-Furusho, J K; de León-Rendón, J L; Rodas, L

2012-01-01

Ulcerative Colitis (UC) is a chronic inflammatory bowel disease characterized by diffuse inflammation of the mucosa of the colon. Up to now, diverse observational studies have implicated a wide variety of pathogenic microorganisms as causal and exacerbating factors in UC. Clostridium difficile (C. difficile) infection has been associated with recurrence and treatment failure and its incidence in patients with UC has been on the rise in the last few years. To determine the frequency of infection by different microorganisms in Mexican UC patients. A total of 150 patients with definitive UC diagnosis were studied. All the stool tests for parasites and ova, stool cultures, tests for the C. difficile toxins A and B, and immunohistochemistry for Cytomegalovirus in colon segment biopsies were analyzed. Other demographic and clinical variables of the disease were recorded for their correlation with infection frequency. Infection frequency in UC patients was 28.00%. C. difficile infection was present in 0.013%. Other pathogens were found, such as Endolimax nana (9.00%), Entamoeba histolytica (3.00%), Cytomegalovirus (2.00%), Salmonella (2.00%), Shigella (0.70%), Toxoplasma gondii (0.70%) and Iodamoeba bütschlii (0.70%). Infection frequency was 28.00% in our study and C. difficile infection represented only 0.013%. Copyright © 2012 Asociación Mexicana de Gastroenterología. Published by Masson Doyma México S.A. All rights reserved.

Cost and resource utilization associated with use of computed tomography to evaluate chest pain in the emergency department: the Rule Out Myocardial Infarction using Computer Assisted Tomography (ROMICAT) study.

PubMed

Hulten, Edward; Goehler, Alexander; Bittencourt, Marcio Sommer; Bamberg, Fabian; Schlett, Christopher L; Truong, Quynh A; Nichols, John; Nasir, Khurram; Rogers, Ian S; Gazelle, Scott G; Nagurney, John T; Hoffmann, Udo; Blankstein, Ron

2013-09-01

Coronary computed tomographic angiography (cCTA) allows rapid, noninvasive exclusion of obstructive coronary artery disease (CAD). However, concern exists whether implementation of cCTA in the assessment of patients presenting to the emergency department with acute chest pain will lead to increased downstream testing and costs compared with alternative strategies. Our aim was to compare observed actual costs of usual care (UC) with projected costs of a strategy including early cCTA in the evaluation of patients with acute chest pain in the Rule Out Myocardial Infarction Using Computer Assisted Tomography I (ROMICAT I) study. We compared cost and hospital length of stay of UC observed among 368 patients enrolled in the ROMICAT I study with projected costs of management based on cCTA. Costs of UC were determined by an electronic cost accounting system. Notably, UC was not influenced by cCTA results because patients and caregivers were blinded to the cCTA results. Costs after early implementation of cCTA were estimated assuming changes in management based on cCTA findings of the presence and severity of CAD. Sensitivity analysis was used to test the influence of key variables on both outcomes and costs. We determined that in comparison with UC, cCTA-guided triage, whereby patients with no CAD are discharged, could reduce total hospital costs by 23% (P<0.001). However, when the prevalence of obstructive CAD increases, index hospitalization cost increases such that when the prevalence of ≥ 50% stenosis is >28% to 33%, the use of cCTA becomes more costly than UC. cCTA may be a cost-saving tool in acute chest pain populations that have a prevalence of potentially obstructive CAD <30%. However, increased cost would be anticipated in populations with higher prevalence of disease.

Chlorine-36 in the Snake River Plain Aquifer at the Idaho National Engineering Laboratory; origin and implications

USGS Publications Warehouse

Beasley, T.M.; Cecil, L.D.; Sharma, P.; Kubik, P.W.; Fehn, U.; Mann, L.J.; Gove, H.E.

1993-01-01

Between 1952 and 1984, low-level radioactive waste was introduced directly into the Snake River Plain aquifer at the Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These wastes were generated, principally, at the nuclear fuel reprocessing facility on the site. Our measurements of 36C1 in monitoring and production well waters, downgradient from disposal wells and seepage ponds, found easily detectable, nonhazardous concentrations of this radionuclide from the point of injection to the INEL southern site boundary. Comparisons are made between 3H and 36Cl concentrations in aquifer water and the advantages of 36C1 as a tracer of subsurface-water dynamics at the site are discussed.

Preparation of UC0.07-0.10N0.90-0.93 spheres for TRISO coated fuel particles

NASA Astrophysics Data System (ADS)

Hunt, R. D.; Silva, C. M.; Lindemer, T. B.; Johnson, J. A.; Collins, J. L.

2014-05-01

The US Department of Energy is considering a new nuclear fuel that would be less susceptible to ruptures during a loss-of-coolant accident. The fuel would consist of tristructural isotropic coated particles with dense uranium nitride (UN) kernels with diameters of 650 or 800 μm. The objectives of this effort are to make uranium oxide microspheres with adequately dispersed carbon nanoparticles and to convert these microspheres into UN spheres, which could be then sintered into kernels. Recent improvements to the internal gelation process were successfully applied to the production of uranium gel spheres with different concentrations of carbon black. After the spheres were washed and dried, a simple two-step heat profile was used to produce porous microspheres with a chemical composition of UC0.07-0.10N0.90-0.93. The first step involved heating the microspheres to 2023 K in a vacuum, and in the second step, the microspheres were held at 1873 K for 6 h in flowing nitrogen.

Developing a concept for a national used fuel interim storage facility in the United States

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

Lewis, Donald Wayne

2013-07-01

In the United States (U.S.) the nuclear waste issue has plagued the nuclear industry for decades. Originally, spent fuel was to be reprocessed but with the threat of nuclear proliferation, spent fuel reprocessing has been eliminated, at least for now. In 1983, the Nuclear Waste Policy Act of 1982 [1] was established, authorizing development of one or more spent fuel and high-level nuclear waste geological repositories and a consolidated national storage facility, called a 'Monitored Retrievable Storage' facility, that could store the spent nuclear fuel until it could be placed into the geological repository. Plans were under way to buildmore » a geological repository, Yucca Mountain, but with the decision by President Obama to terminate the development of Yucca Mountain, a consolidated national storage facility that can store spent fuel for an interim period until a new repository is established has become very important. Since reactor sites have not been able to wait for the government to come up with a storage or disposal location, spent fuel remains in wet or dry storage at each nuclear plant. The purpose of this paper is to present a concept developed to address the DOE's goals stated above. This concept was developed over the past few months by collaboration between the DOE and industry experts that have experience in designing spent nuclear fuel facilities. The paper examines the current spent fuel storage conditions at shutdown reactor sites, operating reactor sites, and the type of storage systems (transportable versus non-transportable, welded or bolted). The concept lays out the basis for a pilot storage facility to house spent fuel from shutdown reactor sites and then how the pilot facility can be enlarged to a larger full scale consolidated interim storage facility. (authors)« less

Conference Paper/Proceedings White Paper Conference Results of March 3, 2005 Workshop in Irvine, CA

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

Deborah Hart Redman; Sarah L. Catz

2005-03-31

A one-day workshop sponsored by UC Irvine's Center for Urban Infrastructure, bringing together 20 state departments of transportation and environmental quality to discuss national coordination on alternative fuels.

Evaluation of various carbon blacks and dispersing agents for use in the preparation of uranium microspheres with carbon

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

Hunt, Rodney Dale; Johnson, Jared A.; Collins, Jack Lee

A comparison study on carbon blacks and dispersing agents was performed to determine their impacts on the final properties of uranium fuel kernels with carbon. The main target compositions in this internal gelation study were 10 and 20 mol % uranium dicarbide (UC 2), which is UC 1.86, with the balance uranium dioxide. After heat treatment at 1900 K in flowing carbon monoxide in argon for 12 h, the density of the kernels produced using a X-energy proprietary carbon suspension, which is commercially available, ranged from 96% to 100% of theoretical density (TD), with full conversion of UC to UCmore » 2 at both carbon concentrations. However, higher carbon concentrations such as a 2.5 mol ratio of carbon to uranium in the feed solutions failed to produce gel spheres with the proprietary carbon suspension. The kernels using our former baseline of Mogul L carbon black and Tamol SN were 90–92% of TD with full conversion of UC to UC 2 at a variety of carbon levels. Raven 5000 carbon black and Tamol SN were used to produce 10 mol % UC2 kernels with 95% of TD. However, an increase in the Raven 5000 concentration led to a kernel density below 90% of TD. Raven 3500 carbon black and Tamol SN were used to make very dense kernels without complete conversion to UC 2. Lastly, the selection of the carbon black and dispersing agent is highly dependent on the desired final properties of the target kernels.« less

Evaluation of various carbon blacks and dispersing agents for use in the preparation of uranium microspheres with carbon

DOE PAGES

Hunt, Rodney Dale; Johnson, Jared A.; Collins, Jack Lee; ...

2017-10-12

A comparison study on carbon blacks and dispersing agents was performed to determine their impacts on the final properties of uranium fuel kernels with carbon. The main target compositions in this internal gelation study were 10 and 20 mol % uranium dicarbide (UC 2), which is UC 1.86, with the balance uranium dioxide. After heat treatment at 1900 K in flowing carbon monoxide in argon for 12 h, the density of the kernels produced using a X-energy proprietary carbon suspension, which is commercially available, ranged from 96% to 100% of theoretical density (TD), with full conversion of UC to UCmore » 2 at both carbon concentrations. However, higher carbon concentrations such as a 2.5 mol ratio of carbon to uranium in the feed solutions failed to produce gel spheres with the proprietary carbon suspension. The kernels using our former baseline of Mogul L carbon black and Tamol SN were 90–92% of TD with full conversion of UC to UC 2 at a variety of carbon levels. Raven 5000 carbon black and Tamol SN were used to produce 10 mol % UC2 kernels with 95% of TD. However, an increase in the Raven 5000 concentration led to a kernel density below 90% of TD. Raven 3500 carbon black and Tamol SN were used to make very dense kernels without complete conversion to UC 2. Lastly, the selection of the carbon black and dispersing agent is highly dependent on the desired final properties of the target kernels.« less

Method for cleaning solution used in nuclear fuel reprocessing

DOEpatents

Tallent, O.K.; Crouse, D.J.; Mailen, J.C.

1980-12-17

Nuclear fuel processing solution consisting of tri-n-butyl phosphate and dodecane, with a complex of uranium, plutonium, or zirconium and with a solvent degradation product such as di-n-butyl phosphate therein, is contacted with an aqueous solution of a salt formed from hydrazine and either a dicarboxylic acid or a hydroxycarboxylic acid, thereby removing the aforesaid complex from the processing solution.

The measurement of U(VI) and Np(IV) mass transfer in a single stage centrifugal contactor

NASA Astrophysics Data System (ADS)

May, I.; Birkett, E. J.; Denniss, I. S.; Gaubert, E. T.; Jobson, M.

2000-07-01

BNFL currently operates two reprocessing plants for the conversion of spent nuclear fuel into uranium and plutonium products for fabrication into uranium oxide and mixed uranium and plutonium oxide (MOX) fuels. To safeguard the future commercial viability of this process, BNFL is developing novel single cycle flowsheets that can be operated in conjunction with intensified centrifugal contactors.

Method for cleaning solution used in nuclear fuel reprocessing

DOEpatents

Tallent, Othar K.; Crouse, David J.; Mailen, James C.

1982-01-01

Nuclear fuel processing solution consisting of tri-n-butyl phosphate and dodecane, with a complex of uranium, plutonium, or zirconium and with a solvent degradation product such as di-n-butyl phosphate therein, is contacted with an aqueous solution of a salt formed from hydrazine and either a dicarboxylic acid or a hydroxycarboxylic acid, thereby removing the aforesaid complex from the processing solution.

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.

10 CFR Appendix B to Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

Code of Federal Regulations, 2010 CFR

2010-01-01

... undue risk to the health and safety of the public. This appendix establishes quality assurance...: reactor physics, stress, thermal, hydraulic, and accident analyses; compatibility of materials...

ONDRAF/NIRAS and high-level radioactive waste management in Belgium

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

Decamps, F.

1993-12-31

The National Agency for Radioactive Waste and Enriched Fissile Materials, ONDRAF/NIRAS, is a public body with legal personality in charge of managing all radioactive waste on Belgian territory, regardless of its origin and source. It is also entrusted with tasks related to the management of enriched fissile materials, plutonium containing materials and used or unused nuclear fuel, and with certain aspects of the dismantling of closed down nuclear facilities. High-level radioactive waste management comprises essentially and for the time being the storage of high-level liquid waste produced by the former EUROCHEMIC reprocessing plant and of high-level and very high-level heatmore » producing waste resulting from the reprocessing in France of Belgian spent fuel, as well as research and development (R and D) with regard to geological disposal in clay of this waste type.« less

Development of spent fuel reprocessing process based on selective sulfurization: Study on the Pu, Np and Am sulfurization

NASA Astrophysics Data System (ADS)

Kirishima, Akira; Amano, Yuuki; Nihei, Toshifumi; Mitsugashira, Toshiaki; Sato, Nobuaki

2010-03-01

For the recovery of fissile materials from spent nuclear fuel, we have proposed a novel reprocessing process based on selective sulfurization of fission products (FPs). The key concept of this process is utilization of unique chemical property of carbon disulfide (CS2), i.e., it works as a reductant for U3O8 but works as a sulfurizing agent for minor actinides and lanthanides. Sulfurized FPs and minor actinides (MA) are highly soluble to dilute nitric acid while UO2 and PuO2 are hardly soluble, therefore, FPs and MA can be removed from Uranium and Plutonium matrix by selective dissolution. As a feasibility study of this new concept, the sulfurization behaviours of U, Pu, Np, Am and Eu are investigated in this paper by the thermodynamical calculation, phase analysis of chemical analogue elements and tracer experiments.

Improvements to the MODIS Land Products in Collection Version 6

NASA Astrophysics Data System (ADS)

Wolfe, R. E.; Devadiga, S.; Masuoka, E. J.; Running, S. W.; Vermote, E.; Giglio, L.; Wan, Z.; Riggs, G. A.; Schaaf, C.; Myneni, R. B.; Friedl, M. A.; Wang, Z.; Sulla-menashe, D. J.; Zhao, M.

2013-12-01

The MODIS (Moderate Resolution Imaging Spectroradiometer) Adaptive Processing System (MODAPS), housed at the NASA Goddard Space Flight Center (GSFC), has been processing the earth view data acquired by the MODIS instrument aboard the Terra (EOS AM) and Aqua (EOS PM) satellites to generate suite of land and atmosphere data products using the science algorithms developed by the MODIS Science Team. These data products are used by diverse set of users in research and other applications from both government and non-government agencies around the world. These validated global products are also being used in interactive Earth system models able to predict global change accurately enough to assist policy makers in making sound decisions concerning the protection of our environment. Hence an increased emphasis is being placed on generation of high quality consistent data records from the MODIS data through reprocessing of the records using improved science algorithms. Since the launch of Terra in December 1999, MODIS land data records have been reprocessed four times. The Collection Version 6 (C6) reprocessing of MODIS Land and Atmosphere products is scheduled to start in Fall 2013 and is expected to complete in Spring 2014. This presentation will describe changes made to the C6 science algorithms to correct issues in the C5 products, additional improvements made to the products as deemed necessary by the data users and science teams, and new products introduced in this reprocessing. In addition to the improvements from product specific changes to algorithms, the C6 products will also see significant improvement in the calibration by the MODIS Calibration Science Team (MCST) of the C6 L1B Top of the Atmosphere (TOA) reflectance and radiance product, more accurate geolocation, and an improved Land Water mask. For the a priori land cover input, this reprocessing will use the multi-year land cover product generated with three years of MODIS data as input as opposed to one single land cover product used for the entire mission in the C5 reprocessing. The C6 products are expected to be released from the Distributed Active Archive Center (DAAC) soon after the reprocessing begins. To facilitate user acquaintance with products from the new version and independent evaluation of C6 by comparison of two versions, MODAPS plans to continue generation of products from both versions for at least a year after completion of the C6 reprocessing after which C5 processing will be discontinued.

Osteocalcin is not a strong determinant of serum testosterone and sperm count in men from infertile couples.

PubMed

Schwetz, V; Gumpold, R; Graupp, M; Hacker, N; Schweighofer, N; Trummer, O; Pieber, T R; Ballon, M; Lerchbaum, E; Obermayer-Pietsch, B

2013-07-01

Osteocalcin (OC) - released by osteoblasts and known as a marker of bone turnover - has been suggested to influence male fertility in murine models by enhancing testosterone production and sperm count. Results from clinical studies are scarce, however. The aim of this cross-sectional study was to investigate the proposed association of OC, undercarboxylated osteocalcin (ucOC) or carboxylated osteocalcin (cOC) with testosterone and sperm count in a cohort of 159 young male adults from infertile couples. Semen analysis was performed. Testosterone, free testosterone, LH, OC and ucOC were measured in serum samples after an overnight fast. cOC and OC correlated weakly but significantly with testosterone (OC: r = 0.165, p = 0.040, cOC: r = 0.193, p = 0.017), but not after adjusting for age and body mass index (BMI) or waist-hip ratio (WHR). %ucOC (ucOC levels expressed as percentage of total OC) correlated inversely with LH (r = -0.184, p = 0.023) and remained significant after the same adjustment. No significant correlations were observed between OC, cOC, ucOC, %ucOC and sperm count, semen volume and number of vital spermatozoa. In binary logistic regression analyses, none of the parameters of OC were predictors of oligozoospermia after adjusting for age and BMI or WHR. The weak association between %ucOC and LH has marginal clinical importance because of the lack of associations of parameters of OC with testosterone and sperm count. The current data thus cannot support the notion that OC is associated with male fertility in young men from infertile couples. © 2013 American Society of Andrology and European Academy of Andrology.

Maximizing the Efficiency of MAGTF Airlift Capacity in WestPac

DTIC Science & Technology

2013-03-27

respectively, cover the realm of medium-long range, medium lift capabilities. The UC -35 and the UC -12 aircraft, for short-medium range, light lift...requirements, are variations similar to the Cessna Citation and Beechcraft King Air respectively. In addition, the upgraded UC -12W model possesses an...airlift are resident to the VMGR and H&HS squadrons, specifically, the KC-130 and the OSA C- 12 and UC -35 aircraft, respectively. Each of these units

Chk1 Promotes DNA Damage Response Bypass following Oxidative Stress in a Model of Hydrogen Peroxide-Associated Ulcerative Colitis through JNK Inactivation and Chromatin Binding.

PubMed

Reissig, Kathrin; Silver, Andrew; Hartig, Roland; Schinlauer, Antje; Walluscheck, Diana; Guenther, Thomas; Siedentopf, Sandra; Ross, Jochen; Vo, Diep-Khanh; Roessner, Albert; Poehlmann-Nitsche, Angela

2017-01-01

Dysregulation of c-Jun N -terminal kinase (JNK) activation promoted DNA damage response bypass and tumorigenesis in our model of hydrogen peroxide-associated ulcerative colitis (UC) and in patients with quiescent UC (QUC), UC-related dysplasia, and UC-related carcinoma (UC-CRC), thereby adapting to oxidative stress. In the UC model, we have observed features of oncogenic transformation: increased proliferation, undetected DNA damage, and apoptosis resistance. Here, we show that Chk1 was downregulated but activated in the acute and quiescent chronic phases. In both phases, Chk1 was linked to DNA damage response bypass by suppressing JNK activation following oxidative stress, promoting cell cycle progression despite DNA damage. Simultaneously, activated Chk1 was bound to chromatin. This triggered histone acetylation and the binding of histone acetyltransferases and transcription factors to chromatin. Thus, chromatin-immobilized activated Chk1 executed a dual function by suppressing DNA damage response and simultaneously inducing chromatin modulation. This caused undetected DNA damage and increased cellular proliferation through failure to transmit the appropriate DNA damage signal. Findings in vitro were corroborated by chromatin accumulation of activated Chk1, Ac-H3, Ac-H4, and c-Jun in active UC (AUC) in vivo. Targeting chromatin-bound Chk1, GCN5, PCAF, and p300/CBP could be a novel therapeutic strategy to prevent UC-related tumor progression.

The polymorphism rs3024505 proximal to IL-10 is associated with risk of ulcerative colitis and Crohns disease in a Danish case-control study.

PubMed

Andersen, Vibeke; Ernst, Anja; Christensen, Jane; Østergaard, Mette; Jacobsen, Bent A; Tjønneland, Anne; Krarup, Henrik B; Vogel, Ulla

2010-05-28

Crohn's disease (CD) and ulcerative colitis (UC) are characterized by a dysregulated inflammatory response to normal constituents of the intestinal flora in the genetically predisposed host. Heme oxygenase-1 (HO-1/HMOX1) is a powerful anti-inflammatory and anti-oxidant enzyme, whereas the pro-inflammatory interleukin 1 beta (IL-1 beta/IL1B) and anti-inflammatory interleukin 10 (IL-10/IL10) are key modulators for the initiation and maintenance of inflammation. We investigated whether single nucleotide polymorphisms (SNPs) in the IL-1 beta, IL-10, and HO-1 genes, together with smoking, were associated with risk of CD and UC. Allele frequencies of the IL-1 beta T-31C (rs1143627), and IL-10 rs3024505, G-1082A (rs1800896), C-819T (rs1800871), and C-592A (rs1800872) and HO-1 A-413T (rs2071746) SNPs were assessed using a case-control design in a Danish cohort of 336 CD and 498 UC patients and 779 healthy controls. Odds ratio (OR) and 95% confidence interval (95% CI) were estimated by logistic regression models. Carriers of rs3024505, a marker polymorphism flanking the IL-10 gene, were at increased risk of CD (OR = 1.40, 95% CI: 1.06-1.85, P = 0.02) and UC (OR = 1.43, 95% CI: 1.12-1.82, P = 0.004) and, furthermore, with risk of a diagnosis of CD and UC at young age (OR = 1.47, 95% CI: 1.10-1.96) and OR = 1.35, 95% CI: 1.04-1.76), respectively). No association was found between the IL-1 beta, IL-10 G-1082A, C-819T, C-592A, and HO-1 gene polymorphisms and CD or UC. No consistent interactions between smoking status and CD or UC genotypes were demonstrated. The rs3024505 marker polymorphism flanking the IL-10 gene was significantly associated with risk of UC and CD, whereas no association was found between IL-1 beta or HO-1 gene polymorphisms and risk of CD and UC in this Danish study, suggesting that IL-10, but not IL-1 beta or HO-1, has a role in IBD etiology in this population.

The polymorphism rs3024505 proximal to IL-10 is associated with risk of ulcerative colitis and Crohns disease in a Danish case-control study

PubMed Central

2010-01-01

Background Crohns disease (CD) and ulcerative colitis (UC) are characterized by a dysregulated inflammatory response to normal constituents of the intestinal flora in the genetically predisposed host. Heme oxygenase-1 (HO-1/HMOX1) is a powerful anti-inflammatory and anti-oxidant enzyme, whereas the pro-inflammatory interleukin 1β (IL-1β/IL1B) and anti-inflammatory interleukin 10 (IL-10/IL10) are key modulators for the initiation and maintenance of inflammation. We investigated whether single nucleotide polymorphisms (SNPs) in the IL-1β, IL-10, and HO-1 genes, together with smoking, were associated with risk of CD and UC. Methods Allele frequencies of the IL-1β T-31C (rs1143627), and IL-10 rs3024505, G-1082A (rs1800896), C-819T (rs1800871), and C-592A (rs1800872) and HO-1 A-413T (rs2071746) SNPs were assessed using a case-control design in a Danish cohort of 336 CD and 498 UC patients and 779 healthy controls. Odds ratio (OR) and 95% confidence interval (95% CI) were estimated by logistic regression models. Results Carriers of rs3024505, a marker polymorphism flanking the IL-10 gene, were at increased risk of CD (OR = 1.40, 95% CI: 1.06-1.85, P = 0.02) and UC (OR = 1.43, 95% CI: 1.12-1.82, P = 0.004) and, furthermore, with risk of a diagnosis of CD and UC at young age (OR = 1.47, 95% CI: 1.10-1.96) and OR = 1.35, 95% CI: 1.04-1.76), respectively). No association was found between the IL-1β, IL-10 G-1082A, C-819T, C-592A, and HO-1 gene polymorphisms and CD or UC. No consistent interactions between smoking status and CD or UC genotypes were demonstrated. Conclusions The rs3024505 marker polymorphism flanking the IL-10 gene was significantly associated with risk of UC and CD, whereas no association was found between IL-1β or HO-1 gene polymorphisms and risk of CD and UC in this Danish study, suggesting that IL-10, but not IL-1β or HO-1, has a role in IBD etiology in this population. PMID:20509889

Ulcerative colitis in a multiracial Asian country: Racial differences and clinical presentation among Malaysian patients

PubMed Central

Tan, Yan-Mei; Goh, Khean-Lee

2005-01-01

AIM: TO determine the prevalence of ulcerative colitis (UC) in Malaysian patients and to establish the spectrum of the disease seen in Malaysian patients. METHODS: Data were obtained retrospectively from a review of the medical records of in- and out-patients with a diagnosis of UC at the University Hospital, Kuala Lumpur between 1985 and 1998. RESULTS: There were 45 confirmed cases of UC, of which 3 were foreigners, who were excluded from analysis. Thirty new cases of UC were diagnosed during the study period. Their mean age at presentation was 33.0 ± 10.0 years. The highest prevalence of UC was 17.9/100 000 hospital admissions in the Indians, followed by 11.2/100 000 hospital admissions in the Chinese. The lowest prevalence was 3.7/100 000 hospital admissions in the Malays. The prevalence of UC was significantly higher in the Indians and the Chinese when compared with the Malays with an OR of 4.89 (CI = 2.02-12.24; c2 = 15.45,P < 0.001) and 3.06 (CI = 1.24-7.78; c2 = 6.30; P = 0.012) respectively. The extent of colonic disease was similar in the Malay and Indian patients. In contrast, distal or left-sided colitis predominated in the Chinese with an OR of 8.17 (95%CI = 1.31-64.87; c2 = 5.53, P = 0.02). Extraintestinal manifestations were uncommon (11.9%). CONCLUSION: UC is an uncommon disease in Malaysia, but racial differences exist. The Indians had the highest prevalence of UC with the Chinese demonstrating the least extensive disease. PMID:16270398

XRCC1 Arg194Trp and Arg399Gln polymorphisms and arsenic methylation capacity are associated with urothelial carcinoma

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

Chiang, Chien-I; Huang, Ya-Li; Chen, Wei-Jen

The association between DNA repair gene polymorphisms and bladder cancer has been widely studied. However, few studies have examined the correlation between urothelial carcinoma (UC) and arsenic or its metabolites. The aim of this study was to examine the association between polymorphisms of the DNA repair genes, XRCC1 Arg194Trp, XRCC1 Arg399Gln, XRCC3 Thr241Met, and XPD Lys751Gln, with urinary arsenic profiles and UC. To this end, we conducted a hospital-based case–control study with 324 UC patients and 647 age- and gender-matched non-cancer controls. Genomic DNA was used to examine the genotype of XRCC1 Arg194Trp, XRCC1 Arg399Gln, XRCC3 Thr241Met, and XPD Lys751Glnmore » by PCR-restriction fragment length polymorphism analysis (PCR-RFLP). Urinary arsenic profiles were measured by high performance liquid chromatography (HPLC) linked with hydride generator and atomic absorption spectrometry. The XRCC1 399 Gln/Gln and 194 Arg/Trp and Trp/Trp genotypes were significantly related to UC, and the odds ratio (OR) and 95% confidence interval (95%CI) were 1.68 (1.03–2.75) and 0.66 (0.48–0.90), respectively. Participants with higher total urinary arsenic levels, a higher percentage of inorganic arsenic (InAs%) and a lower percentage of dimethylarsinic acid (DMA%) had a higher OR of UC. Participants carrying XRCC1 risk diplotypes G-C/G-C, A-C/A-C, and A-T/G-T, and who had higher total arsenic levels, higher InAs%, or lower DMA% compared to those with other XRCC1 diplotypes had a higher OR of UC. Our results suggest that the XRCC1 399 Gln/Gln and 194 Arg/Arg DNA repair genes play an important role in poor arsenic methylation capacity, thereby increasing the risk of UC in non-obvious arsenic exposure areas. - Highlights: • The XRCC1 399Gln/Gln genotype was significantly associated with increased OR of UC. • The XRCC1 194 Arg/Trp and Trp/Trp genotype had a significantly decreased OR of UC. • Combined effect of the XRCC1 genotypes and poor arsenic methylation capacity on UC.« less

Toll-like receptor 6 and connective tissue growth factor are significantly upregulated in mitomycin-C-treated urothelial carcinoma cells under hydrostatic pressure stimulation.

PubMed

Chen, Shao-Kuan; Chung, Chih-Ang; Cheng, Yu-Che; Huang, Chi-Jung; Chen, Wen-Yih; Ruaan, Ruoh-Chyu; Li, Chuan; Tsao, Chia-Wen; Hu, Wei-Wen; Chien, Chih-Cheng

2014-06-01

Urothelial carcinoma (UC) is the most common histologic subtype of bladder cancer. The administration of mitomycin C (MMC) into the bladder after transurethral resection of the bladder tumor (TURBT) is a common treatment strategy for preventing recurrence after surgery. We previously applied hydrostatic pressure combined with MMC in UC cells and found that hydrostatic pressure synergistically enhanced MMC-induced UC cell apoptosis through the Fas/FasL pathways. To understand the alteration of gene expressions in UC cells caused by hydrostatic pressure and MMC, oligonucleotide microarray was used to explore all the differentially expressed genes. After bioinformatics analysis and gene annotation, Toll-like receptor 6 (TLR6) and connective tissue growth factor (CTGF) showed significant upregulation among altered genes, and their gene and protein expressions with each treatment of UC cells were validated by quantitative real-time PCR and immunoblotting. Under treatment with MMC and hydrostatic pressure, UC cells showed increasing apoptosis using extrinsic pathways through upregulation of TLR6 and CTGF.

Synthesis and properties of reprocessable sulfonated polyimides cross-linked via acid stimulation for use as proton exchange membranes

NASA Astrophysics Data System (ADS)

Zhang, Boping; Ni, Jiangpeng; Xiang, Xiongzhi; Wang, Lei; Chen, Yongming

2017-01-01

Cross-linked sulfonated polyimides are one of the most promising materials for proton exchange membrane (PEM) applications. However, these cross-linked membranes are difficult to reprocess because they are insoluble. In this study, a series of cross-linkable sulfonated polyimides with flexible pendant alkyl side chains containing trimethoxysilyl groups is successfully synthesized. The cross-linkable polymers are highly soluble in common solvents and can be used to prepare tough and smooth films. Before the cross-linking reaction is complete, the membranes can be reprocessed, and the recovery rate of the prepared films falls within an acceptable range. The cross-linked membranes are obtained rapidly when the cross-linkable membranes are immersed in an acid solution, yielding a cross-linking density of the gel fraction of greater than 90%. The cross-linked membranes exhibit high proton conductivities and tensile strengths under hydrous conditions. Compared with those of pristine membranes, the oxidative and hydrolytic stabilities of the cross-linked membranes are significantly higher. The CSPI-70 membrane shows considerable power density in a direct methanol fuel cell (DMFC) test. All of these results suggest that the prepared cross-linked membranes have great potential for applications in proton exchange membrane fuel cells.

The use of nuclear data in the field of nuclear fuel recycling

NASA Astrophysics Data System (ADS)

Martin, Julie-Fiona; Launay, Agnès; Grassi, Gabriele; Binet, Christophe; Lelandais, Jacques; Lecampion, Erick

2017-09-01

AREVA NC La Hague facility is the first step of the nuclear fuel recycling process implemented in France. The processing of the used fuel is governed by high standards of criticality-safety, and strong expectations on the quality of end-products. From the received used fuel assemblies, the plutonium and the uranium are extracted for further energy production purposes within the years following the reprocessing. Furthermore, the ultimate waste - fission products and minor actinides on the one hand, and hulls and end-pieces on the other hand - is adequately packaged for long term disposal. The used fuel is therefore separated into very different materials, and time scales which come into account may be longer than in some other nuclear fields of activity. Given the variety of the handled nuclear materials, as well as the time scales at stake, the importance given to some radionuclides, and hence to the associated nuclear data, can also be specific to the AREVA NC La Hague plant. A study has thus been led to identify a list of the most important radionuclides for the AREVA NC La Hague plant applications, relying on the running constraints of the facility, and the end-products expectations. The activities at the AREVA NC La Hague plant are presented, and the methodology to extract the most important radionuclides for the reprocessing process is detailed.

The myth of the ``proliferation-resistant'' closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Lyman, Edwin S.

2000-07-01

National nuclear energy programs that engage in reprocessing of spent nuclear fuel (SNF) and the development of "closed" nuclear fuel cycles based on the utilization of plutonium process and store large quantities of weapons-usable nuclear materials in forms vulnerable to diversion or theft by national or subnational groups. Proliferation resistance, an idea dating back at least as far as the International Fuel Cycle Evaluation (INFCE) of the late 1970s, is a loosely defined term referring to processes for chemical separation of SNF that do not extract weapons-usable materials in a purified form.

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

Lilga, Michael A.; Padmaperuma, Asanga B.; Auberry, Deanna L.

We studied a new process for direct conversion of either levulinic acid (LA) or γ-valerolactone (GVL) to hydrocarbon fuel precursors. The process involves passing an aqueous solution of LA or GVL containing a reducing agent, such as ethylene glycol or formic acid, over a ketonization catalyst at 380–400 °C and atmospheric pressure to form a biphasic liquid product. The organic phase is significantly oligomerized and deoxygenated and comprises a complex mixture of open-chain alkanes and olefins, aromatics, and low concentrations of ketones, alcohols, ethers, and carboxylates or lactones. Carbon content in the aqueous phase decreases with decreasing feed rate; themore » aqueous phase can be reprocessed through the same catalyst to form additional organic oils to improve carbon yield. Catalysts are readily regenerated to restore initial activity. Furthermore, the process might be valuable in converting cellulosics to biorenewable gasoline, jet, and diesel fuels as a means to decrease petroleum use and decrease greenhouse gas emissions.« less

78 FR 7816 - Quality Assurance Program Requirements (Operations)

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-04

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0021] Quality Assurance Program Requirements (Operations...), DG-1300, ``Quality Assurance Program Requirements (Operations).'' DATES: Submit comments by April 1... CFR Part 50, Appendix B, ``Quality Assurance Criteria for Nuclear power Plants and Fuel Reprocessing...

CONSTRUCTION VIEW OF MAIN PROCESSING BUILDING (CPP601) LOOKING NORTHWEST. INL ...

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

CONSTRUCTION VIEW OF MAIN PROCESSING BUILDING (CPP-601) LOOKING NORTHWEST. INL PHOTO NUMBER NRTS-51-1390. Unknown Photographer, 1/31/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO OF REMOTE ANALYTICAL FACILITY (CPP627). INL PHOTO ...

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

CONSTRUCTION PROGRESS PHOTO OF REMOTE ANALYTICAL FACILITY (CPP-627). INL PHOTO NUMBER NRTS-54-12124. Unknown Photographer, 9/21/1954 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

DETAILS OF REMOTE ANALYTICAL FACILITY (CPP627). INL DRAWING NUMBER 200062700098105071. ...

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

DETAILS OF REMOTE ANALYTICAL FACILITY (CPP-627). INL DRAWING NUMBER 200-0627-00-098-105071. ALTERNATE ID NUMBER 4272-14-108. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION VIEW OF MAIN PROCESSING BUILDING (CPP601) LOOKING EAST. INL ...

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

CONSTRUCTION VIEW OF MAIN PROCESSING BUILDING (CPP-601) LOOKING EAST. INL PHOTO NUMBER NRTS-51-1547. Unknown Photographer, 2/28/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

The potential of volatile organic compounds for the detection of active disease in patients with ulcerative colitis.

PubMed

Smolinska, A; Bodelier, A G L; Dallinga, J W; Masclee, A A M; Jonkers, D M; van Schooten, F-J; Pierik, M J

2017-05-01

To optimise treatment of ulcerative colitis (UC), patients need repeated assessment of mucosal inflammation. Current non-invasive biomarkers and clinical activity indices do not accurately reflect disease activity in all patients and cannot discriminate UC from non-UC colitis. Volatile organic compounds (VOCs) in exhaled air could be predictive of active disease or remission in Crohn's disease. To investigate whether VOCs are able to differentiate between active UC, UC in remission and non-UC colitis. UC patients participated in a 1-year study. Clinical activity index, blood, faecal and breath samples were collected at each out-patient visit. Patients with clear defined active faecal calprotectin >250 μg/g and inactive disease (Simple Clinical Colitis Activity Index <3, C-reactive protein <5 mg/L and faecal calprotectin <100 μg/g) were included for cross-sectional analysis. Non-UC colitis was confirmed by stool culture or radiological evaluation. Breath samples were analysed by gas chromatography time-of-flight mass spectrometry and kernel-based method to identify discriminating VOCs. In total, 72 UC (132 breath samples; 62 active; 70 remission) and 22 non-UC-colitis patients (22 samples) were included. Eleven VOCs predicted active vs. inactive UC in an independent internal validation set with 92% sensitivity and 77% specificity (AUC 0.94). Non-UC colitis patients could be clearly separated from active and inactive UC patients with principal component analysis. Volatile organic compounds can accurately distinguish active disease from remission in UC and profiles in UC are clearly different from profiles in non-UC colitis patients. VOCs have demonstrated potential as new non-invasive biomarker to monitor inflammation in UC. © 2017 John Wiley & Sons Ltd.

Endoscopic retrograde cholangiopancreatography-associated AmpC Escherichia coli outbreak.

PubMed

Wendorf, Kristen A; Kay, Meagan; Baliga, Christopher; Weissman, Scott J; Gluck, Michael; Verma, Punam; D'Angeli, Marisa; Swoveland, Jennifer; Kang, Mi-Gyeong; Eckmann, Kaye; Ross, Andrew S; Duchin, Jeffrey

2015-06-01

We identified an outbreak of AmpC-producing Escherichia coli infections resistant to third-generation cephalosporins and carbapenems (CR) among 7 patients who had undergone endoscopic retrograde cholangiopancreatography at hospital A during November 2012-August 2013. Gene sequencing revealed a shared novel mutation in a bla CMY gene and a distinctive fumC/ fimH typing profile. To determine the extent and epidemiologic characteristics of the outbreak, identify potential sources of transmission, design and implement infection control measures, and determine the association between the CR E. coli and AmpC E. coli circulating at hospital A. We reviewed laboratory, medical, and endoscopy reports, and endoscope reprocessing procedures. We obtained cultures from endoscopes after reprocessing as well as environmental samples and conducted pulsed-field gel electrophoresis and gene sequencing on phenotypic AmpC isolates from patients and endoscopes. Cases were those infected with phenotypic AmpC isolates (both carbapenem-susceptible and CR) and identical bla CMY-2, fumC, and fimH alleles or related pulsed-field gel electrophoresis patterns. Thirty-five of 49 AmpC E. coli tested met the case definition, including all CR isolates. All cases had complicated biliary disease and had undergone at least 1 endoscopic retrograde cholangiopancreatography at hospital A. Mortality at 30 days was 16% for all patients and 56% for CR patients. Two of 8 reprocessed endoscopic retrograde cholangiopancreatography scopes harbored AmpC that matched case isolates by pulsed-field gel electrophoresis. Environmental cultures were negative. No breaches in infection control were identified. Endoscopic reprocessing exceeded manufacturer's recommended cleaning guidelines. Recommended reprocessing guidelines are not sufficient.

Fusion Applications and Market Evaluation (FAME) Study

DTIC Science & Technology

1988-02-01

fuel from the breeder. Pyrochemical reprocessing is identified as having the potential for low cost, but needs development . The fast-fission designs... Development Administration, "Alternatives for Man- aging Wastes from Reactors and Post-Fission Operations in the LWR Fuel Cycle," ERDA-76-43 (1976). 5...of the ICF program to produce pulsed radiation for military development applications. X-rays can be converted into UV at about 50% energy efficiency

U.S. and South Korean Cooperation in the World Nuclear Energy Market: Major Policy Considerations

DTIC Science & Technology

2010-01-21

a laboratory-scale research program on reprocessing spent fuel with an advanced pyroprocessing technique. However, the level of consensus over the... pyroprocessing option among government agencies, Korean electric utilities, and the public remains uncertain. The current U.S.-Korea 123 agreement...permission. KAERI’s pyroprocessing technology would partially separate plutonium and uranium from spent fuel, but the United States has not allowed the

JPRS Report, Science & Technology, Japan

DTIC Science & Technology

1987-11-12

Change (4) Future Direction Anyway, it has become almost clear that the effect of power recovery cannot be expected from the insulation of...process spent fuels in greater safety and to recover the uranium or plutonium from spent fuels for effective reapplication. In 1974, the PNC began...constructed to serve as a pilot plant that could be used to establish reprocessing technology for the next practical stage. 32 As for enriched uranium

NUCLEAR MATERIAL ATTRACTIVENESS: AN ASSESSMENT OF MATERIAL ASSOCIATED WITH A CLOSED FUEL CYCLE

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

Bathke, C. G.; Ebbinghaus, B.; Sleaford, Brad W.

2010-06-11

This paper examines the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with the various processing steps required for a closed fuel cycle. This paper combines the results from earlier studies that examined the attractiveness of SNM associated with the processing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR with new results for the final, repeated burning of SNM in fast-spectrum reactors: fast reactors and accelerator driven systems (ADS). The results of this paper suggest that all reprocessing products evaluated so farmore » need to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of "attractiveness levels" that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities. The methodology and key findings will be presented. Additionally, how these attractiveness levels relate to proliferation resistance (e.g. by increasing impediments to the diversion, theft, or undeclared production of SNM for the purpose of acquiring a nuclear weapon), and how they could be used to help inform policy makers, will be discussed.« less

The impact of Clostridum difficile on surgical rate among ulcerative colitis patients: A systemic review and meta-analysis.

PubMed

Peng, Jiang-Chen; Shen, Jun; Zhu, Qi; Ran, Zhi-Hua

2015-01-01

There is growing recognition of the impact of Clostridum difficile infection (CDI) on patients with inflammatory bowel disease. Clostridium difficile infection causes greater morbidity and mortality. This study aimed to evaluate the impact of C. difficile on surgical risk among ulcerative colitis (UC) patients. We searched the following databases: MEDLINE, EMBASE, the Cochrane Central Register of Controlled Trials, ACP Journal Club, DARE, CMR, and HTA. Studies were included if fulfilled the following criteria: (1) Cohort or case-control studies, which involved a comparison group that lacked CDI, (2) Patients were given a primary diagnosis of UC, (3) Comorbidity of CDI was evaluated by enzyme immunoassay of stool for C. difficile toxin A and B or C. difficile stool culture, (4) Studies evaluated surgical rate, and (5) Studies reported an estimate of odds ratio, accompanied by a corresponding measure of uncertainty. Five studies with 2380 patients fulfilled the inclusion criteria. Overall, meta-analysis showed that UC with CDI patients had a significant higher surgical rate than patients with UC alone. (OR=1.76, 95% CI=1.36-2.28). C. difficile infection increased the surgical rate in UC patients. However, results should be interpreted with caution, given the limitations of this stud.

Hydrostatic pressure enhances mitomycin C induced apoptosis in urothelial carcinoma cells.

PubMed

Chen, Shao-Kuan; Chung, Chih-Ang; Cheng, Yu-Che; Huang, Chi-Jung; Ruaan, Ruoh-Chyu; Chen, Wen-Yih; Li, Chuan; Tsao, Chia-Wen; Hu, Wei-Wen; Chien, Chih-Cheng

2014-01-01

Urothelial carcinoma (UC) of the bladder is the second most common cancer of the genitourinary system. Clinical UC treatment usually involves transurethral resection of the bladder tumor followed by adjuvant intravesical immunotherapy or chemotherapy to prevent recurrence. Intravesical chemotherapy induces fewer side effects than immunotherapy but is less effective at preventing tumor recurrence. Improvement to intravesical chemotherapy is, therefore, needed. Cellular effects of mitomycin C (MMC) and hydrostatic pressure on UC BFTC905 cells were assessed. The viability of the UC cells was determined using cellular proliferation assay. Changes in apoptotic function were evaluated by caspase 3/7 activities, expression of FasL, and loss of mitochondrial membrane potential. Reduced cell viability was associated with increasing hydrostatic pressure. Caspase 3/7 activities were increased following treatment of the UC cells with MMC or hydrostatic pressure. In combination with 10 kPa hydrostatic pressure, MMC treatment induced increasing FasL expression. The mitochondria of UC cells displayed increasingly impaired membrane potentials following a combined treatment with 10 μg/ml MMC and 10 kPa hydrostatic pressure. Both MMC and hydrostatic pressure can induce apoptosis in UC cells through an extrinsic pathway. Hydrostatic pressure specifically increases MMC-induced apoptosis and might minimize the side effects of the chemotherapy by reducing the concentration of the chemical agent. This study provides a new and alternative approach for treatment of patients with UC following transurethral resection of the bladder tumor. Copyright © 2014 Elsevier Inc. All rights reserved.

Chk1 Promotes DNA Damage Response Bypass following Oxidative Stress in a Model of Hydrogen Peroxide-Associated Ulcerative Colitis through JNK Inactivation and Chromatin Binding

PubMed Central

Silver, Andrew; Guenther, Thomas; Siedentopf, Sandra; Ross, Jochen; Vo, Diep-Khanh; Roessner, Albert

2017-01-01

Dysregulation of c-Jun N-terminal kinase (JNK) activation promoted DNA damage response bypass and tumorigenesis in our model of hydrogen peroxide-associated ulcerative colitis (UC) and in patients with quiescent UC (QUC), UC-related dysplasia, and UC-related carcinoma (UC-CRC), thereby adapting to oxidative stress. In the UC model, we have observed features of oncogenic transformation: increased proliferation, undetected DNA damage, and apoptosis resistance. Here, we show that Chk1 was downregulated but activated in the acute and quiescent chronic phases. In both phases, Chk1 was linked to DNA damage response bypass by suppressing JNK activation following oxidative stress, promoting cell cycle progression despite DNA damage. Simultaneously, activated Chk1 was bound to chromatin. This triggered histone acetylation and the binding of histone acetyltransferases and transcription factors to chromatin. Thus, chromatin-immobilized activated Chk1 executed a dual function by suppressing DNA damage response and simultaneously inducing chromatin modulation. This caused undetected DNA damage and increased cellular proliferation through failure to transmit the appropriate DNA damage signal. Findings in vitro were corroborated by chromatin accumulation of activated Chk1, Ac-H3, Ac-H4, and c-Jun in active UC (AUC) in vivo. Targeting chromatin-bound Chk1, GCN5, PCAF, and p300/CBP could be a novel therapeutic strategy to prevent UC-related tumor progression. PMID:28751935

CONSTRUCTION PROGRESS PHOTO SHOWING EXCAVATION PIT FOR MAIN PROCESSING BUILDING ...

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

CONSTRUCTION PROGRESS PHOTO SHOWING EXCAVATION PIT FOR MAIN PROCESSING BUILDING (CPP-601) LOOKING SOUTH. INL PHOTO NUMBER NRTS-50-693. Unknown Photographer, 1950 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

MISCELLANEOUS ARCHITECTURAL DETAILS OF REMOTE ANALYTICAL FACILITY (CPP627). INL DRAWING ...

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

MISCELLANEOUS ARCHITECTURAL DETAILS OF REMOTE ANALYTICAL FACILITY (CPP-627). INL DRAWING NUMBER 200-0627-00-098-105631. ALTERNATE ID NUMBER 4272-814-134. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO SHOWING MAIN PROCESSING BUILDING (CPP601) LOOKING NORTH. ...

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

CONSTRUCTION PROGRESS PHOTO SHOWING MAIN PROCESSING BUILDING (CPP-601) LOOKING NORTH. INL PHOTO NUMBER NRTS-51-1387. Unknown Photographer, 1/31/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

SOUTH ELEVATION OF HOT PILOT PLANT (CPP640) LOOKING NORTH. INL ...

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

SOUTH ELEVATION OF HOT PILOT PLANT (CPP-640) LOOKING NORTH. INL PHOTO NUMBER HD-22-3-1. Mike Crane, Photographer, 11/1998 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO OF REMOTE ANALYTICAL FACILITY (CPP627). INL PHOTO ...

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

CONSTRUCTION PROGRESS PHOTO OF REMOTE ANALYTICAL FACILITY (CPP-627). INL PHOTO NUMBER NRTS-54-12573. R.G. Larsen, Photographer, 10/20/1954 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO OF REMOTE ANALYTICAL FACILITY (CPP627) SHOWING INITIAL ...

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

CONSTRUCTION PROGRESS PHOTO OF REMOTE ANALYTICAL FACILITY (CPP-627) SHOWING INITIAL EXCAVATION. INL PHOTO NUMBER NRTS-54-10703. Unknown Photographer, 5/21/1954 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

SOUTH ELEVATION OF MAIN PROCESSING BUILDING (CPP601) LOOKING NORTH. INL ...

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

SOUTH ELEVATION OF MAIN PROCESSING BUILDING (CPP-601) LOOKING NORTH. INL PHOTO NUMBER HD-22-5-3. Mike Crane, Photographer, 11/1998 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Nitrogen Trifluoride-Based Fluoride- Volatility Separations Process: Initial Studies

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

McNamara, Bruce K.; Scheele, Randall D.; Casella, Andrew M.

2011-09-28

This document describes the results of our investigations on the potential use of nitrogen trifluoride as the fluorinating and oxidizing agent in fluoride volatility-based used nuclear fuel reprocessing. The conceptual process uses differences in reaction temperatures between nitrogen trifluoride and fuel constituents that produce volatile fluorides to achieve separations and recover valuable constituents. We provide results from our thermodynamic evaluations, thermo-analytical experiments, kinetic models, and provide a preliminary process flowsheet. The evaluations found that nitrogen trifluoride can effectively produce volatile fluorides at different temperatures dependent on the fuel constituent.

Florid urticarial vasculitis heralding a flare up of ulcerative colitis.

PubMed

Boules, Evon; Lyon, Calum

2014-12-22

A 75-year-old man with ulcerative colitis (UC) and diet controlled diabetes mellitus presented with a 3-week history of slightly itchy, red plaques on both lower limbs ascending gradually to cover the trunk and arms. One week later, he developed a flare up of his UC. Routine blood tests showed modest drop in haemoglobin (122 g/L) and C reactive protein (85 mg/L). Serology was remarkable for high antiproteinase 3 (c-ANCA). Serum electrophoresis showed a mildly positive paraprotein band (γ region). Stool culture was negative. Urine analysis showed proteinuria. Skin biopsy showed features of urticarial vasculitis (UV). He underwent a flexible sigmoidoscopy after the flare up showed mildly active UC. The patient was given hydrocortisone for 7 days and then prednisolone. Both rash and UC subsided. Electrophoresis was repeated 4 weeks later showing normal pattern. Prednisolone has been gradually reduced. Although rare, UV can be considered as one of the skin manifestations of UC. 2014 BMJ Publishing Group Ltd.

Structural and optical characterization of NaGdF{sub 4}: Ho{sup 3+}/Yb{sup 3+} UC nano-particles for lateral finger mark detections

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

Kumar, A., E-mail: ak.phy87@gmail.com; Tiwari, S. P.; Krishna, K. M.

2016-05-23

Ho{sup 3+}/Yb{sup 3+} co-doped NaGdF{sub 4} up-conversion (UC) nano-particles were synthesized by thermal decomposition method. X-ray diffraction and FE-SEM image analysis were done to confirm the structure, morphology and determination of particle size. The UC emission spectra for as prepared as well as 100°C, 200°C, 300°C, 400°C, 800°C, 1000°C and 1200°C heated for 3h samples were recorded and there emission intensities were compared at a constant pump power of excitations 98.1 W/cm{sup 2}. The effect of emission intensity on decay time was also studied through focused and unfocused excitations. The synthesized material was successfully utilized in lateral finger mark detections onmore » the glass substrate through powder dusting method.« less

Evaluation and development plan of NRTA measurement methods for the Rokkasho Reprocessing Plant

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

Li, T.K.; Hakkila, E.A.; Flosterbuer, S.F.

Near-real-time accounting (NRTA) has been proposed as a safeguards method at the Rokkasho Reprocessing Plant (RRP), a large-scale commercial boiling water and pressurized water reactors spent-fuel reprocessing facility. NRTA for RRP requires material balance closures every month. To develop a more effective and practical NRTA system for RRP, we have evaluated NRTA measurement techniques and systems that might be implemented in both the main process and the co-denitration process areas at RRP to analyze the concentrations of plutonium in solutions and mixed oxide powder. Based on the comparative evaluation, including performance, reliability, design criteria, operation methods, maintenance requirements, and estimatedmore » costs for each possible measurement method, recommendations for development were formulated. This paper discusses the evaluations and reports on the recommendation of the NRTA development plan for potential implementation at RRP.« less

A MODEL FOR FISSION-GAS RELEASE FROM POROUS FUELS IN LOW-PERMEABILITY CONTAINERS

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

Prados, J.W.

1961-08-25

A simple mathematical model was developed to describe the steady-state release rate of gaseous fission products from porous ceramic fuels in low- permeability containers. The resulting equations are used to analyze experimental release rate results obtained from a UC/sub 2/-fueled graphite fuel body enclosed in a low-permeability impregnated graphite container. The relative release rates of the fission-product species Kr/sup 85m/, Kr/sup 88/, and Xe/sup 133/ were predicted with reasonable success. Absolute-rate predictions were not possible due to lack of information on true permeability and porosity profiles in the graphite container. (auth)

Recent advances in computational actinoid chemistry.

PubMed

Wang, Dongqi; van Gunsteren, Wilfred F; Chai, Zhifang

2012-09-07

We briefly review advances in computational actinoid (An) chemistry during the past ten years in regard to two issues: the geometrical and electronic structures, and reactions. The former addresses the An-O, An-C, and M-An (M is a metal atom including An) bonds in the actinoid molecular systems, including actinoid oxo and oxide species, actinoid-carbenoid, dinuclear and diatomic systems, and the latter the hydration and ligand exchange, the disproportionation, the oxidation, the reduction of uranyl, hydroamination, and the photolysis of uranium azide. Concerning their relevance to the electronic structures and reactions of actinoids and their importance in the development of an advanced nuclear fuel cycle, we also mentioned the work on actinoid carbides and nitrides, which have been proposed to be candidates of the next generation of nuclear fuel, and the oxidation of PuO(x), which is important to understand the speciation of actinoids in the environment, followed by a brief discussion on the urgent need for a heavier involvement of computational actinoid chemistry in developing advanced reprocessing protocols of spent nuclear fuel. The paper is concluded with an outlook.

Text Messaging to Improve Disease Management in Patients With Painful Diabetic Peripheral Neuropathy.

PubMed

Bauer, Victoria; Goodman, Nancy; Lapin, Brittany; Cooley, Camille; Wang, Ed; Craig, Terri L; Glosner, Scott E; Juhn, Mark S; Cappelleri, Joseph C; Sadosky, Alesia B; Masi, Christopher

2018-06-01

Purpose The purpose of the study was to determine the impact of educational text messages on diabetes self-management activities and outcomes in patients with painful diabetic peripheral neuropathy (pDPN). Methods Patients with pDPN identified from a large integrated health system who agreed to participate were randomized to 6 months of usual care (UC) or UC plus twice-daily diabetes self-management text messages (UC+TxtM). Outcomes included the Pain Numerical Rating Scale, Summary of Diabetes Self-Care Activities (SDSCA), questions on diabetes health beliefs, and glycated hemoglobin (A1C). Changes from baseline were evaluated at 6 months and compared between groups. Results Demographic characteristics were balanced between groups (N = 62; 53% female, mean age = 63 years, 94% type 2 diabetes), as were baseline measures. After 6 months, pain decreased with UC+TxtM from 6.3 to 5.5 and with UC from 6.5 to 6.0, with no difference between groups. UC+TxtM but not UC was associated with significant improvements from baseline on all SDSCA subscales. On diabetes health beliefs, UC+TxtM patients reported significantly increased benefits and reduced barriers and susceptibility relative to UC at 6 months. A1C declined in both groups, but neither change was significant relative to baseline. Conclusions Patients with pDPN who receive twice-daily text messages regarding diabetes management reported reduced pain relative to baseline, although this change was not significant compared with usual care. In addition, text messaging was associated with increased self-management activities and improved diabetes health beliefs and total self-care. These results warrant further investigation.

The Military Significance of Small Uranium Enrichment Facilities Fed with Low-Enrichment Uranium (Redacted)

DTIC Science & Technology

1969-12-01

a five-year supply of enriched uranium for reactor fuel . Nevertheless, it seems clear that some foreign enrichment developments are approaching a...produc- tion of fissile material could powerfully influence the assessment of risks and benefits of a nuclear weapons development program . Since... program is likely to include the production of its own relatively pure fissile plutonium. This would involve more rapid cycling and reprocessing of fuel

Method for photochemical reduction of uranyl nitrate by tri-N-butyl phosphate and application of this method to nuclear fuel reprocessing

DOEpatents

De Poorter, Gerald L.; Rofer-De Poorter, Cheryl K.

1978-01-01

Uranyl ion in solution in tri-n-butyl phosphate is readily photochemically reduced to U(IV). The product U(IV) may effectively be used in the Purex process for treating spent nuclear fuels to reduce Pu(IV) to Pu(III). The Pu(III) is readily separated from uranium in solution in the tri-n-butyl phosphate by an aqueous strip.

University-Community Partnership Models: Employing Organizational Management Theories of Paradox and Strategic Contradiction

ERIC Educational Resources Information Center

Bowers, Amanda M.

2017-01-01

University-Community (U-C) partnerships have the potential to respond to society's most pressing needs through engaged scholarship. Despite this promise, partnerships face paradoxical tensions and inherent contradictions that are often not fully addressed in U-C partnership models or frameworks, or in practice. This article seeks to explore the…

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

Vidal, Jean-Marc; Eschbach, Romain; Launay, Agnes

CEA and AREVA-NC have developed and used a depletion code named CESAR for 30 years. This user-friendly industrial tool provides fast characterizations for all types of nuclear fuel (PWR / UOX or MOX or reprocess Uranium, BWR / UOX or MOX, MTR and SFR) and the wastes associated. CESAR can evaluate 100 heavy nuclides, 200 fission products and 150 activation products (with Helium and Tritium formation). It can also characterize the structural material of the fuel (Zircalloy, stainless steel, M5 alloy). CESAR provides depletion calculations for any reactor irradiation history and from 3 months to 1 million years of coolingmore » time. CESAR5.3 is based on the latest calculation schemes recommended by the CEA and on an international nuclear data base (JEFF-3.1.1). It is constantly checked against the CEA referenced and qualified depletion code DARWIN. CESAR incorporates the CEA qualification based on the dissolution analyses of fuel rod samples and the 'La Hague' reprocessing plant feedback experience. AREVA-NC uses CESAR intensively at 'La Hague' plant, not only for prospective studies but also for characterizations at different industrial facilities all along the reprocessing process and waste conditioning (near 150 000 calculations per year). CESAR is the reference code for AREVA-NC. CESAR is used directly or indirectly with other software, data bank or special equipment in many parts of the La Hague plants. The great flexibility of CESAR has rapidly interested other projects. CESAR became a 'tool' directly integrated in some other softwares. Finally, coupled with a Graphical User Interface, it can be easily used independently, responding to many needs for prospective studies as a support for nuclear facilities or transport. An English version is available. For the principal isotopes of U and Pu, CESAR5 benefits from the CEA experimental validation for the PWR UOX fuels, up to a burnup of 60 GWd/t and for PWR MOX fuels, up to 45 GWd/t. CESAR version 5.3 uses the CEA reference calculation codes for neutron physics with the JEFF-3.1.1 nuclear data set. (authors)« less

10 CFR 2.1105 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Hearing Procedures for Expansion of Spent Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors § 2.1105 Definitions. As used in this part: (a) Civilian nuclear power reactor means a civilian... reactor following irradiation, the constituent elements of which have not been separated by reprocessing. ...

MISCELLANEOUS ARCHITECTURAL DETAILS AND SECTIONS OF REMOTE ANALYTICAL FACILITY (CPP627). ...

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

MISCELLANEOUS ARCHITECTURAL DETAILS AND SECTIONS OF REMOTE ANALYTICAL FACILITY (CPP-627). INL DRAWING NUMBER 200-0627-00-098-105632. ALTERNATE ID NUMBER 4272-814-135. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO SHOWING EXCAVATION PIT FOR MAIN PROCESSING BUILDING ...

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

CONSTRUCTION PROGRESS PHOTO SHOWING EXCAVATION PIT FOR MAIN PROCESSING BUILDING (CPP-601) LOOKING NORTHWEST. INL PHOTO NUMBER NRTS-50-885. Unknown Photographer, 10/30/1950 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

SOUTH ELEVATION AND DETAILS OF MAIN PROCESSING BUILDING (CPP601). INL ...

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

SOUTH ELEVATION AND DETAILS OF MAIN PROCESSING BUILDING (CPP-601). INL DRAWING NUMBER 200-0601-00-291-103082. ALTERNATE ID NUMBER 542-12-B-76. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

ARCHITECTURAL WALL SECTIONS OF HOT PILOT PLANT (CPP640). INL DRAWING ...

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

ARCHITECTURAL WALL SECTIONS OF HOT PILOT PLANT (CPP-640). INL DRAWING NUMBER 200-0640-00-279-111682. ALTERNATE ID NUMBER 8952-CPP-640-A-5. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

NORTH AND SOUTH SECTIONS OF REMOTE ANALYTICAL FACILITY (CPP627). INL ...

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

NORTH AND SOUTH SECTIONS OF REMOTE ANALYTICAL FACILITY (CPP-627). INL DRAWING NUMBER 200-0627-00-098-105068. ALTERNATE ID NUMBER 4272-14-105. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

AERIAL VIEW OF MAIN PROCESSING BUILDING SHOWING CONSTRUCTION PROGRESS AND ...

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

AERIAL VIEW OF MAIN PROCESSING BUILDING SHOWING CONSTRUCTION PROGRESS AND EXCAVATION FOR LABORATORY ON LEFT. INL PHOTO NUMBER NRTS-51-1759. Unknown Photographer, 3/28/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

EAST AND WEST ELEVATIONS OF REMOTE ANALYTICAL FACILITY (CPP627). INL ...

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

EAST AND WEST ELEVATIONS OF REMOTE ANALYTICAL FACILITY (CPP-627). INL DRAWING NUMBER 200-0627-00-098-105067. ALTERNATE ID NUMBER 4272-14-104. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

EAST AND WEST ELEVATIONS OF MAIN PROCESSING BUILDING (CPP601). INL ...

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

EAST AND WEST ELEVATIONS OF MAIN PROCESSING BUILDING (CPP-601). INL DRAWING NUMBER 200-0601-00-291-103081. ALTERNATE ID NUMBER 542-11-B-75. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

INTERIOR PHOTO OF THE REMOTE ANALYTICAL FACILITY OF SHIELDED GLOVE ...

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

INTERIOR PHOTO OF THE REMOTE ANALYTICAL FACILITY OF SHIELDED GLOVE BOXES IN OPERATING CORRIDOR (CPP-627). INL PHOTO NUMBER NRTS-55-1524. Unknown Photographer, 1955 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

MISCELLANEOUS ARCHITECTURAL DETAILS OF HOT PILOT PLANT (CPP640). INL DRAWING ...

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

MISCELLANEOUS ARCHITECTURAL DETAILS OF HOT PILOT PLANT (CPP-640). INL DRAWING NUMBER 200-640-00-279-111684. ALTERNATE ID NUMBER 8952-CPP-640-A-7. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

BUILDING DETAILS AND SECTIONS OF MAIN PROCESSING BUILDING (CPP601). INL ...

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

BUILDING DETAILS AND SECTIONS OF MAIN PROCESSING BUILDING (CPP-601). INL DRAWING NUMBER 200-0601-00-291-103080. ALTERNATE ID NUMBER 542-11-B-74. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

STRUCTURAL DETAILS AND SECTIONS OF MAIN PROCESSING BUILDING (CPP601). INL ...

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

STRUCTURAL DETAILS AND SECTIONS OF MAIN PROCESSING BUILDING (CPP-601). INL DRAWING NUMBER 200-0601-00-291-103079. ALTERNATE ID NUMBER 542-11-B-73. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO OF REMOTE ANALYTICAL FACILITY (CPP627) SHOWING PLACEMENT ...

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

CONSTRUCTION PROGRESS PHOTO OF REMOTE ANALYTICAL FACILITY (CPP-627) SHOWING PLACEMENT OF PIERS. INL PHOTO NUMBER NRTS-54-11716. Unknown Photographer, 8/20/1954 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

10 CFR 110.41 - Executive Branch review.

Code of Federal Regulations, 2011 CFR

2011-01-01

.... (6) An export involving assistance to end uses related to isotope separation, chemical reprocessing, heavy water production, advanced reactors, or the fabrication of nuclear fuel containing plutonium... equipment to a foreign reactor. (8) An export involving radioactive waste. (9) An export to any country...

Local atomic and electronic structure of LaCoO3 /SrTiO3 thin films by HAADF STEM and EELS

NASA Astrophysics Data System (ADS)

Borisevich, Albina; Hyuck Jang, Jae; Kim, Young-Min; Qiao, Liang; Biegalski, Michael

2013-03-01

For perovskite films with several competing functionalities, magnetic and electronic properties can be affected both by structural order parameters and chemical factors. For example, in LaCoO3 (LCO) thin films, magnetic and transport properties are strongly dependent on strain state and oxygen content. For this study, LCO thin films were deposited by pulsed laser deposition method with different thicknesses (2, 5, 15 unit cell and 20 nm thickness) on SrTiO3 substrate. X-ray photoelectron spectroscopy studies of the grown films have demonstrated that Co 3p edges shift up to 2 eV for 15 u.c. and 20 nm films, indicating possible presence of 2D electron gas. The structure of the 5 u.c and 15 u.c LCO films was examined. Atomic position mapping from STEM HAADF and BF images can reveal lattice parameter and octahedral tilt behavior with atomic resolution. BF STEM imaging showed that octahedral tilts were active in the 15 u.c. film but not in the 5 u.c. film. A complex pattern of O K fine structure evolution at the interface was observed; results of the deconvolution of different contributions to this behavior using advanced simulations, as well as data on oxygen vacancy mapping, will be presented. Research supported by the US DOE-BES, Materials Sciences and Engineering Division, and through a user project supported by ORNL's ShaRE User Program.

Direct observation of the core/double-shell architecture of intense dual-mode luminescent tetragonal bipyramidal nanophosphors

NASA Astrophysics Data System (ADS)

Kim, Su Yeon; Jeong, Jong Seok; Mkhoyan, K. Andre; Jang, Ho Seong

2016-05-01

Highly efficient downconversion (DC) green-emitting LiYF4:Ce,Tb nanophosphors have been synthesized for bright dual-mode upconversion (UC) and DC green-emitting core/double-shell (C/D-S) nanophosphors--Li(Gd,Y)F4:Yb(18%),Er(2%)/LiYF4:Ce(15%),Tb(15%)/LiYF4--and the C/D-S structure has been proved by extensive scanning transmission electron microscopy (STEM) analysis. Colloidal LiYF4:Ce,Tb nanophosphors with a tetragonal bipyramidal shape are synthesized for the first time and they show intense DC green light via energy transfer from Ce3+ to Tb3+ under illumination with ultraviolet (UV) light. The LiYF4:Ce,Tb nanophosphors show 65 times higher photoluminescence intensity than LiYF4:Tb nanophosphors under illumination with UV light and the LiYF4:Ce,Tb is adapted into a luminescent shell of the tetragonal bipyramidal C/D-S nanophosphors. The formation of the DC shell on the core significantly enhances UC luminescence from the UC core under irradiation of near infrared light and concurrently generates DC luminescence from the core/shell nanophosphors under UV light. Coating with an inert inorganic shell further enhances the UC-DC dual-mode luminescence by suppressing the surface quenching effect. The C/D-S nanophosphors show 3.8% UC quantum efficiency (QE) at 239 W cm-2 and 73.0 +/- 0.1% DC QE. The designed C/D-S architecture in tetragonal bipyramidal nanophosphors is rigorously verified by an energy dispersive X-ray spectroscopy (EDX) analysis, with the assistance of line profile simulation, using an aberration-corrected scanning transmission electron microscope equipped with a high-efficiency EDX. The feasibility of these C/D-S nanophosphors for transparent display devices is also considered.Highly efficient downconversion (DC) green-emitting LiYF4:Ce,Tb nanophosphors have been synthesized for bright dual-mode upconversion (UC) and DC green-emitting core/double-shell (C/D-S) nanophosphors--Li(Gd,Y)F4:Yb(18%),Er(2%)/LiYF4:Ce(15%),Tb(15%)/LiYF4--and the C/D-S structure has been proved by extensive scanning transmission electron microscopy (STEM) analysis. Colloidal LiYF4:Ce,Tb nanophosphors with a tetragonal bipyramidal shape are synthesized for the first time and they show intense DC green light via energy transfer from Ce3+ to Tb3+ under illumination with ultraviolet (UV) light. The LiYF4:Ce,Tb nanophosphors show 65 times higher photoluminescence intensity than LiYF4:Tb nanophosphors under illumination with UV light and the LiYF4:Ce,Tb is adapted into a luminescent shell of the tetragonal bipyramidal C/D-S nanophosphors. The formation of the DC shell on the core significantly enhances UC luminescence from the UC core under irradiation of near infrared light and concurrently generates DC luminescence from the core/shell nanophosphors under UV light. Coating with an inert inorganic shell further enhances the UC-DC dual-mode luminescence by suppressing the surface quenching effect. The C/D-S nanophosphors show 3.8% UC quantum efficiency (QE) at 239 W cm-2 and 73.0 +/- 0.1% DC QE. The designed C/D-S architecture in tetragonal bipyramidal nanophosphors is rigorously verified by an energy dispersive X-ray spectroscopy (EDX) analysis, with the assistance of line profile simulation, using an aberration-corrected scanning transmission electron microscope equipped with a high-efficiency EDX. The feasibility of these C/D-S nanophosphors for transparent display devices is also considered. Electronic supplementary information (ESI) available: XRD patterns, PL and PLE spectra, SEM and HR-TEM images, PL decay times, photographs showing the transparent nanophosphor solutions and their dual-mode luminescence, and additional EDX data. See DOI: 10.1039/c5nr05722a

Online disease management of diabetes: engaging and motivating patients online with enhanced resources-diabetes (EMPOWER-D), a randomized controlled trial.

PubMed

Tang, Paul C; Overhage, J Marc; Chan, Albert Solomon; Brown, Nancy L; Aghighi, Bahar; Entwistle, Martin P; Hui, Siu Lui; Hyde, Shauna M; Klieman, Linda H; Mitchell, Charlotte J; Perkins, Anthony J; Qureshi, Lubna S; Waltimyer, Tanya A; Winters, Leigha J; Young, Charles Y

2013-05-01

To evaluate an online disease management system supporting patients with uncontrolled type 2 diabetes. Engaging and Motivating Patients Online With Enhanced Resources for Diabetes was a 12-month parallel randomized controlled trial of 415 patients with type 2 diabetes with baseline glycosylated hemoglobin (A1C) values ≥7.5% from primary care sites sharing an electronic health record. The intervention included: (1) wirelessly uploaded home glucometer readings with graphical feedback; (2) comprehensive patient-specific diabetes summary status report; (3) nutrition and exercise logs; (4) insulin record; (5) online messaging with the patient's health team; (6) nurse care manager and dietitian providing advice and medication management; and (7) personalized text and video educational 'nuggets' dispensed electronically by the care team. A1C was the primary outcome variable. Compared with usual care (UC, n=189), patients in the intervention (INT, n=193) group had significantly reduced A1C at 6 months (-1.32% INT vs -0.66% UC; p<0.001). At 12 months, the differences were not significant (-1.14% INT vs -0.95% UC; p=0.133). In post hoc analysis, significantly more INT patients had improved diabetes control (>0.5% reduction in A1C) than UC patients at 12 months (69.9 (95% CI 63.2 to 76.5) vs 55.4 (95% CI 48.4 to 62.5); p=0.006). A nurse-led, multidisciplinary health team can manage a population of diabetic patients in an online disease management program. INT patients achieved greater decreases in A1C at 6 months than UC patients, but the differences were not sustained at 12 months. More INT than UC patients achieved improvement in A1C (>0.5% decrease). Trial registered in clinical trials.gov: #NCT00542204.

Online disease management of diabetes: Engaging and Motivating Patients Online With Enhanced Resources-Diabetes (EMPOWER-D), a randomized controlled trial

PubMed Central

Tang, Paul C; Overhage, J Marc; Chan, Albert Solomon; Brown, Nancy L; Aghighi, Bahar; Entwistle, Martin P; Hui, Siu Lui; Hyde, Shauna M; Klieman, Linda H; Mitchell, Charlotte J; Perkins, Anthony J; Qureshi, Lubna S; Waltimyer, Tanya A; Winters, Leigha J; Young, Charles Y

2013-01-01

Objective To evaluate an online disease management system supporting patients with uncontrolled type 2 diabetes. Materials and methods Engaging and Motivating Patients Online With Enhanced Resources for Diabetes was a 12-month parallel randomized controlled trial of 415 patients with type 2 diabetes with baseline glycosylated hemoglobin (A1C) values ≥7.5% from primary care sites sharing an electronic health record. The intervention included: (1) wirelessly uploaded home glucometer readings with graphical feedback; (2) comprehensive patient-specific diabetes summary status report; (3) nutrition and exercise logs; (4) insulin record; (5) online messaging with the patient's health team; (6) nurse care manager and dietitian providing advice and medication management; and (7) personalized text and video educational ‘nuggets’ dispensed electronically by the care team. A1C was the primary outcome variable. Results Compared with usual care (UC, n=189), patients in the intervention (INT, n=193) group had significantly reduced A1C at 6 months (−1.32% INT vs −0.66% UC; p<0.001). At 12 months, the differences were not significant (−1.14% INT vs −0.95% UC; p=0.133). In post hoc analysis, significantly more INT patients had improved diabetes control (>0.5% reduction in A1C) than UC patients at 12 months (69.9 (95% CI 63.2 to 76.5) vs 55.4 (95% CI 48.4 to 62.5); p=0.006). Conclusions A nurse-led, multidisciplinary health team can manage a population of diabetic patients in an online disease management program. INT patients achieved greater decreases in A1C at 6 months than UC patients, but the differences were not sustained at 12 months. More INT than UC patients achieved improvement in A1C (>0.5% decrease). Trial registered in clinical trials.gov: #NCT00542204 PMID:23171659

Patient satisfaction with different interpreting methods: a randomized controlled trial.

PubMed

Gany, Francesca; Leng, Jennifer; Shapiro, Ephraim; Abramson, David; Motola, Ivette; Shield, David C; Changrani, Jyotsna

2007-11-01

Growth of the foreign-born population in the U.S. has led to increasing numbers of limited-English-proficient (LEP) patients. Innovative medical interpreting strategies, including remote simultaneous medical interpreting (RSMI), have arisen to address the language barrier. This study evaluates the impact of interpreting method on patient satisfaction. 1,276 English-, Spanish-, Mandarin-, and Cantonese-speaking patients attending the primary care clinic and emergency department of a large New York City municipal hospital were screened for enrollment in a randomized controlled trial. Language-discordant patients were randomized to RSMI or usual and customary (U&C) interpreting. Patients with language-concordant providers received usual care. Demographic and patient satisfaction questionnaires were administered to all participants. 541 patients were language-concordant with their providers and not randomized; 371 were randomized to RSMI, 167 of whom were exposed to RSMI; and 364 were randomized to U&C, 198 of whom were exposed to U&C. Patients randomized to RSMI were more likely than those with U&C to think doctors treated them with respect (RSMI 71%, U&C 64%, p < 0.05), but they did not differ in other measures of physician communication/care. In a linear regression analysis, exposure to RSMI was significantly associated with an increase in overall satisfaction with physician communication/care (beta 0.10, 95% CI 0.02-0.18, scale 0-1.0). Patients randomized to RSMI were more likely to think the interpreting method protected their privacy (RSMI 51%, U&C 38%, p < 0.05). Patients randomized to either arm of interpretation reported less comprehension and satisfaction than patients in language-concordant encounters. While not a substitute for language-concordant providers, RSMI can improve patient satisfaction and privacy among LEP patients. Implementing RSMI should be considered an important component of a multipronged approach to addressing language barriers in health care.

Patient Satisfaction with Different Interpreting Methods: A Randomized Controlled Trial

PubMed Central

Leng, Jennifer; Shapiro, Ephraim; Abramson, David; Motola, Ivette; Shield, David C.; Changrani, Jyotsna

2007-01-01

Background Growth of the foreign-born population in the U.S. has led to increasing numbers of limited-English-proficient (LEP) patients. Innovative medical interpreting strategies, including remote simultaneous medical interpreting (RSMI), have arisen to address the language barrier. This study evaluates the impact of interpreting method on patient satisfaction. Methods 1,276 English-, Spanish-, Mandarin-, and Cantonese-speaking patients attending the primary care clinic and emergency department of a large New York City municipal hospital were screened for enrollment in a randomized controlled trial. Language-discordant patients were randomized to RSMI or usual and customary (U&C) interpreting. Patients with language-concordant providers received usual care. Demographic and patient satisfaction questionnaires were administered to all participants. Results 541 patients were language-concordant with their providers and not randomized; 371 were randomized to RSMI, 167 of whom were exposed to RSMI; and 364 were randomized to U&C, 198 of whom were exposed to U&C. Patients randomized to RSMI were more likely than those with U&C to think doctors treated them with respect (RSMI 71%, U&C 64%, p < 0.05), but they did not differ in other measures of physician communication/care. In a linear regression analysis, exposure to RSMI was significantly associated with an increase in overall satisfaction with physician communication/care (β 0.10, 95% CI 0.02–0.18, scale 0–1.0). Patients randomized to RSMI were more likely to think the interpreting method protected their privacy (RSMI 51%, U&C 38%, p < 0.05). Patients randomized to either arm of interpretation reported less comprehension and satisfaction than patients in language-concordant encounters. Conclusions While not a substitute for language-concordant providers, RSMI can improve patient satisfaction and privacy among LEP patients. Implementing RSMI should be considered an important component of a multipronged approach to addressing language barriers in health care. PMID:17957417

CTLA-4 polymorphisms and susceptibility to inflammatory bowel disease: a meta-analysis.

PubMed

Lee, Young Ho; Kim, Jae-Hoon; Seo, Young Ho; Choi, Sung Jae; Ji, Jong Dae; Song, Gwan Gyu

2014-05-01

The aim of this study was to explore whether the cytotoxic T lymphocyte associated antigen-4 (CTLA-4) polymorphisms are associated with susceptibility to ulcerative colitis (UC) and Crohn's disease (CD). The authors conducted a meta-analysis on associations between CTLA-4 +49 A/G, -318 C/T, CT60 A/G polymorphisms, and (AT)n repeat in the 3' untranslated region (UTR) and UC and CD susceptibility. A total of 15 comparison studies were considered in our meta-analysis. Meta-analysis revealed no association between UC and the CTLA-4 +49 G and CTLA-4 -318 T alleles in all subjects (OR=0.982, 95% CI=0.851-1.1339, p=0.804; OR=0.500, 95% CI=0.223-1.124, p=0.094). No association was found between UC and the CTLA-4 CT60 A/G polymorphism in Europeans. However, a significant association was observed between the longer allele (⩾118bp) of the (AT)n and UC in Asian population (OR=6.073, 95% CI=4.246-8.684, p=1.0×10(-9)). Meta-analysis of the CTLA-4 +49 A/G, -318 C/T, CT60 A/G polymorphisms showed no association with CD. This meta-analysis demonstrates that the CTLA-4 (AT)n repeat in 3' UTR may be associated with susceptibility to UC in Asians, while no association was found between the CTLA-4 +49 A/G, -318 C/T, and CD60 A/G polymorphism and susceptibility to UC and CD. Copyright © 2014 American Society for Histocompatibility and Immunogenetics. Published by Elsevier Inc. All rights reserved.

Gla-Rich Protein Is a Potential New Vitamin K Target in Cancer: Evidences for a Direct GRP-Mineral Interaction

PubMed Central

Viegas, Carla S. B.; Herfs, Marjolein; Rafael, Marta S.; Enriquez, José L.; Teixeira, Alexandra; Luís, Inês M.; van ‘t Hoofd, Cynthia M. R.; João, Alexandre; Maria, Vera L.; Cavaco, Sofia; Ferreira, Ana; Serra, Manuel; Theuwissen, Elke; Vermeer, Cees; Simes, Dina C.

2014-01-01

Gla-rich protein (GRP) was described in sturgeon as a new vitamin-K-dependent protein (VKDP) with a high density of Gla residues and associated with ectopic calcifications in humans. Although VKDPs function has been related with γ-carboxylation, the Gla status of GRP in humans is still unknown. Here, we investigated the expression of recently identified GRP spliced transcripts, the γ-carboxylation status, and its association with ectopic calcifications, in skin basal cell and breast carcinomas. GRP-F1 was identified as the predominant splice variant expressed in healthy and cancer tissues. Patterns of γ-carboxylated GRP (cGRP)/undercarboxylated GRP (ucGRP) accumulation in healthy and cancer tissues were determined by immunohistochemistry, using newly developed conformation-specific antibodies. Both GRP protein forms were found colocalized in healthy tissues, while ucGRP was the predominant form associated with tumor cells. Both cGRP and ucGRP found at sites of microcalcifications were shown to have in vitro calcium mineral-binding capacity. The decreased levels of cGRP and predominance of ucGRP in tumor cells suggest that GRP may represent a new target for the anticancer potential of vitamin K. Also, the direct interaction of cGRP and ucGRP with BCP crystals provides a possible mechanism explaining GRP association with pathological mineralization. PMID:24949434

Progress on Cleaning Up the Only Commercial Nuclear Fuel Reprocessing Facility to Operate in the United States

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

Jackson, T. J.; MacVean, S. A.; Szlis, K. A.

2002-02-26

This paper describes the progress on cleanup of the West Valley Demonstration Project (WVDP), an environmental management project located south of Buffalo, NY. The WVDP was the site of the only commercial nuclear fuel reprocessing facility to have operated in the United States (1966 to 1972). Former fuel reprocessing operations generated approximately 600,000 gallons of liquid high-level radioactive waste stored in underground tanks. The U.S. Congress passed the WVDP Act in 1980 (WVDP Act) to authorize cleanup of the 220-acre facility. The facility is unique in that it sits on the 3,345-acre Western New York Nuclear Service Center (WNYNSC), whichmore » is owned by New York State through the New York State Energy Research and Development Authority (NYSERDA). The U.S. Department of Energy (DOE) has overall responsibility for the cleanup that is authorized by the WVDP Act, paying 90 percent of the WVDP costs; NYSERDA pays 10 percent. West Valley Nuclear Services Company (WVNSCO) is the management contractor at the WVDP. This paper will provide a description of the many accomplishments at the WVDP, including the pretreatment and near completion of vitrification of all the site's liquid high-level radioactive waste, a demonstration of technologies to characterize the remaining material in the high-level waste tanks, the commencement of decontamination and decommissioning (D&D) activities to place the site in a safe configuration for long-term site management options, and achievement of several technological firsts. It will also include a discussion of the complexities involved in completing the WVDP due to the various agency interests that require integration for future cleanup decisions.« less

Current state of nuclear fuel cycles in nuclear engineering and trends in their development according to the environmental safety requirements

NASA Astrophysics Data System (ADS)

Vislov, I. S.; Pischulin, V. P.; Kladiev, S. N.; Slobodyan, S. M.

2016-08-01

The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium-plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel assemblies. Based on analysis of modern engineering solutions on SNF regeneration, it has been concluded that new reprocessing technologies should meet the ecological safety requirements, provide a more extensive use of the resource base of nuclear engineering, allow the production of valuable and trace elements on an industrial scale, and decrease radioactive waste release.

Next Generation Safeguards Initiative research to determine the Pu mass in spent fuel assemblies: Purpose, approach, constraints, implementation, and calibration

NASA Astrophysics Data System (ADS)

Tobin, S. J.; Menlove, H. O.; Swinhoe, M. T.; Schear, M. A.

2011-10-01

The Next Generation Safeguards Initiative (NGSI) of the U.S. Department of Energy has funded a multi-lab/multi-university collaboration to quantify the plutonium mass in spent nuclear fuel assemblies and to detect the diversion of pins from them. The goal of this research effort is to quantify the capability of various non-destructive assay (NDA) technologies as well as to train a future generation of safeguards practitioners. This research is "technology driven" in the sense that we will quantify the capabilities of a wide range of safeguards technologies of interest to regulators and policy makers; a key benefit to this approach is that the techniques are being tested in a unified manner. When the results of the Monte Carlo modeling are evaluated and integrated, practical constraints are part of defining the potential context in which a given technology might be applied. This paper organizes the commercial spent fuel safeguard needs into four facility types in order to identify any constraints on the NDA system design. These four facility types are the following: future reprocessing plants, current reprocessing plants, once-through spent fuel repositories, and any other sites that store individual spent fuel assemblies (reactor sites are the most common facility type in this category). Dry storage is not of interest since individual assemblies are not accessible. This paper will overview the purpose and approach of the NGSI spent fuel effort and describe the constraints inherent in commercial fuel facilities. It will conclude by discussing implementation and calibration of measurement systems. This report will also provide some motivation for considering a couple of other safeguards concepts (base measurement and fingerprinting) that might meet the safeguards need but not require the determination of plutonium mass.

Recovery of transplutonium elements from nuclear reactor waste

DOEpatents

Campbell, David O.; Buxton, Samuel R.

1977-05-24

A method of separating actinide values from nitric acid waste solutions resulting from reprocessing of irradiated nuclear fuels comprises oxalate precipitation of the major portion of actinide and lanthanide values to provide a trivalent fraction suitable for subsequent actinide/lanthanide partition, exchange of actinide and lanthanide values in the supernate onto a suitable cation exchange resin to provide an intermediate-lived raffinate waste stream substantially free of actinides, and elution of the actinide values from the exchange resin. The eluate is then used to dissolve the trivalent oxalate fraction prior to actinide/lanthanide partition or may be combined with the reprocessing waste stream and recycled.

EAST WEST NORTH ELEVATIONS OF MULTICURIE CELL ARCHITECTURAL DETAILS REMOTE ...

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

EAST WEST NORTH ELEVATIONS OF MULTICURIE CELL ARCHITECTURAL DETAILS REMOTE ANALYTICAL FACILITY (CPP-627). INL DRAWING NUMBER 200-00627-00-706-050245. ALTERNATE ID NUMBER AED-D-245. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

ARCHITECTURAL FLOOR PLAN OF OPERATING AREA HOT PILOT PLANT (CPP640). ...

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

ARCHITECTURAL FLOOR PLAN OF OPERATING AREA HOT PILOT PLANT (CPP-640). INL DRAWING NUMBER 200-0640-00-279-111678. ALTERNATE ID NUMBER 8952-CPP-640-A-1. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO REMOTE ANALYTICAL FACILITY (CPP627) SHOWING EMPLACEMENT OF ...

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

CONSTRUCTION PROGRESS PHOTO REMOTE ANALYTICAL FACILITY (CPP-627) SHOWING EMPLACEMENT OF ROOF SLABS. INL PHOTO NUMBER NRTS-54-13463. R.G. Larsen, Photographer, 12/20/1954 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

PLAN SECTIONS AND DETAILS OF CELL HATCHES MAIN PROCESSING BUILDING ...

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

PLAN SECTIONS AND DETAILS OF CELL HATCHES MAIN PROCESSING BUILDING (CPP-601). INL DRAWING NUMBER 200-0601-00-291-103256. ALTERNATE ID NUMBER 542-11-F-302. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

ARCHITECTURAL DOOR DETAILS AND SCHEDULE OF HOT PILOT PLANT (CPP640). ...

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

ARCHITECTURAL DOOR DETAILS AND SCHEDULE OF HOT PILOT PLANT (CPP-640). INL DRAWING NUMBER 200-640-00-279-111683. ALTERNATE ID NUMBER 8952-CPP-640-A-6. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Characterization of a cold-active bacterium isolated from the South Pole “Ice Tunnel”

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

Madigan, Michael T.; Kempher, Megan L.; Bender, Kelly S.

Abstract Extremely cold microbial habitats on Earth (those below -30 °C) are rare and have not been surveyed for microbes as extensively as environments in the 0 to -20 °C range. Using cryoprotected growth media incubated at -5 °C, we enriched a cold-active Pseudomonas species from -50 °C ice collected from a utility tunnel for wastewater pipes under Amundsen–Scott South Pole Station, Antarctica. The isolate, strain UC-1, is related to other cold-active Pseudomonas species, most notably P. psychrophila, and grew at -5 °C to +34–37 °C; growth of UC-1 at +3 °C was significantly faster than at +34 °C. Strainmore » UC-1 synthesized a surface exopolymer and high levels of unsaturated fatty acids under cold growth conditions. A 16S rRNA gene diversity screen of the ice sample that yielded strain UC-1 revealed over 1200 operational taxonomic units (OTUs) distributed across eight major classes of Bacteria. Many of the OTUs were Clostridia and Bacteriodia and some of these were probably of wastewater origin. However, a significant fraction of the OTUs were Proteobacteria and Actinobacteria of likely environmental origin. Our results shed light on the lower temperature limits to life and the possible existence of functional microbial communities in ultra-cold environments.« less

Longitudinal microbiome analysis of single donor fecal microbiota transplantation in patients with recurrent Clostridium difficile infection and/or ulcerative colitis

PubMed Central

Grewal, Suman; LaComb, Joseph F.; Park, Jiyhe; Channer, Breana; Rajapakse, Ramona; Bucobo, Juan Carlos; Buscaglia, Jonathan M.; Monzur, Farah; Chawla, Anupama; Yang, Jie; Robertson, Charlie E.; Frank, Daniel N.; Li, Ellen

2018-01-01

Background Studies of colonoscopic fecal microbiota transplant (FMT) in patients with recurrent CDI, indicate that this is a very effective treatment for preventing further relapses. In order to provide this service at Stony Brook University Hospital, we initiated an open-label prospective study of single colonoscopic FMT among patients with ≥ 2 recurrences of CDI, with the intention of monitoring microbial composition in the recipient before and after FMT, as compared with their respective donor. We also initiated a concurrent open label prospective trial of single colonoscopic FMT of patients with ulcerative colitis (UC) not responsive to therapy, after obtaining an IND permit (IND 15642). To characterize how FMT alters the fecal microbiota in patients with recurrent Clostridia difficile infections (CDI) and/or UC, we report the results of a pilot microbiome analysis of 11 recipients with a history of 2 or more recurrences of C. difficile infections without inflammatory bowel disease (CDI-only), 3 UC recipients with recurrent C. difficile infections (CDI + UC), and 5 UC recipients without a history of C. difficile infections (UC-only). Method V3V4 Illumina 16S ribosomal RNA (rRNA) gene sequencing was performed on the pre-FMT, 1-week post-FMT, and 3-months post-FMT recipient fecal samples along with those collected from the healthy donors. Fitted linear mixed models were used to examine the effects of Group (CDI-only, CDI + UC, UC-only), timing of FMT (Donor, pre-FMT, 1-week post-FMT, 3-months post-FMT) and first order Group*FMT interactions on the diversity and composition of fecal microbiota. Pairwise comparisons were then carried out on the recipient vs. donor and between the pre-FMT, 1-week post-FMT, and 3-months post-FMT recipient samples within each group. Results Significant effects of FMT on overall microbiota composition (e.g., beta diversity) were observed for the CDI-only and CDI + UC groups. Marked decreases in the relative abundances of the strictly anaerobic Bacteroidetes phylum, and two Firmicutes sub-phyla associated with butyrate production (Ruminococcaceae and Lachnospiraceae) were observed between the CDI-only and CDI + UC recipient groups. There were corresponding increases in the microaerophilic Proteobacteria phylum and the Firmicutes/Bacilli group in the CDI-only and CDI + UC recipient groups. At a more granular level, significant effects of FMT were observed for 81 genus-level operational taxonomic units (OTUs) in at least one of the three recipient groups (p<0.00016 with Bonferroni correction). Pairwise comparisons of the estimated pre-FMT recipient/donor relative abundance ratios identified 6 Gammaproteobacteria OTUs, including the Escherichia-Shigella genus, and 2 Fusobacteria OTUs with significantly increased relative abundance in the pre-FMT samples of all three recipient groups (FDR < 0.05), however the magnitude of the fold change was much larger in the CDI-only and CDI + UC recipients than in the UC-only recipients. Depletion of butyrate producing OTUs, such as Faecalibacterium, in the CDI-only and CDI + UC recipients, were restored after FMT. Conclusion The results from this pilot study suggest that the microbial imbalances in the CDI + UC recipients more closely resemble those of the CDI-only recipients than the UC-only recipients. PMID:29385143

Longitudinal microbiome analysis of single donor fecal microbiota transplantation in patients with recurrent Clostridium difficile infection and/or ulcerative colitis.

PubMed

Mintz, Michael; Khair, Shanawaj; Grewal, Suman; LaComb, Joseph F; Park, Jiyhe; Channer, Breana; Rajapakse, Ramona; Bucobo, Juan Carlos; Buscaglia, Jonathan M; Monzur, Farah; Chawla, Anupama; Yang, Jie; Robertson, Charlie E; Frank, Daniel N; Li, Ellen

2018-01-01

Studies of colonoscopic fecal microbiota transplant (FMT) in patients with recurrent CDI, indicate that this is a very effective treatment for preventing further relapses. In order to provide this service at Stony Brook University Hospital, we initiated an open-label prospective study of single colonoscopic FMT among patients with ≥ 2 recurrences of CDI, with the intention of monitoring microbial composition in the recipient before and after FMT, as compared with their respective donor. We also initiated a concurrent open label prospective trial of single colonoscopic FMT of patients with ulcerative colitis (UC) not responsive to therapy, after obtaining an IND permit (IND 15642). To characterize how FMT alters the fecal microbiota in patients with recurrent Clostridia difficile infections (CDI) and/or UC, we report the results of a pilot microbiome analysis of 11 recipients with a history of 2 or more recurrences of C. difficile infections without inflammatory bowel disease (CDI-only), 3 UC recipients with recurrent C. difficile infections (CDI + UC), and 5 UC recipients without a history of C. difficile infections (UC-only). V3V4 Illumina 16S ribosomal RNA (rRNA) gene sequencing was performed on the pre-FMT, 1-week post-FMT, and 3-months post-FMT recipient fecal samples along with those collected from the healthy donors. Fitted linear mixed models were used to examine the effects of Group (CDI-only, CDI + UC, UC-only), timing of FMT (Donor, pre-FMT, 1-week post-FMT, 3-months post-FMT) and first order Group*FMT interactions on the diversity and composition of fecal microbiota. Pairwise comparisons were then carried out on the recipient vs. donor and between the pre-FMT, 1-week post-FMT, and 3-months post-FMT recipient samples within each group. Significant effects of FMT on overall microbiota composition (e.g., beta diversity) were observed for the CDI-only and CDI + UC groups. Marked decreases in the relative abundances of the strictly anaerobic Bacteroidetes phylum, and two Firmicutes sub-phyla associated with butyrate production (Ruminococcaceae and Lachnospiraceae) were observed between the CDI-only and CDI + UC recipient groups. There were corresponding increases in the microaerophilic Proteobacteria phylum and the Firmicutes/Bacilli group in the CDI-only and CDI + UC recipient groups. At a more granular level, significant effects of FMT were observed for 81 genus-level operational taxonomic units (OTUs) in at least one of the three recipient groups (p<0.00016 with Bonferroni correction). Pairwise comparisons of the estimated pre-FMT recipient/donor relative abundance ratios identified 6 Gammaproteobacteria OTUs, including the Escherichia-Shigella genus, and 2 Fusobacteria OTUs with significantly increased relative abundance in the pre-FMT samples of all three recipient groups (FDR < 0.05), however the magnitude of the fold change was much larger in the CDI-only and CDI + UC recipients than in the UC-only recipients. Depletion of butyrate producing OTUs, such as Faecalibacterium, in the CDI-only and CDI + UC recipients, were restored after FMT. The results from this pilot study suggest that the microbial imbalances in the CDI + UC recipients more closely resemble those of the CDI-only recipients than the UC-only recipients.

Improving the Estimates of Waste from the Recycling of Used Nuclear Fuel - 13410

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

Phillips, Chris; Willis, William; Carter, Robert

2013-07-01

Estimates are presented of wastes arising from the reprocessing of 50 GWD/tonne, 5 year and 50 year cooled used nuclear fuel (UNF) from Light Water Reactors (LWRs), using the 'NUEX' solvent extraction process. NUEX is a fourth generation aqueous based reprocessing system, comprising shearing and dissolution in nitric acid of the UNF, separation of uranium and mixed uranium-plutonium using solvent extraction in a development of the PUREX process using tri-n-butyl phosphate in a kerosene diluent, purification of the plutonium and uranium-plutonium products, and conversion of them to uranium trioxide and mixed uranium-plutonium dioxides respectively. These products are suitable for usemore » as new LWR uranium oxide and mixed oxide fuel, respectively. Each unit process is described and the wastes that it produces are identified and quantified. Quantification of the process wastes was achieved by use of a detailed process model developed using the Aspen Custom Modeler suite of software and based on both first principles equilibrium and rate data, plus practical experience and data from the industrial scale Thermal Oxide Reprocessing Plant (THORP) at the Sellafield nuclear site in the United Kingdom. By feeding this model with the known concentrations of all species in the incoming UNF, the species and their concentrations in all product and waste streams were produced as the output. By using these data, along with a defined set of assumptions, including regulatory requirements, it was possible to calculate the waste forms, their radioactivities, volumes and quantities. Quantification of secondary wastes, such as plant maintenance, housekeeping and clean-up wastes, was achieved by reviewing actual operating experience from THORP during its hot operation from 1994 to the present time. This work was carried out under a contract from the United States Department of Energy (DOE) and, so as to enable DOE to make valid comparisons with other similar work, a number of assumptions were agreed. These include an assumed reprocessing capacity of 800 tonnes per year, the requirement to remove as waste forms the volatile fission products carbon-14, iodine-129, krypton-85, tritium and ruthenium-106, the restriction of discharge of any water from the facility unless it meets US Environmental Protection Agency drinking water standards, no intentional blending of wastes to lower their classification, and the requirement for the recovered uranium to be sufficiently free from fission products and neutron-absorbing species to allow it to be re-enriched and recycled as nuclear fuel. The results from this work showed that over 99.9% of the radioactivity in the UNF can be concentrated via reprocessing into a fission-product-containing vitrified product, bottles of compressed krypton storage and a cement grout containing the tritium, that together have a volume of only about one eighth the volume of the original UNF. The other waste forms have larger volumes than the original UNF but contain only the remaining 0.1% of the radioactivity. (authors)« less

Assessing Withering Syndrome Resistance in California Black Abalone: Implications for Conservation and Restoration

DTIC Science & Technology

2009-01-29

worldwide. California Sea Grant College Program UC San Diego Title: Assessing withering syndrome resistance in California black abalone : Implications...Hunter S.(2009). Assessing withering syndrome resistance in California black abalone : Implications for conservation and restoration. UC San Diego...California Sea Grant College Program. Retrieved from: http://escholarship.org/uc/item/7c39q78n Keywords: black abalone , Haliotis cracherodii, captive breeding

Direct measurement of 235U in spent fuel rods with Gamma-ray mirrors

NASA Astrophysics Data System (ADS)

Ruz, J.; Brejnholt, N. F.; Alameda, J. B.; Decker, T. A.; Descalle, M. A.; Fernandez-Perea, M.; Hill, R. M.; Kisner, R. A.; Melin, A. M.; Patton, B. W.; Soufli, R.; Ziock, K.; Pivovaroff, M. J.

2015-03-01

Direct measurement of plutonium and uranium X-rays and gamma-rays is a highly desirable non-destructive analysis method for the use in reprocessing fuel environments. The high background and intense radiation from spent fuel make direct measurements difficult to implement since the relatively low activity of uranium and plutonium is masked by the high activity from fission products. To overcome this problem, we make use of a grazing incidence optic to selectively reflect Kα and Kβ fluorescence of Special Nuclear Materials (SNM) into a high-purity position-sensitive germanium detector and obtain their relative ratios.

Ethical and hygiene aspects of the reprocessing of medical devices in Germany

PubMed Central

Kramer, Axel; Assadian, Ojan

2008-01-01

Based on safety and quality principles, for each medical device (MD), regardless of its declared status as single- or multi-use device, careful considerations must be made. This includes assessment whether reprocessing is economical and ecological meaningful, and technical feasible. So far, however, in Germany reprocessing of declared single use MD is legally allowed, provided that the above aspects are well covered. The purpose of this paper is to elucidate, when circumstances allow reprocessing of declared single-use MD. For reprocessing of single use MD the following preconditions must be fulfilled: The security level of the reprocessed MD must be equivalent to the status of the newly delivered item; this means that a patient is not exposed to a higher risk through a reprocessed disposable MD than through the new, i.e. un-processed product. The reprocessing must be based on a detailed risk assessment and risk analysis, and must be described in detail regarding selection of the reprocessing method. Additionally, all necessary safety- and quality assurance measures must be stated. The reprocessing measure needs to be accompanied with a quality management system which determines and documents the responsibility of all stages of reprocessing; where the corresponding reprocessing procedures are well defined; and the efficacy of the procedure is proven by product-specific or product-group-specific tests and reports. The process must be validated according to recognised methods of science and technology, taking into account potential negative influences of the reprocessing on the properties of the material and the technical and functional safety. For reprocessing of MDs of the category Critical C the quality assurance must be certified by an accredited certifying body. PMID:20204097

Prototype Demonstration of Gamma- Blind Tensioned Metastable Fluid Neutron/Multiplicity/Alpha Detector – Real Time Methods for Advanced Fuel Cycle Applications

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

McDeavitt, Sean M.

The content of this report summarizes a multi-year effort to develop prototype detection equipment using the Tensioned Metastable Fluid Detector (TMFD) technology developed by Taleyarkhan [1]. The context of this development effort was to create new methods for evaluating and developing advanced methods for safeguarding nuclear materials along with instrumentation in various stages of the fuel cycle, especially in material balance areas (MBAs) and during reprocessing of used nuclear fuel. One of the challenges related to the implementation of any type of MBA and/or reprocessing technology (e.g., PUREX or UREX) is the real-time quantification and control of the transuranic (TRU)more » isotopes as they move through the process. Monitoring of higher actinides from their neutron emission (including multiplicity) and alpha signatures during transit in MBAs and in aqueous separations is a critical research area. By providing on-line real-time materials accountability, diversion of the materials becomes much more difficult. The Tensioned Metastable Fluid Detector (TMFD) is a transformational technology that is uniquely capable of both alpha and neutron spectroscopy while being “blind” to the intense gamma field that typically accompanies used fuel – simultaneously with the ability to provide multiplicity information as well [1-3]. The TMFD technology was proven (lab-scale) as part of a 2008 NERI-C program [1-7]. The bulk of this report describes the advancements and demonstrations made in TMFD technology. One final point to present before turning to the TMFD demonstrations is the context for discussing real-time monitoring of SNM. It is useful to review the spectrum of isotopes generated within nuclear fuel during reactor operations. Used nuclear fuel (UNF) from a light water reactor (LWR) contains fission products as well as TRU elements formed through neutron absorption/decay chains. The majority of the fission products are gamma and beta emitters and they represent the more significant hazards from a radiation protection standpoint. However, alpha and neutron emitting uranium and TRU elements represent the more significant safeguards and security concerns. Table 1.1 presents a representative PWR inventory of the uranium and actinide isotopes present in a used fuel assembly. The uranium and actinide isotopes (chiefly the Pu, Am and Cm elements) are all emitters of alpha particles and some of them release significant quantities of neutrons through spontaneous fissions« less

WEST ELEVATION OF REMOTE ANALYTICAL FACILITY (CPP627) AND HOT PILOT ...

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

WEST ELEVATION OF REMOTE ANALYTICAL FACILITY (CPP-627) AND HOT PILOT PLANT (CPP-640) LOOKING NORTHEAST. INL PHOTO NUMBER HD-22-2-1. Mike Crane, Photographer, 11/1998 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

ARCHITECTURAL FLOOR PLAN OF PROCESS AND ACCESS AREAS HOT PILOT ...

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

ARCHITECTURAL FLOOR PLAN OF PROCESS AND ACCESS AREAS HOT PILOT PLANT (CPP-640). INL DRAWING NUMBER 200-0640-00-279-111679. ALTERNATE ID NUMBER 8952-CPP-640-A-2. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP640) OVERALL VIEW ...

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

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP-640) OVERALL VIEW LOOKING SOUTHEAST; CONSTRUCTION 34 PERCENT COMPLETE. INL PHOTO NUMBER NRTS-60-3034. Holmes, Photographer, 6/23/1960 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

NORTH ELEVATION OF HOT PILOT PLANT (CPP640) LOOKING SOUTH AFTER ...

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

NORTH ELEVATION OF HOT PILOT PLANT (CPP-640) LOOKING SOUTH AFTER REMOTE ANALYTICAL FACILITY (CPP-627) WAS REMOVED. PHOTO NUMBER HD-54-33-2. Mike Crane, Photographer, 7/2006 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION VIEW OF MAIN PROCESSING BUILDING (CPP601) ON THE RIGHT ...

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

CONSTRUCTION VIEW OF MAIN PROCESSING BUILDING (CPP-601) ON THE RIGHT AND LABORATORY (CPP-602) ON THE LEFT. INL PHOTO NUMBER NRTS-51-3373. Unknown Photographer, 9/28/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

ARCHITECTURAL ROOF PLAN AND WESTSOUTHEAST ELEVATIONS OF HOT PILOT PLANT ...

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

ARCHITECTURAL ROOF PLAN AND WEST-SOUTHEAST ELEVATIONS OF HOT PILOT PLANT (CPP-640). INL DRAWING NUMBER 200-0640-00-279-111680. ALTERNATE ID NUMBER 8952-CPP-640-A-3. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Process for the extraction of technetium from uranium

DOEpatents

Gong, Cynthia-May S.; Poineau, Frederic; Czerwinski, Kenneth R.

2010-12-21

A spent fuel reprocessing method contacts an aqueous solution containing Technetium(V) and uranyl with an acidic solution comprising hydroxylamine hydrochloride or acetohydroxamic acid to reduce Tc(V) to Tc(II, and then extracts the uranyl with an organic phase, leaving technetium(II) in aqueous solution.

10 CFR 110.41 - Executive Branch review.

Code of Federal Regulations, 2010 CFR

2010-01-01

... export involving assistance to end uses related to isotope separation, chemical reprocessing, heavy water production, advanced reactors, or the fabrication of nuclear fuel containing plutonium, except for exports of... foreign reactor. (8) An export involving radioactive waste. (9) An export to any country listed in § 110...

Uncarboxylated matrix Gla protein (ucMGP) is associated with coronary artery calcification in haemodialysis patients.

PubMed

Cranenburg, Ellen C M; Brandenburg, Vincent M; Vermeer, Cees; Stenger, Melanie; Mühlenbruch, Georg; Mahnken, Andreas H; Gladziwa, Ulrich; Ketteler, Markus; Schurgers, Leon J

2009-02-01

Matrix gamma-carboxyglutamate (Gla) protein (MGP) is a potent local inhibitor of cardiovascular calcification and accumulates at areas of calcification in its uncarboxylated form (ucMGP). We previously found significantly lower circulating ucMGP levels in patients with a high vascular calcification burden. Here we report on the potential of circulating ucMGP to serve as a biomarker for vascular calcification in haemodialysis (HD) patients. Circulating ucMGP levels were measured with an ELISA-based assay in 40 HD patients who underwent multi-slice computed tomography (MSCT) scanning to quantify the extent of coronary artery calcification (CAC). The mean ucMGP level in HD patients (193 +/- 65 nM) was significantly lower as compared to apparently healthy subjects of the same age (441 +/- 97 nM; p < 0.001) and patients with rheumatoid arthritis (RA) without CAC (560 +/- 140 nM; p < 0.001). Additionally, ucMGP levels correlated inversely with CAC scores (r = -0.41; p = 0.009), and this correlation persisted after adjustment for age, dialysis vintage and high-sensitivity C-reactive protein (hs-CRP). Since circulating ucMGP levels are significantly and inversely correlated with the extent of CAC in HD patients, ucMGP may become a tool for identifying HD patients with a high probability of cardiovascular calcification.

A Report Guide to Radiographic Testing Literature. Volume 6

DTIC Science & Technology

1975-04-01

Sources and Applications IITRI, Chicago, 111., 21-22 October 1964, ORNL -11C5, UC-23-Isotopes-Industrial Technology, November 1965 This...Applications IITRI, Chicago, IU., 21-22 October 1964. ORNL -11C5, UC-23lsotopes-lndustria3 Technology November 1965 The design of radioactive sources...Mich. Proceedings of Symposium on Low-Energy X and Gamma Sources and Applications IITRI, Chicago, HI., 21-22 October 1964. ORNL -11C5, US-23-Isotopes

An integrated approach for determining plutonium mass in spent fuel assemblies with nondestructive assay

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

Swinhoe, Martyn T; Tobin, Stephen J; Fensin, Mike L

2009-01-01

There are a variety of reasons for quantifying plutonium (Pu) in spent fuel. Below, five motivations are listed: (1) To verify the Pu content of spent fuel without depending on unverified information from the facility, as requested by the IAEA ('independent verification'). New spent fuel measurement techniques have the potential to allow the IAEA to recover continuity of knowledge and to better detect diversion. (2) To assure regulators that all of the nuclear material of interest leaving a nuclear facility actually arrives at another nuclear facility ('shipper/receiver'). Given the large stockpile of nuclear fuel at reactor sites around the world,more » it is clear that in the coming decades, spent fuel will need to be moved to either reprocessing facilities or storage sites. Safeguarding this transportation is of significant interest. (3) To quantify the Pu in spent fuel that is not considered 'self-protecting.' Fuel is considered self-protecting by some regulatory bodies when the dose that the fuel emits is above a given level. If the fuel is not self-protecting, then the Pu content of the fuel needs to be determined and the Pu mass recorded in the facility's accounting system. This subject area is of particular interest to facilities that have research-reactor spent fuel or old light-water reactor (LWR) fuel. It is also of interest to regulators considering changing the level at which fuel is considered self-protecting. (4) To determine the input accountability value at an electrochemical processing facility. It is not expected that an electrochemical reprocessing facility will have an input accountability tank, as is typical in an aqueous reprocessing facility. As such, one possible means of determining the input accountability value is to measure the Pu content in the spent fuel that arrives at the facility. (5) To fully understand the composition of the fuel in order to efficiently and safely pack spent fuel into a long-term repository. The NDA of spent fuel can be part of a system that cost-effectively meets the burnup credit needs of a repository. Behind each of these reasons is a regulatory structure with MC&A requirements. In the case of the IAEA, the accountable quantity is elemental plutonium. The material in spent fuel (fissile isotopes, fission products, etc.) emits signatures that provide information about the content and history of the fuel. A variety of nondestructive assay (NDA) techniques are available to quantify these signatures. The effort presented in this paper is investigation of the capabilities of 12 NDA techniques. For these 12, none is conceptually capable of independently determining the Pu content in a spent fuel assembly while at the same time being able to detect the diversion of a significant quantity of rods. For this reason the authors are investigating the capability of 12 NDA techniques with the end goal of integrating a few techniques together into a system that is capable of measuring Pu mass in an assembly. The work described here is the beginning of what is anticipated to be a five year effort: (1) two years of modeling to select the best technologies, (2) one year fabricating instruments and (3) two years measuring spent fuel. This paper describes the first two years of this work. In order to cost effectively and robustly model the performance of the 12 NDA techniques, an 'assembly library' was created. The library contains the following: (a) A diverse range of PWR spent fuel assemblies (burnup, enrichment, cooling time) similar to that which exists in spent pools today and in the future. (b) Diversion scenarios that capture a range of possible rod removal options. (c) The spatial and isotopic detail needed to accurately quantify the capability of all the NDA techniques so as to enable integration. It is our intention to make this library available to other researchers in the field for inter-comparison purposes. The performance of each instrument will be quantified for the full assembly library for measurements in three different media: air, water and borated water. The 12 NDA techniques being researched are the following: Delayed Gamma, Delayed Neutrons, Differential Die-Away, Lead Slowing Down Spectrometer, Neutron Multiplicity, Nuclear Resonance Fluorescence, Passive Prompt Gamma, Passive Neutron Albedo Reactivity, Self-integration Neutron Resonance Densitometry, Total Neutron (Gross Neutron), X-Ray Fluorescence, {sup 252}Cf Interrogation with Prompt Neutron Detection.« less

Disposition and metabolism of 2,3-( UC)dichloropropene in rats after inhalation

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

Bond, J.A.; Medinsky, M.A.; Dutcher, J.S.

1985-01-01

2,3-Dichloropropene (2,3-DCP) is a constituent of some commercially available preplant soil fumigants for the control of plant parasitic nematodes. The purpose of this investigation was to determine the disposition and metabolism of 2,3-( UC)DCP in rats after inhalation. Male Fischer-344 rats were exposed nose-only to a vapor concentration of 250 nmol 2,3-( UC)DCP/liter air (7.5 ppm; 25C, 620 Torr) for 6 hr. Blood samples were taken during exposure, and urine, feces, expired air, and tissues were collected for up to 65 hr after exposure. Urinary excretion was the major route of elimination of UC (55% of estimated absorbed 2,3-DCP). Half-timemore » for elimination of UC in urine was 9.8 +/- 0.05 hr (anti x +/- SE). Half-time for elimination of UC feces (17% of absorbed 2,3-DCP) was 12.9 +/- 0.14 hr (anti x +/- SE). Approximately 1 and 3% of the estimated absorbed 2,3-( UC)DCP were exhaled as either 2,3-( UC)DCP or UCO2, respectively. Concentrations of UC in blood increased during 240 min of exposure, after which no further increases in blood concentration of UC were seen. UC was widely distributed in tissues analyzed after a 6-hr exposure of rats to 2,3-( UC)DCP. Urinary bladder (150 nmol/g), nasal turbinates (125 nmol/g), kidneys (84 nmol/g), small intestine (61 nmol/g), and liver (35 nmol/g) were tissues with the highest concentrations of UC immediately after exposure. Over 90% of the UC in tissues analyzed was 2,3-DCP metabolites. Half-times for elimination of UC from tissues examined ranged from 3 to 11 hr. The data from this study indicate that after inhalation 2,3-DCP is metabolized in tissues and readily excreted. 21 references. 2 figures, 4 tables.« less

A meta-analysis of the relationship between MYO9B gene polymorphisms and susceptibility to Crohn's disease and ulcerative colitis.

PubMed

Li, Peng; Yang, Xiao-Ke; Wang, Xiu; Zhao, Meng-Qin; Zhang, Chao; Tao, Sha-Sha; Zhao, Wei; Huang, Qing; Li, Lian-Ju; Pan, Hai-Feng; Ye, Dong-Qing

2016-10-01

Both Crohn's disease (CD) and ulcerative colitis (UC) have a complex etiology involving multiple genetic and environmental factors. Multiple UC and CD susceptibility genes have been identified through genome-wide association studies and subsequent meta-analyses. The aim of this meta-analysis was to clarify the impact of MYO9B gene polymorphisms on CD and UC risk. The PubMed, Elsevier Science Direct and Embase databases were searched to identify eligible studies that were published before October 2014. Data were extracted and pooled crude odds ratios (ORs) and 95% confidence intervals (95% CIs) were calculated. A total of 11 studies, containing 3297 CD cases, 3903 UC cases and 8174 controls were included in this meta-analysis. Bonferroni correction results showed that rs1545620 A/C polymorphism of MYO9B gene was associated with both CD and UC susceptibility in Caucasians (OR=0.88, 95% CI=0.82∼0.95, P=0.001; OR=0.82, 95% CI=0.76∼0.89, P<0.001), but not in Chinese. rs1457092 G/T and rs2305764 C/T polymorphisms are associated with UC susceptibility (OR=0.85, 95% CI=0.79∼0.91, P<0.001; OR=0.88, 95% CI=0.83∼0.93, P<0.001), but not with CD susceptibility in Caucasians. This meta-analysis suggested that rs1545620 is both CD and UC susceptible locus in Caucasians; rs1457092 and rs2305764 are UC susceptible loci, but are not CD susceptible loci in Caucasians. Further studies with more sample size are needed for a definitive conclusion. Copyright © 2016 American Society for Histocompatibility and Immunogenetics. Published by Elsevier Inc. All rights reserved.

Greening Transportation and Parking at University of Coimbra

ERIC Educational Resources Information Center

Cruz, Luís; Barata, Eduardo; Ferreira, João-Pedro; Freire, Fausto

2017-01-01

Purpose: This paper aims to explore the potential contribution of integrated traffic and parking management strategies to ensure more rational use of available parking spaces and to reduce fuel consumption and greenhouse gas emissions by commuters traveling to the University of Coimbra (UC) main campus. Design/methodology/approach: An integrated…

Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR&PP)

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

Moses, David Lewis

2011-10-01

This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR&PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physicalmore » Protection Working Group (PR&PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR&PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR&PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet to be deployed commercially and have only been demonstrated in testing at a laboratory scale.« less

Electrolysis cell for reprocessing plutonium reactor fuel

DOEpatents

Miller, William E.; Steindler, Martin J.; Burris, Leslie

1986-01-01

An electrolytic cell for refining a mixture of metals including spent fuel containing U and Pu contaminated with other metals, the cell including a metallic pot containing a metallic pool as one anode at a lower level, a fused salt as the electrolyte at an intermediate level and a cathode and an anode basket in spaced-apart positions in the electrolyte with the cathode and anode being retractable to positions above the electrolyte during which spent fuel may be added to the anode basket and the anode basket being extendable into the lower pool to dissolve at least some metallic contaminants, the anode basket containing the spent fuel acting as a second anode when in the electrolyte.

Electrolysis cell for reprocessing plutonium reactor fuel

DOEpatents

Miller, W.E.; Steindler, M.J.; Burris, L.

1985-01-04

An electrolytic cell for refining a mixture of metals including spent fuel containing U and Pu contaminated with other metals is claimed. The cell includes a metallic pot containing a metallic pool as one anode at a lower level, a fused salt as the electrolyte at an intermediate level and a cathode and an anode basket in spaced-apart positions in the electrolyte with the cathode and anode being retractable to positions above the electrolyte during which spent fuel may be added to the anode basket. The anode basket is extendable into the lower pool to dissolve at least some metallic contaminants; the anode basket contains the spent fuel acting as a second anode when in the electrolyte.

Studies of thermionic materials for space power applications

NASA Technical Reports Server (NTRS)

1972-01-01

The effect of microstructures of tungsten cladding on the transport rates of carbide fuel components was studied at 2073 K. hyperstoichiometric 90UC-10ZrC containing 4 wt% tungsten was clad with six types of tungsten material of 40 mil thickness. Screening tests of 1000 hours were carried out, and then selected samples were subjected to long-term tests up to 10,000 hours. The results indicate that the microstructures strongly affect the transport rates of carbide fuel components. The conditions for preparing (110) oriented cylindrical chloride tungsten emitters of high vacuum work functions were also investigated. Specimen sets were deposited on fluoride tungsten substrates for evaluating the effects of various deposition parameters on the degree and uniformity of the (110) preferred orientation and the vacuum work function. Long-term tests showed that the high vacuum work function of a cylindrical emitter was stable and the chloride tungsten to fluoride tungsten bond remained in excellent shape after 4850 hours at 2073 K.

Umbilical cord mesenchymal stromal cell transplantations: A systemic analysis of clinical trials.

PubMed

Can, Alp; Celikkan, Ferda Topal; Cinar, Ozgur

2017-12-01

The advances and success of umbilical cord-derived mesenchymal stromal cells (UC-MSCs) in experimental disease animal models have fueled the development of targeted therapies in humans. The therapeutic potential of allogeneic transplantation of UC-MSCs has been under examination since 2009. The purpose of this systematic analysis was to review the published results, limitations and obstacles for UC-MSC transplantation. An extensive search strategy was applied to the published literature, 93 peer-reviewed full-text articles and abstracts were found published by early August 2017 that investigated the safety, efficacy and feasibility of UC-MSCs in 2001 patients with 53 distinct pathologies including many systemic/local, acute/chronic conditions. Few data were extracted from the abstracts and/or Chinese-written articles (n = 7, 8%). Importantly, no long-term adverse effects, tumor formation or cell rejection were reported. All studies noted certain degrees of therapeutic benefit as evidenced by clinical symptoms and/or laboratory findings. Thirty-seven percent (n = 34) of studies were found published as a single case (n = 10; 11%) or 2-10 case reports (n = 24; 26%) with no control group. Due to the nature of many stem cell-based studies, the majority of patients also received conventional therapy regimens, which obscured the pure efficacy of the cells transplanted. Randomized, blind, phase 1/2 trials with control groups (placebo-controlled) showed more plausible results. Given that most UC-MSC trials are early phase, the internationally recognized cell isolation and preparation standards should be extended to future phase 2/3 trials to reach more convincing conclusions regarding the safety and efficacy of UC-MSC therapies. Copyright © 2017 International Society for Cellular Therapy. Published by Elsevier Inc. All rights reserved.

Celecoxib-Induced Cytotoxic Effect Is Potentiated by Inhibition of Autophagy in Human Urothelial Carcinoma Cells

PubMed Central

Ho, I-Lin; Chang, Hong-Chiang; Chuang, Yuan-Ting; Lin, Wei-Chou; Lee, Ping-Yi; Chang, Shih-Chen; Chiang, Chih-Kang; Pu, Yeong-Shiau; Chou, Chien-Tso; Hsu, Chen-Hsun; Liu, Shing-Hwa

2013-01-01

Celecoxib, a cyclooxygenase-2 (COX-2) inhibitor, can elicit anti-tumor effects in various malignancies. Here, we sought to clarify the role of autophagy in celecoxib-induced cytotoxicity in human urothelial carcinoma (UC) cells. The results shows celecoxib induced cellular stress response such as endoplasmic reticulum (ER) stress, phosopho-SAPK/JNK, and phosopho-c-Jun as well as autophagosome formation in UC cells. Inhibition of autophagy by 3-methyladenine (3-MA), bafilomycin A1 or ATG7 knockdown potentiated celecoxib-induced apoptosis. Up-regulation of autophagy by rapamycin or GFP-LC3B-transfection alleviated celecoxib-induced cytotoxicity in UC cells. Taken together, the inhibition of autophagy enhances therapeutic efficacy of celecoxib in UC cells, suggesting a novel therapeutic strategy against UC. PMID:24349176

Interleukin 1 β (IL-1B) and IL-1 antagonist receptor (IL-1RN) gene polymorphisms are associated with the genetic susceptibility and steroid dependence in patients with ulcerative colitis.

PubMed

Yamamoto-Furusho, Jesús K; Santiago-Hernández, Jean J; Pérez-Hernández, Nonanzit; Ramírez-Fuentes, Silvestre; Fragoso, José Manuel; Vargas-Alarcón, Gilberto

2011-07-01

Ulcerative colitis (UC) is an inflammatory bowel disease of unknown etiology. Among cytokines induced in UC, interleukin 1 antagonist (IL-1ra) and interleukin 1 β (IL-1β) seems to have a central role because of its immunoregulatory and proinflammatory activities. To determine the association between IL-1RA and IL-1B gene polymorphisms and the clinical features of UC in the Mexican Mestizo population. Five polymorphisms in the IL-1 gene cluster members IL-1B (rs16944), IL1F10 (rs3811058), and IL-1RN (rs419598, rs315952, and rs315951) were genotyped by 5' exonuclease TaqMan genotyping assays in a group of 200 Mexican patients with UC and 248 ethnically matched unrelated healthy controls. We found a significant increased frequencies of IL-1RN6/1 TC (rs315952) and RN6/2 CC (rs315951) and decreased frequency of IL-1B-511 TC (rs16944) genotypes in UC patients as compared with healthy controls. In the subgroup analysis, we found a significant association between the RN6/2 GG (rs315951) and IL-1B-511 CC (rs16944) genotypes and the presence of steroid-dependence in UC patients (pC=00001, OR=15.6 and pC=0.008, OR=4.09, respectively). Patients with UC showed increased frequencies of IL-1RN "CTC" and "TCG" haplotypes when compared with healthy controls (P=0.019, OR=1.43 and P<10(-7), OR=2.63, respectively). Two haplotypes (TTG and CTG) showed decreased frequency in patients when compared with healthy controls (P=9×10(-7), OR=0.11 and P=8×10(-6), OR=0.11, respectively). IL-1 RN and IL-1B polymorphisms were associated with the genetic susceptibility to develop UC and might be associated with the presence of steroid-dependence in UC patients.

Air Force Office of Scientific Research Technical Report Summaries January - March 1991.

DTIC Science & Technology

1991-04-01

SUS 4SU Uc m W1( -W 0x t - 030 L O F-C-UC W QIMS O-UC tN -C 0 Lo - in 4 LC 4--- 41 UU 00W142 U4 (a 4- C- L 0) -0*-LU0 u ZaW inUi x& Ii- .0 5C4 L US)-L...C 3 L4-in > L 1 4 0 0 4) - DUC 0 L -0M Li nnU 0 3O> C U 0IX 9 z L f -) i- njo e CL .-- w* 1- - .I& - tUn I-. (A in 0 CQ4O 3 041 M -CC 0- LO A A L0...0 ’ 1- - -4-0 -V40 A -- 0 -- W;z a-a 0 Sn C’) 4A OLD tun , z 1’.S oEV mmU OW =3 ) ml > 23 S3 N S 01 1 54 V -0 a0 5 -0 .<Z - 0 4U.~ >jI~S𔃺~.C 42 r - C

Theoretical study of actinide monocarbides (ThC, UC, PuC, and AmC)

NASA Astrophysics Data System (ADS)

Pogány, Peter; Kovács, Attila; Visscher, Lucas; Konings, Rudy J. M.

2016-12-01

A study of four representative actinide monocarbides, ThC, UC, PuC, and AmC, has been performed with relativistic quantum chemical calculations. The two applied methods were multireference complete active space second-order perturbation theory (CASPT2) including the Douglas-Kroll-Hess Hamiltonian with all-electron basis sets and density functional theory with the B3LYP exchange-correlation functional in conjunction with relativistic pseudopotentials. Beside the ground electronic states, the excited states up to 17 000 cm-1 have been determined. The molecular properties explored included the ground-state geometries, bonding properties, and the electronic absorption spectra. According to the occupation of the bonding orbitals, the calculated electronic states were classified into three groups, each leading to a characteristic bond distance range for the equilibrium geometry. The ground states of ThC, UC, and PuC have two doubly occupied π orbitals resulting in short bond distances between 1.8 and 2.0 Å, whereas the ground state of AmC has significant occupation of the antibonding orbitals, causing a bond distance of 2.15 Å.

New approaches for MOX multi-recycling

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

Gain, T.; Bouvier, E.; Grosman, R.

Due to its low fissile content after irradiation, Pu from used MOX fuel is considered by some as not recyclable in LWR (Light Water Reactors). The point of this paper is hence to go back to those statements and provide a new analysis based on AREVA extended experience in the fields of fissile and fertile material management and optimized waste management. This is done using the current US fuel inventory as a case study. MOX Multi-recycling in LWRs is a closed cycle scenario where U and Pu management through reprocessing and recycling leads to a significant reduction of the usedmore » assemblies to be stored. The recycling of Pu in MOX fuel is moreover a way to maintain the self-protection of the Pu-bearing assemblies. With this scenario, Pu content is also reduced repetitively via a multi-recycling of MOX in LWRs. Simultaneously, {sup 238}Pu content decreases. All along this scenario, HLW (High-Level Radioactive Waste) vitrified canisters are produced and planned for deep geological disposal. Contrary to used fuel, HLW vitrified canisters do not contain proliferation materials. Moreover, the reprocessing of used fuel limits the space needed on current interim storage. With MOX multi-recycling in LWR, Pu isotopy needs to be managed carefully all along the scenario. The early introduction of a limited number of SFRs (Sodium Fast Reactors) can therefore be a real asset for the overall system. A few SFRs would be enough to improve the Pu isotopy from used LWR MOX fuel and provide a Pu-isotopy that could be mixed back with multi-recycled Pu from LWRs, hence increasing the Pu multi-recycling potential in LWRs.« less

INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP601) ...

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

INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP-601) LOOKING SOUTHWEST. PHOTO TAKEN FROM NORTHEAST CORNER. INL PHOTO NUMBER HD-50-4-2. Mike Crane, Photographer, 6/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

FIRST FLOOR PLAN OF REMOTE ANALYTICAL FACILITY (CPP627) SHOWING REMOTE ...

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

FIRST FLOOR PLAN OF REMOTE ANALYTICAL FACILITY (CPP-627) SHOWING REMOTE ANALYTICAL LABORATORY, DECONTAMINATION ROOM, AND MULTICURIE CELL ROOM. INL DRAWING NUMBER 200-0627-00-008-105065. ALTERNATE ID NUMBER 4272-14-102. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP601) ...

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

INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP-601) LOOKING NORTH. PHOTO TAKEN FROM SOUTHWEST CORNER. INL PHOTO NUMBER HD-50-1-3. Mike Crane, Photographer, 6/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP640) LOOKING NORTHEAST ...

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

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP-640) LOOKING NORTHEAST SHOWING OVERALL BLOCK EXTERIOR WALLS; CONSTRUCTION 65 PERCENT COMPLETE. INL PHOTO NUMBER NRTS-60-4976. Holmes, Photographer, 9/26/1960 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP640) LOOKING EAST ...

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

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP-640) LOOKING EAST SHOWING EXCAVATION AND FORMING; CONSTRUCTION 6 PERCENT COMPLETE. INL PHOTO NUMBER NRTS-59-4935. J. Anderson, Photographer, 9/21/1959 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

INTERIOR PHOTO OF MAIN PROCESSING BUILDING (CPP601) PROCESS MAKEUP AREA ...

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

INTERIOR PHOTO OF MAIN PROCESSING BUILDING (CPP-601) PROCESS MAKEUP AREA LOOKING SOUTH. PHOTO TAKEN FROM CENTER OF WEST WALL. INL PHOTO NUMBER HD-50-1-4. Mike Crane, Photographer, 6/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

NORTH AND WEST ELEVATIONS OF REMOTE ANALYTICAL FACILITY (CPP627) LOOKING ...

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

NORTH AND WEST ELEVATIONS OF REMOTE ANALYTICAL FACILITY (CPP-627) LOOKING SOUTHEAST. HEADEND PLANT (CPP-640) APPEARS IN THE BACKGROUND. INL PHOTO NUMBER HD-22-1-4. Mike Crane, Photographer, 11/1998 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

INTERIOR PHOTO OF HOT PILOT PLANT SECOND FLOOR DEPICTING DETAIL ...

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

INTERIOR PHOTO OF HOT PILOT PLANT SECOND FLOOR DEPICTING DETAIL OF SHIELDED CAVE (CPP-640) LOOKING SOUTHWEST. PHOTO TAKEN FROM NORTH. INL PHOTO NUMBER HD-54-40-2. Mike Crane, Photographer, 7/2006 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

PLAN SECTIONS AND ELEVATIONS OF VESSEL SAMPLING STATIONS "P", "Q", ...

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

PLAN SECTIONS AND ELEVATIONS OF VESSEL SAMPLING STATIONS "P", "Q", "S" CELLS MAIN PROCESSING BUILDING (CPP-601). INL DRAWING NUMBER 200-0601-00-291-053694. ALTERNATE ID NUMBER CPP-E-1394. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

EAST ELEVATION OF MAIN PROCESSING BUILDING (CPP601) LOOKING NORTHWEST. MAINTENANCE ...

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

EAST ELEVATION OF MAIN PROCESSING BUILDING (CPP-601) LOOKING NORTHWEST. MAINTENANCE SHOP AND OFFICE BUILDING (CPP-630) ON RIGHT IN PHOTO. INL PHOTO NUMBER HD-22-3-2. Mike Crane, Photographer, 11/1998 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

INTERIOR PHOTO OF HOT PILOT PLANT SECOND FLOOR WITH SOUTH ...

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

INTERIOR PHOTO OF HOT PILOT PLANT SECOND FLOOR WITH SOUTH SECTION OF SHIELDED CAVE IN FOREGROUND (CPP-640) LOOKING NORTHWEST. INL PHOTO NUMBER HD-54-40-1. Mike Crane, Photographer, 7/2006 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

EQUIPMENT LAYOUT OF MAIN PROCESSING BUILDING (CPP601) LCELL PLAN AND ...

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

EQUIPMENT LAYOUT OF MAIN PROCESSING BUILDING (CPP-601) L-CELL PLAN AND SECTION SHOWS COMPLEXITY OF CELLS. INL DRAWING NUMBER 200-0601-00-098-105687. ALTERNATE ID NUMBER 4289-20-301. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP601) ...

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

INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP-601) LOOKING NORTHWEST. PHOTO TAKEN FROM MIDDLE OF CORRIDOR. INL PHOTO NUMBER HD-50-2-3. Mike Crane, Photographer, 6/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP601) ...

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

INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP-601) LOOKING SOUTH. PHOTO TAKEN FROM MIDDLE OF CORRIDOR. INL PHOTO NUMBER HD-50-3-2. Mike Crane, Photographer, 6/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

10 CFR Appendix D to Subpart D of... - Classes of Actions That Normally Require EISs

Code of Federal Regulations, 2010 CFR

2010-01-01

... average megawatts or more over a 12 month period. This applies to power marketing operations and to siting... Systems D2. Siting/construction/operation/decommissioning of nuclear fuel reprocessing facilities D3. Siting/construction/operation/decommissioning of uranium enrichment facilities D4. Siting/construction...

Experimental studies at the Idaho Chemical Processing Plant on actinide partitioning from acidic nuclear wastes

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

McIssaac, L. D.; Baker, J. D.; Meikrantz, D. H.

1980-01-01

Wastes generated at ICPP and in the reprocessing of LWR fuel is discussed separately. DHDECMP is used as extractant. Studies on DHDECMP purification and toxicity, diluent effects, reaction kinetics, radioloysis, mixer-settler performance, etc. are reported. 10 tables, 3 figures. (DLC)

10 CFR Appendix B to Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

Code of Federal Regulations, 2012 CFR

2012-01-01

... executing. III. Design Control Measures shall be established to assure that applicable regulatory... control of design interfaces and for coordination among participating design organizations. These measures..., approval, release, distribution, and revision of documents involving design interfaces. The design control...

10 CFR Appendix B to Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

Code of Federal Regulations, 2013 CFR

2013-01-01

... executing. III. Design Control Measures shall be established to assure that applicable regulatory... control of design interfaces and for coordination among participating design organizations. These measures..., approval, release, distribution, and revision of documents involving design interfaces. The design control...

New measurement of the 242Pu(n,γ) cross section at n_TOF

NASA Astrophysics Data System (ADS)

Lerendegui-Marco, J.; Guerrero, C.; Cortés-Giraldo, M. A.; Quesada, J. M.; Mendoza, E.; Cano-Ott, D.; Eberhardt, K.; Junghans, A.

2016-03-01

The use of MOX fuel (mixed-oxide fuel made of UO2 and PuO2) in nuclear reactors allows substituting a large fraction of the enriched Uranium by Plutonium reprocessed from spent fuel. With the use of such new fuel composition rich in Pu, a better knowledge of the capture and fission cross sections of the Pu isotopes becomes very important. In particular, a new series of cross section evaluations have been recently carried out jointly by the European (JEFF) and United States (ENDF) nuclear data agencies. For the case of 242Pu, the two only neutron capture time-of-flight measurements available, from 1973 and 1976, are not consistent with each other, which calls for a new time-of flight capture cross section measurement. In order to contribute to a new evaluation, we have perfomed a neutron capture cross section measurement at the n_TOF-EAR1 facility at CERN using four C6D6 detectors, using a high purity target of 95 mg. The preliminary results assessing the quality and limitations (background, statistics and γ-flash effects) of this new experimental data are presented and discussed, taking into account that the aimed accuracy of the measurement ranges between 7% and 12% depending on the neutron energy region.

Radiocarbon behaviour in seawater and the brown algae Fucus serratus in the vicinity of the COGEMA La Hague spent fuel reprocessing plant (Goury)--France.

PubMed

Douville, Eric; Fiévet, Bruno; Germain, Pierre; Fournier, Marc

2004-01-01

Extensive studies of the radiocarbon (14C) distribution and transfer in the marine environment of the North-Cotentin peninsula and along the English Channel have been carried out. The main aims of these studies have been to estimate the spatial and temporal variation of the 14C concentration in seawater and to calculate 14C concentration factors for some biological species. Such information will be helpful in order to calculate precisely radiation doses to humans. First results obtained in the vicinity of the COGEMA La Hague nuclear plant (Goury) indicate a 14C labelling of the dissolved inorganic carbon (DIC) in seawater (8.0-26.2 Bq.m(-3)) and a tight relationship between the 14C in the liquid releases from the plant and the 14C concentrations in DIC. The particulate organic carbon (POC) is also labelled. The concentration factor calculations for the brown algae (Fucus serratus) sampled from Goury, and also along the English Channel, give 14C values around 3000 Bq.kg(-1) fresh weight / Bq.L(-1).

Efficacy and safety of glycosylated undenatured type-II collagen (UC-II) in therapy of arthritic dogs.

PubMed

Deparle, L A; Gupta, R C; Canerdy, T D; Goad, J T; D'Altilio, M; Bagchi, M; Bagchi, D

2005-08-01

DeParle L. A., Gupta R. C., Canerdy T. D., Goad J. T., D'Altilio M., Bagchi M., Bagchi D. Efficacy and safety of glycosylated undenatured type-II collagen (UC-II) in therapy of arthritic dogs. J. vet. Pharmacol. Therap.28, 385-390. In large breed dogs, arthritis is very common because of obesity, injury, aging, immune disorder, or genetic predispositions. This study was therefore undertaken to evaluate clinical efficacy and safety of undenatured type-II collagen (UC-II) in obese-arthritic dogs. Fifteen dogs in three groups received either no UC-II (Group I) or UC-II with 1 mg/day (Group II) or 10 mg/day (Group III) for 90 days. Lameness and pain were measured on a weekly basis for 120 days (90 days treatment plus 30 days post-treatment). Blood samples were assayed for creatinine and blood urea nitrogen (markers of renal injury); and alanine aminotransferase and aspartate aminotransferase (evidence of hepatic injury). Dogs receiving 1 mg or 10 mg UC-II/day for 90 days showed significant declines in overall pain and pain during limb manipulation and lameness after physical exertion, with 10 mg showed greater improvement. At either dose of UC-II, no adverse effects were noted and no significant changes were noted in serum chemistry, suggesting that UC-II was well tolerated. In addition, dogs receiving UC-II for 90 days showed increased physical activity level. Following UC-II withdrawal for a period of 30 days, all dogs experienced a relapse of overall pain, exercise-associated lameness, and pain upon limb manipulation. These results suggest that daily treatment of arthritic dogs with UC-II ameliorates signs and symptoms of arthritis, and UC-II is well tolerated as no adverse effects were noted.

Polymorphism of inflammatory genes and arsenic methylation capacity are associated with urothelial carcinoma

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

Wu, Chia-Chang; Department of Urology, Taipei Medical University—Shuang Ho Hospital, Taipei, Taiwan; Huang, Yung-Kai

2013-10-01

Chronic exposure to arsenic can generate reactive oxidative species, which can induce certain proinflammatory cytokines such as tumor necrosis factor-alpha (TNF-α), interleukin-6 (IL-6) and interleukin-8 (IL-8). TNF-α, IL-6 and IL-8 have been shown to be involved in the development and progression of various cancers, including bladder cancer. This study aimed to investigate the joint effect of the polymorphism of TNF-α − 308 G/A, IL-6 − 174 G/C, IL-8 − 251 T/A and urinary arsenic profiles on urothelial carcinoma (UC) risk. This study evaluated 300 pathologically-confirmed cases of UC and 594 cancer-free controls. Urinary arsenic species were detected using high-performance liquidmore » chromatography-linked hydride generator and atomic absorption spectrometry. The polymorphism of TNF-α − 308 G/A, IL-6 − 174 G/C and IL-8 − 251 T/A was determined using polymerase chain reaction-restriction fragment length polymorphism. The joint effects on UC risk were estimated by odds ratios and 95% confidence intervals using unconditional logistic regression. We found that the TNF-α − 308 A/A and IL-8 − 251 T/T polymorphisms were significantly associated with UC. Moreover, significant dose–response joint effect of TNF-α − 308 A/A or IL-8 − 251 T/T genotypes and arsenic methylation indices were seen to affect UC risk. The present results also showed a significant increase in UC risk in subjects with the IL-8 − 251 T/T genotype for each SD increase in urinary total arsenic and MMA%. In contrast, a significant decrease in UC risk was found in subjects who carried the IL-8 − 251 T/T genotype for each SD increase in DMA%. - Highlights: • Joint effect of the TNF-α -308 A/A genotype and urinary total arsenic affected UC. • Joint effect of the IL-8 -251 T/T genotype and urinary total arsenic affected UC. • Urinary total arsenic level, TNF-α -308 A/A and IL-8 -251 T/T genotype affected UC.« less

Tn4556, a 6.8-kilobase-pair transposable element of Streptomyces fradiae.

PubMed Central

Chung, S T

1987-01-01

A 6.8-kilobase-pair (kbp) transposable element (Tn4556) was found in a neomycin-producing strain of Streptomyces fradiae. This element was first observed in two 30.3-kbp plasmids (pUC1123 and pUC1124) which arose when a thiostrepton resistance gene (1 kbp) was ligated with the BclI-2 fragment (22.5 kbp) that contains the origin of replication of phage SF1. The Tn4556 segment was deleted when these plasmids were transduced into another S. fradiae host with phage SF1. These deletion plasmids (pUC1210 and pUC1211) had copy numbers of less than 1 per chromosome and were unstable. In contrast, pUC1123 and pUC1124, with copy numbers of 12 to 15 per chromosome, respectively, were relatively stable. When pUC1210 and pUC1211 were reintroduced into S. fradiae by protoplast transformation, the Tn4556 element transposed again to the plasmids at numerous new locations in either of two orientations. A copy of Tn4556 was found in the S. fradiae chromosome by hybridization studies. It appears that Tn4556 originated from the chromosome, transposed into unstable pUC1210 and pUC1211, and made stable plasmids. A temperature-sensitive hybrid plasmid carrying a viomycin resistance derivative of Tn4556 (pMT660::Tn4556::vph) was constructed. When Streptomyces lividans UC8390 containing the hybrid plasmid was grown at 39 degrees C, Tn4556::vph (Tn4560) transposed to random positions in the host chromosome. Images PMID:2820925

Use of apigenin from Cordia dichotoma in the treatment of colitis.

PubMed

Ganjare, Anjali B; Nirmal, Sunil A; Patil, Anuja N

2011-10-01

Cordia dichotoma f. (Boraginaceae) is a small deciduous tree from India. The bark of was used in the treatment of ulcerative colitis (UC) and colic pain traditionally hence present work was undertaken to identify the phytoconstituent responsible for this activity. Apigenin is isolated by column chromatography from methanol fraction of crude methanol extract of C. dichotoma bark. Structure of apigenin is established by various spectroscopic studies. Apigenin (5mg/kg, p.o.) showed significant healing and reduction in inflammatory enzymes when screened for UC. It can be concluded that apigenin from C. dichotoma bark may be responsible for the treatment of UC. Copyright © 2011 Elsevier B.V. All rights reserved.

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

Casella, Amanda J.; Hylden, Laura R.; Campbell, Emily L.

Knowledge of real-time solution properties and composition is a necessity for any spent nuclear fuel reprocessing method. Metal-ligand speciation in aqueous solutions derived from the dissolved commercial spent fuel is highly dependent upon the acid concentration/pH, which influences extraction efficiency and the resulting speciation in the organic phase. Spectroscopic process monitoring capabilities, incorporated in a counter current centrifugal contactor bank, provide a pathway for on-line real-time measurement of solution pH. The spectroscopic techniques are process-friendly and can be easily configured for on-line applications, while classic potentiometric pH measurements require frequent calibration/maintenance and have poor long-term stability in aggressive chemical andmore » radiation environments. Our research is focused on developing a general method for on-line determination of pH of aqueous solutions through chemometric analysis of Raman spectra. Interpretive quantitative models have been developed and validated under the range of chemical composition and pH using a lactic acid/lactate buffer system. The developed model was applied to spectra obtained on-line during solvent extractions performed in a centrifugal contactor bank. The model predicted the pH within 11% for pH > 2, thus demonstrating that this technique could provide the capability of monitoring pH on-line in applications such as nuclear fuel reprocessing.« less

Sources of the transuranic elements plutonium and neptunium in arctic marine sediments.

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

Cooper, L. W.; Kelley, J. M.; Bond, L. A.

2000-01-01

We report here thermal ionization mass spectrometry measurements of {sup 239}Pu, {sup 240}Pu, {sup 241}Pu, {sup 242}Pu, and {sup 237}Np isolated from oceanic, estuarine, and riverine sediments from the Arctic Ocean Basin. {sup 238}Pu/{sup 239+240}Pu activity ratios are also reported for alpha spectrometric analyses undertaken on a subset of these samples. Our results indicate that the Pu in sediments on the Alaskan shelf and slope, as well as that in the deep basins (Amerasian and Eurasian) of the Arctic Ocean, has its origin in stratospheric and tropospheric fallout. Sediments from the Ob and Yenisei Rivers show isotopic Pu signatures thatmore » are distinctly different from those of northern-hemisphere stratospheric fallout and indicate the presence of weapons-grade Pu originating from nuclear fuel reprocessing wastes generated at Russian facilities within these river catchments. Consequently, sediments of the Eurasian Arctic Ocean, particularly those in the Barents and Kara Seas, probably contain a mixture of Pu from stratospheric fallout, tropospheric fallout, and fuel-reprocessing wastes of riverine origin. In particular, the {sup 241}Pu/{sup 239}Pu ratios observed in these sediments are inconsistent with significant contributions of Pu to the arctic sediments studied from western European reprocessing facilities, principally Sellafield in the UK. Several other potential sources of Pu to arctic sediments can also be excluded as significant based upon the transuranic isotope ratios presented.« less

Aging of Iodine-Loaded Silver Mordenite in NO2

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

Bruffey, Stephanie H.; Jubin, Robert Thomas; Patton, Kaara K.

2014-04-01

Used nuclear fuel facilities need to control and minimize radioactive emissions. Off-gas systems are designed to remove radioactive contaminants, such as 85Kr, 14C, 3H, and 129I. In an off-gas system, any capture material will be exposed to a gas stream for months at a time. This gas stream may be at elevated temperature and could contain water, NOx gas, or a variety of other constituents comprising the dissolver off-gas stream in a nuclear fuel reprocessing plant. For this reason, it is important to evaluate the effects of long-term exposure, or aging, on proposed capture materials. One material under consideration ismore » reduced silver mordenite (Ag0Z), which is recognized for its efficient iodine capture properties. Iodine is immobilized on Ag0Z as AgI, a solid with low volatility (m.p. ≥ 500°C). The aim of this study was to determine whether extended aging at elevated temperature in a nominally 2% NO2 environment would result in a loss of immobilized iodine from this material due to either physical or chemical changes that might occur during aging. Charges of iodine-loaded reduced silver mordenite (I2-Ag0Z) were exposed to a 2% NO2 environment for 1, 2, 3, and 4 months at 150°C, then analyzed for iodine losses The aging study was completed successfully. The material did not visibly change color or form. The results demonstrate that no significant iodine loss was observed over the course of 4 months of 2% NO2 aging of I2-Ag0Z at elevated temperature within the margin of error and the variability (~10%) in the loading along the beds. This provides assurance that iodine will remain immobilized on Ag0Z during extended online use in an off-gas capture treatment system. Future tests should expose I2-Ag0Z to progressively more complex feed gases in an effort to accurately replicate the conditions expected in a reprocessing facility.« less

Characterization of Used Nuclear Fuel with Multivariate Analysis for Process Monitoring

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

Dayman, Kenneth J.; Coble, Jamie B.; Orton, Christopher R.

2014-01-01

The Multi-Isotope Process (MIP) Monitor combines gamma spectroscopy and multivariate analysis to detect anomalies in various process streams in a nuclear fuel reprocessing system. Measured spectra are compared to models of nominal behavior at each measurement location to detect unexpected changes in system behavior. In order to improve the accuracy and specificity of process monitoring, fuel characterization may be used to more accurately train subsequent models in a full analysis scheme. This paper presents initial development of a reactor-type classifier that is used to select a reactor-specific partial least squares model to predict fuel burnup. Nuclide activities for prototypic usedmore » fuel samples were generated in ORIGEN-ARP and used to investigate techniques to characterize used nuclear fuel in terms of reactor type (pressurized or boiling water reactor) and burnup. A variety of reactor type classification algorithms, including k-nearest neighbors, linear and quadratic discriminant analyses, and support vector machines, were evaluated to differentiate used fuel from pressurized and boiling water reactors. Then, reactor type-specific partial least squares models were developed to predict the burnup of the fuel. Using these reactor type-specific models instead of a model trained for all light water reactors improved the accuracy of burnup predictions. The developed classification and prediction models were combined and applied to a large dataset that included eight fuel assembly designs, two of which were not used in training the models, and spanned the range of the initial 235U enrichment, cooling time, and burnup values expected of future commercial used fuel for reprocessing. Error rates were consistent across the range of considered enrichment, cooling time, and burnup values. Average absolute relative errors in burnup predictions for validation data both within and outside the training space were 0.0574% and 0.0597%, respectively. The errors seen in this work are artificially low, because the models were trained, optimized, and tested on simulated, noise-free data. However, these results indicate that the developed models may generalize well to new data and that the proposed approach constitutes a viable first step in developing a fuel characterization algorithm based on gamma spectra.« less

A Pebble-Bed Breed-and-Burn Reactor

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

Greenspan, Ehud

2016-03-31

The primary objective of this project is to use three-dimensional fuel shuffling in order to reduce the minimum peak radiation damage of ~550 dpa present Breed-and-Burn (B&B) fast nuclear reactor cores designs (they feature 2-D fuel shuffling) call for to as close as possible to the presently accepted value of 200 dpa thereby enabling earlier commercialization of B&B reactors which could make substantial contribution to energy sustainability and economic stability without need for fuel recycling. Another objective is increasing the average discharge burnup for the same peak discharge burnup thereby (1) increasing the fuel utilization of 2-D shuffled B&B reactorsmore » and (2) reducing the reprocessing capacity required to support a given capacity of FRs that are to recycle fuel.« less

The scheme for evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Saldikov, I. S.; Ternovykh, M. Yu; Fomichenko, P. A.; Gerasimov, A. S.

2017-01-01

The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of power. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. To solve the closed nuclear fuel modeling tasks REPRORYV code was developed. It simulates the mass flow for nuclides in the closed fuel cycle. This paper presents the results of modeling of a closed nuclear fuel cycle, nuclide flows considering the influence of the uncertainty on the outcome of neutron-physical characteristics of the reactor.

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

Rabin, S.A.; Martin, M.M.; Lotts, A.L.

The fabricability of dispersion fuels using UO/sub 2/ or UC as the dispersoid and uranium combined with 10 to 15 wt% Mo as the matrix was investigated. Cores containing l7.8 wt% UO/sub 2/ dispersed in U-- 15 wt.% Mo were successfully fabricated to about 80% of theoretical density by cold pressing at 50 tsi, sintering at 1100 deg C, and cold coining at 50 tsi. Comparable results were obtained with UC as the dispersoid. Core fabrication results varied greatly with the type of matrix powder used. Occluded gases, pour density, and surface cleanliness bore important relations to the fabrication behaviormore » of powders. Suitable pressing and sintering results were obtained with prealloyed, calcium-reduced U--Mo powder and with molybdenum and calcium-reduced uranium as elemental powders. Shotted prealloyed powders were difficult to press and sinter, as were elemental and prealloyed powders prepared by hydriding. The cores containing UO/sub 2/ were picture-frame, hot-roll-clad as miniature plates. Molybdenum, Fansteel 82, and Zr--3 wt% Al were investigated as cladding materials. While each bonded well to itself, only the molybdenum-clad core, rolled at 1150 deg C to 10/1 reduction, resulted in dispersions free of ruptures and UO/sub 2/ fragmentation and in strong bonding to the core, evaluated by metallography, mechanical peel, and thermal shock tests. The matrix phase was homogeneous, but the UO/sub 2/ dispersoid showed stringering characteristic of cores worked by hot rolling. Core densities as high as 99% of theoretical were obtained. (auth)« less

THE EFFECT OF TEMPERATURE AND UNIAXIAL PRESSURE ON THE DENSIFICATION BEHAVIOR OF SILICA AEROGEL GRANULES

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

Matyas, Josef; Robinson, Matthew J.; Fryxell, Glen E.

Materials are being developed in U.S. for the removal and immobilization of iodine from gaseous products of nuclear fuel reprocessing in support of the Fuel Cycle Technology Separations and Waste Forms Campaign. The silver-functionalized silica aerogel proved to be an excellent candidate for this treatment because of its high selectivity and sorption capacity for radioiodine and its possible conversion to a durable silica-based waste form. The present study investigated with nitrogen sorption and helium pycnometry the effect of pressureless isothermal sintering at temperatures of 900-1400°C for 2.5-90 min or isothermal hot-pressing at 1200°C for 2.5 min on densification of rawmore » and silver-functionalized silica aerogel granules. Rapid sintering was observed at 1050 and 1200°C. Only 15 min of pressureless sintering at 1200°C resulted in almost complete densification. The macropores disappeared, surface area decreased from 1114 m2/g to 25 m2/g, pore volume from 7.41 cm3/g to 0.09 cm3/g, and adsorption pore size from 18.7 to 7 nm. The skeletal density of sintered granules was similar to the bulk density of amorphous silica (2.2 g/cm3). The hot-pressing accelerated the sintering process, decreasing significantly the pore size and volume.« less

Cobalt-Free Permanent Magnet Alloys.

DTIC Science & Technology

1984-10-01

carbide co- UC CbC lumbium carbide M003 Uranium carbide - tho- UC 2 25ThC rium carbide ZrO2 MgO WOs Use of this Process for MnAlC As indicated in the...cobalt. Free World Cobal Consumption Estimated Breakdown by End Uses Magnetic alloys 20% Cemented carbides - 5% 30 SuPerolloy _ 15% Other steels and...would normally result in the formation of binary alloy of TbFe 2 and preventing the formation of amorphous alloy (Fe-B) contain- ing Tb. The

A two-dimensional, finite-difference model of the oxidation of a uranium carbide fuel pellet

NASA Astrophysics Data System (ADS)

Shepherd, James; Fairweather, Michael; Hanson, Bruce C.; Heggs, Peter J.

2015-12-01

The oxidation of spent uranium carbide fuel, a candidate fuel for Generation IV nuclear reactors, is an important process in its potential reprocessing cycle. However, the oxidation of uranium carbide in air is highly exothermic. A model has therefore been developed to predict the temperature rise, as well as other useful information such as reaction completion times, under different reaction conditions in order to help in deriving safe oxidation conditions. Finite difference-methods are used to model the heat and mass transfer processes occurring during the reaction in two dimensions and are coupled to kinetics found in the literature.

Radioactive Waste Management, its Global Implication on Societies, and Political Impact

NASA Astrophysics Data System (ADS)

Matsui, Kazuaki

2009-05-01

Reprocessing plant in Rokkasho, Japan is under commissioning at the end of 2008, and it starts soon to reprocess about 800 Mt of spent fuel per annum, which have been stored at each nuclear power plant sites in Japan. Fission products together with minor actinides separated from uranium and plutonium in the spent fuel contain almost all radioactivity of it and will be vitrified with glass matrix, which then will fill the canisters. The canisters with the high level radioactive waste (HLW) are so hot in both thermal and radiological meanings that they have to be cooled off for decades before bringing out to any destination. Where is the final destination for HLW in Japan, which is located at the rim of the Pacific Ocean with volcanoes? Although geological formation in Japan is not so static and rather active as the other parts of the planet, experts concluded with some intensive studies and researches that there will be a lot of variety of geological formations even in Japan which can host the HLW for so long times of more than million years. Then an organization to implement HLW disposal program was set up and started to campaign for volunteers to accept the survey on geological suitability for HLW disposal. Some local governments wanted to apply, but were crashed down by local and neighbor governments and residents. The above development is not peculiar only to Japan, but generally speaking more or less common for those with radioactive waste programs. This is why the radioactive waste management is not any more science and technology issue but socio-political one. It does not mean further R&D on geological disposal is not any more necessary, but rather we, each of us, should face much more sincerely the societal and political issues caused by the development of the science and technology. Second topic might be how effective partitioning and transformation technology may be to reduce the burden of waste disposal and denature the waste toxicity? The third one might be the proposal of international nuclear fuel centers which supply nuclear fuel to the nuclear power plants in the region and take back spent fuel which will be reprocessed to recover useful energy resources of uranium and plutonium. This may help non proliferation issue due to world nuclear development beyond renaissance.

Healing acceleration of acetic acid-induced colitis by marigold (Calendula officinalis) in male rats.

PubMed

Tanideh, Nader; Jamshidzadeh, Akram; Sepehrimanesh, Masood; Hosseinzadeh, Masood; Koohi-Hosseinabadi, Omid; Najibi, Asma; Raam, Mozhdeh; Daneshi, Sajad; Asadi-Yousefabad, Seyedeh-Leili

2016-01-01

Ulcerative colitis (UC) is a type of chronic inflammatory bowel disease with unknown etiology. Several therapeutic strategies such as consumption of medicinal plants have been used for its treatment. The aim of this study was to evaluate healing effects of Calendula officinalis hydroalcoholic extract in experimentally induced UC in rat. Ninety-six rats, weighing 200 ± 20 g, were randomly divided into eight equal groups. UC induced by 3% acetic acid and oral doses of C. officinalis extract, 1500 and 3000 mg/kg, and enema (gel 10% and 20%) were given. Two groups as positive controls were given asacol (enema) and oral mesalamine. Negative control groups were given normal saline and base gel. On days 3 and 7, intestinal histopathology and weight changes, plus oxidative stress indices including malondialdehyde (MDA) level and myeloperoxidase (MPO) activity were assayed. A significant increase in the body weight of rats was seen in the group given C. officinalis extract 3000 mg/kg orally, oral mesalamine, and 20% intracolonic gel form of marigold extract compared with negative control and base gel groups during the experimental period. Acute inflammation and granular atrophy after UC induction were resolved completely completely by both 20% intracolonic gel and 3000 mg/kg orally. An increase in MPO activity and a decrease in MDA level in response to oral and intracolonic gel form of C. officinalis were observed 3 and and 7 days after treatment (P < 0.05). Our results indicate that oral and enema forms of hydroalcoholic extract of C. officinalis can be offered as are potential therapeutic agents for UC induced in rats.

HLA-B is the best candidate of susceptibility genes in HLA for Japanese ulcerative colitis.

PubMed

Aizawa, H; Kinouchi, Y; Negoro, K; Nomura, E; Imai, G; Takahashi, S; Takagi, S; Kakuta, Y; Tosa, M; Mochida, A; Matsumura, Y; Endo, K; Shimosegawa, T

2009-06-01

Recently, a genome-wide association study for ulcerative colitis (UC) in the UK population was reported, and several susceptibility loci including the human leukocyte antigen (HLA) region were identified. The strongest association in the HLA region was found at a 400 kb haplotype block containing HLA-DRB1. In Japanese population, previous study suggested the association between UC and HLA-B*52; however, HLA typing was determined using serotyping with the small sample size. The purpose of this study was to perform an association study in HLA-B by genotyping. A total of 320 patients with UC and 322 healthy controls were recruited in this case-control study. All subjects were Japanese. Genotyping of HLA-B was performed by polymerase chain reaction using a sequence-specific primer. When the allele frequencies were compared, significant associations were found with B*52 [odds ratio (OR) = 3.65, P = 1.6 x 10(-17), P(c) = 3.7 x 10(-16)] and B*4002 (OR = 0.52, P = 0.00030, P(c) = 0.0068). The allele frequency of B*52 was significantly higher in patients diagnosed before 40 years of age than in those diagnosed after 40 years (OR = 1.79, P = 0.010, P(c) = 0.020). A combination association map of Japanese UC using our current and previous studies showed two equal peaks of association on HLA-DRB1 and HLA-B, indicating the possible existence of two casual variants in the HLA region inside and outside the 400 kb block found in UK. We conclude that HLA-B contributes to the susceptibility to Japanese UC, especially cases with younger age of onset. The strength of association for HLA-B was equal to that for HLA-DRB1 in Japanese UC, in contrast to the UK population.

Further evaluations of the toxicity of irradiated advanced heavy water reactor fuels.

PubMed

Edwards, Geoffrey W R; Priest, Nicholas D

2014-11-01

The neutron economy and online refueling capability of heavy water moderated reactors enable them to use many different fuel types, such as low enriched uranium, plutonium mixed with uranium, or plutonium and/or U mixed with thorium, in addition to their traditional natural uranium fuel. However, the toxicity and radiological protection methods for fuels other than natural uranium are not well established. A previous paper by the current authors compared the composition and toxicity of irradiated natural uranium to that of three potential advanced heavy water fuels not containing plutonium, and this work uses the same method to compare irradiated natural uranium to three other fuels that do contain plutonium in their initial composition. All three of the new fuels are assumed to incorporate plutonium isotopes characteristic of those that would be recovered from light water reactor fuel via reprocessing. The first fuel investigated is a homogeneous thorium-plutonium fuel designed for a once-through fuel cycle without reprocessing. The second fuel is a heterogeneous thorium-plutonium-U bundle, with graded enrichments of U in different parts of a single fuel assembly. This fuel is assumed to be part of a recycling scenario in which U from previously irradiated fuel is recovered. The third fuel is one in which plutonium and Am are mixed with natural uranium. Each of these fuels, because of the presence of plutonium in the initial composition, is determined to be considerably more radiotoxic than is standard natural uranium. Canadian nuclear safety regulations require that techniques be available for the measurement of 1 mSv of committed effective dose after exposure to irradiated fuel. For natural uranium fuel, the isotope Pu is a significant contributor to the committed effective dose after exposure, and thermal ionization mass spectrometry is sensitive enough that the amount of Pu excreted in urine is sufficient to estimate internal doses, from all isotopes, as low as 1 mSv. In addition, if this method is extended so that Pu is also measured, then the combined amount of Pu and Pu is sufficiently high in the thorium-plutonium fuel that a committed effective dose of 1 mSv would be measurable. However, the fraction of Pu and Pu in the other two fuels is sufficiently low that a 1 mSv dose would remain below the detection limit using this technique. Thus new methods, such as fecal measurements of Pu (or other alpha emitters), will be required to measure exposure to these new fuels.

The used nuclear fuel problem - can reprocessing and consolidated storage be complementary?

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

Phillips, C.; Thomas, I.

2013-07-01

This paper describes our CISF (Consolidated Interim Storage Facilities) and Reprocessing Facility concepts and show how they can be combined with a geologic repository to provide a comprehensive system for dealing with spent fuels in the USA. The performance of the CISF was logistically analyzed under six operational scenarios. A 3-stage plan has been developed to establish the CISF. Stage 1: the construction at the CISF site of only a rail receipt interface and storage pad large enough for the number of casks that will be received. The construction of the CISF Canister Handling Facility, the Storage Cask Fabrication Facility,more » the Cask Maintenance Facility and supporting infrastructure are performed during stage 2. The construction and placement into operation of a water-filled pool repackaging facility is completed for Stage 3. By using this staged approach, the capital cost of the CISF is spread over a number of years. It also allows more time for a final decision on the geologic repository to be made. A recycling facility will be built, this facility will used the NUEX recycling process that is based on the aqueous-based PUREX solvent extraction process, using a solvent of tri-N-butyl phosphate in a kerosene diluent. It is capable of processing spent fuels at a rate of 5 MT per day, at burn-ups up to 50 GWD per ton of spent fuels and a minimum of 5 years out-of-reactor cooling.« less

Spent Fuel Working Group Report. Volume 1

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

O`Toole, T.

1993-11-01

The Department of Energy is storing large amounts of spent nuclear fuel and other reactor irradiated nuclear materials (herein referred to as RINM). In the past, the Department reprocessed RINM to recover plutonium, tritium, and other isotopes. However, the Department has ceased or is phasing out reprocessing operations. As a consequence, Department facilities designed, constructed, and operated to store RINM for relatively short periods of time now store RINM, pending decisions on the disposition of these materials. The extended use of the facilities, combined with their known degradation and that of their stored materials, has led to uncertainties about safety.more » To ensure that extended storage is safe (i.e., that protection exists for workers, the public, and the environment), the conditions of these storage facilities had to be assessed. The compelling need for such an assessment led to the Secretary`s initiative on spent fuel, which is the subject of this report. This report comprises three volumes: Volume I; Summary Results of the Spent Fuel Working Group Evaluation; Volume II, Working Group Assessment Team Reports and Protocol; Volume III; Operating Contractor Site Team Reports. This volume presents the overall results of the Working Group`s Evaluation. The group assessed 66 facilities spread across 11 sites. It identified: (1) facilities that should be considered for priority attention. (2) programmatic issues to be considered in decision making about interim storage plans and (3) specific vulnerabilities for some of these facilities.« less

Fuel cycle for a fusion neutron source

NASA Astrophysics Data System (ADS)

Ananyev, S. S.; Spitsyn, A. V.; Kuteev, B. V.

2015-12-01

The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion-fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium-tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m3Pa/s, and temperature of reactor elements up to 650°C). The deuterium-tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.

JPRS Report, Proliferation Issues

DTIC Science & Technology

1993-06-07

Ruta Skatikaite; RESPUBLIKA, 19 May 93] ................................................................................ 20 Radioactive Beryllium...nuclear fuel will be transported around 2000 to a reprocessing facility in [Yi] IAEA surveillance cameras are said to take four photos Tokai, Ibaraki...Comparing these two methods, the method of extracting May 93 pp 342-346. plutonium is similar to carrying a backpack to transport goods, while

10 CFR Appendix B to Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

Code of Federal Regulations, 2014 CFR

2014-01-01

... related to the design, fabrication, construction, and testing of the structures, systems, and components... components. The pertinent requirements of this appendix apply to all activities affecting the safety-related..., which comprises those quality assurance actions related to the physical characteristics of a material...

FLOOR PLAN OF MAIN PROCESSING BUILDING (CPP601) BASEMENT SHOWING PROCESS ...

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

FLOOR PLAN OF MAIN PROCESSING BUILDING (CPP-601) BASEMENT SHOWING PROCESS CORRIDOR AND EIGHTEEN CELLS. TO LEFT IS LABORATORY BUILDING (CPP-602). INL DRAWING NUMBER 200-0601-00-706-051981. ALTERNATE ID NUMBER CPP-E-1981. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP640) LOOKING NORTHEAST ...

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

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP-640) LOOKING NORTHEAST SHOWING DECK FORMING FOR SOUTH SECTION OF OPERATING CORRIDOR; CONSTRUCTION 44 PERCENT COMPLETE. INL PHOTO NUMBER NRTS-60-3624. Holmes, Photographer, 7/25/1960 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

SECOND FLOOR PLAN OF REMOTE ANALYTICAL FACILITY (CPP627) WARM LABORATORY ...

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

SECOND FLOOR PLAN OF REMOTE ANALYTICAL FACILITY (CPP-627) WARM LABORATORY ROOM, DECONTAMINATION ROOM, HOT CHEMISTRY LABORATORY, AND MULTICURIE CELL ROOM. INL DRAWING NUMBER 200-0627-00-098-105066. ALTERNATE ID NUMBER 4272-14-103. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP640) LOOKING NORTHWEST, ...

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

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP-640) LOOKING NORTHWEST, SHOWING FORMING FOR NORTH WALLS OF CELLS 1, 4 AND 5; CONSTRUCTION 21 PERCENT COMPLETE. INL PHOTO NUMBER NRTS-60-1874. Holmes, Photographer, 4/21/1960 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

SOUTH SECTION OF WEST ELEVATION OF MAIN PROCESSING BUILDING (CPP601) ...

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

SOUTH SECTION OF WEST ELEVATION OF MAIN PROCESSING BUILDING (CPP-601) LOOKING EAST. HEADEND PLANT BUILDING (CPP-640) APPEARS ON LEFT IN PHOTO. INL PHOTO NUMBER HD-22-3-3. Mike Crane, Photographer, 11/1998 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

OBLIQUE PHOTO OF NORTH AND WEST ELEVATIONS OF REMOTE ANALYTICAL ...

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

OBLIQUE PHOTO OF NORTH AND WEST ELEVATIONS OF REMOTE ANALYTICAL FACILITY (CPP-627) LOOKING SOUTHEAST. LABORATORY AND OFFICE BUILDING (CPP-602) APPEAR ON LEFT IN PHOTO. INL PHOTO NUMBER HD-22-2-2. Mike Crane, Photographer, 11/1998 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

76 FR 40943 - Notice of Issuance of Regulatory Guide

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-12

..., Revision 3, ``Criteria for Use of Computers in Safety Systems of Nuclear Power Plants.'' FOR FURTHER..., ``Criteria for Use of Computers in Safety Systems of Nuclear Power Plants,'' was issued with a temporary... Fuel Reprocessing Plants,'' to 10 CFR part 50 with regard to the use of computers in safety systems of...

15 CFR Supplement No. 3 to Part 730 - Other U.S. Government Departments and Agencies With Export Control Responsibilities

Code of Federal Regulations, 2014 CFR

2014-01-01

.... Nuclear Technologies and Services Which Contribute to the Production of Special Nuclear Material (Snm). Technologies Covered Include Nuclear Reactors, Enrichment, Reprocessing, Fuel Fabrication, and Heavy Water...-6050. 10 CFR 205.300 through 205.379 and part 590. Nuclear Materials and Equipment * Nuclear Regulatory...

Radiation chemistry for modern nuclear energy development

NASA Astrophysics Data System (ADS)

Chmielewski, Andrzej G.; Szołucha, Monika M.

2016-07-01

Radiation chemistry plays a significant role in modern nuclear energy development. Pioneering research in nuclear science, for example the development of generation IV nuclear reactors, cannot be pursued without chemical solutions. Present issues related to light water reactors concern radiolysis of water in the primary circuit; long-term storage of spent nuclear fuel; radiation effects on cables and wire insulation, and on ion exchangers used for water purification; as well as the procedures of radioactive waste reprocessing and storage. Radiation effects on materials and enhanced corrosion are crucial in current (II/III/III+) and future (IV) generation reactors, and in waste management, deep geological disposal and spent fuel reprocessing. The new generation of reactors (III+ and IV) impose new challenges for radiation chemists due to their new conditions of operation and the usage of new types of coolant. In the case of the supercritical water-cooled reactor (SCWR), water chemistry control may be the key factor in preventing corrosion of reactor structural materials. This paper mainly focuses on radiation effects on long-term performance and safety in the development of nuclear power plants.

Macrophage migration inhibitory factor gene polymorphisms in inflammatory bowel disease: an association study in New Zealand Caucasians and meta-analysis.

PubMed

Falvey, James D; Bentley, Robert W; Merriman, Tony R; Hampton, Mark B; Barclay, Murray L; Gearry, Richard B; Roberts, Rebecca L

2013-10-21

To investigate the association of macrophage migration inhibitory factor (MIF) promoter polymorphisms with inflammatory bowel disease (IBD) risk. One thousand and six New Zealand Caucasian cases and 540 Caucasian controls were genotyped for the MIF SNP -173G > C (rs755622) and the repeat polymorphism CATT₅₋₈ (rs5844572) using a pre-designed TaqMan SNP assay and capillary electrophoresis, respectively. Data were analysed for single site and haplotype association with IBD risk and phenotype. Meta-analysis was employed, to assess cumulative evidence of association of MIF -173G > C with IBD. All published genotype data for MIF -173G > C in IBD were identified using PubMed and subsequently searching the references of all PubMed-identified studies. Imputed genotypes for MIF -173G > C were generated from the Wellcome Trust Case Control Consortium (and National Institute of Diabetes and Digestive and Kidney Diseases). Separate meta-analyses were performed on Caucasian Crohn's disease (CD) (3863 patients, 6031 controls), Caucasian ulcerative colitis (UC) (1260 patients, 1987 controls), and East Asian UC (416 patients and 789 controls) datasets using the Mantel-Haenszel method. The New Zealand dataset had 93% power, and the meta-analyses had 100% power to detect an effect size of OR = 1.40 at α = 0.05, respectively. In our New Zealand dataset, single-site analysis found no evidence of association of MIF polymorphisms with overall risk of CD, UC, and IBD or disease phenotype (all P values > 0.05). Haplotype analysis found the CATT₅/-173C haplotype occurred at a higher frequency in New Zealand controls compared to IBD patients (0.6 vs 0.01; P = 0.03, OR = 0.22; 95%CI: 0.05-0.99), but this association did not survive bonferroni correction. Meta-analysis of our New Zealand MIF -173G > C data with data from seven additional Caucasian datasets using a random effects model found no association of MIF polymorphisms with CD, UC, or overall IBD. Similarly, meta-analysis of all published MIF -173G > C data from East Asian datasets (416 UC patients, 789 controls) found no association of this promoter polymorphism with UC. We found no evidence of association of MIF promoter polymorphisms with IBD.

A Non-Proliferating Fuel Cycle: No Enrichment, Reprocessing or Accessible Spent Fuel - 12375

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

Parker, Frank L.

2012-07-01

Current fuel cycles offer a number of opportunities for access to plutonium, opportunities to create highly enriched uranium and access highly radioactive wastes to create nuclear weapons and 'dirty' bombs. The non-proliferating fuel cycle however eliminates or reduces such opportunities and access by eliminating the mining, milling and enrichment of uranium. The non-proliferating fuel cycle also reduces the production of plutonium per unit of energy created, eliminates reprocessing and the separation of plutonium from the spent fuel and the creation of a stream of high-level waste. It further simplifies the search for land based deep geologic repositories and interim storagemore » sites for spent fuel in the USA by disposing of the spent fuel in deep sub-seabed sediments after storing the spent fuel at U.S. Navy Nuclear Shipyards that have the space and all of the necessary equipment and security already in place. The non-proliferating fuel cycle also reduces transportation risks by utilizing barges for the collection of spent fuel and transport to the Navy shipyards and specially designed ships to take the spent fuel to designated disposal sites at sea and to dispose of them there in deep sub-seabed sediments. Disposal in the sub-seabed sediments practically eliminates human intrusion. Potential disposal sites include Great Meteor East and Southern Nares Abyssal Plain. Such sites then could easily become international disposal sites since they occur in the open ocean. It also reduces the level of human exposure in case of failure because of the large physical and chemical dilution and the elimination of a major pathway to man-seawater is not potable. Of course, the recovery of uranium from sea water and the disposal of spent fuel in sub-seabed sediments must be proven on an industrial scale. All other technologies are already operating on an industrial scale. If externalities, such as reduced terrorist threats, environmental damage (including embedded emissions), long term care, reduced access to 'dirty' bomb materials, the social and political costs of siting new facilities and the psychological impact of no solution to the nuclear waste problem, were taken into account, the costs would be far lower than those of the present fuel cycle. (authors)« less

Composite uranium carbide targets at TRIUMF: Development and characterization with SEM, XRD, XRF and L-edge densitometry

NASA Astrophysics Data System (ADS)

Kunz, Peter; Bricault, Pierre; Dombsky, Marik; Erdmann, Nicole; Hanemaayer, Vicky; Wong, John; Lützenkirchen, Klaus

2013-09-01

The production of radioactive ion beams (RIB) from spallation targets by irradiation with a continuous 500 MeV proton beam, has been routine at TRIUMF for several years. Based on the experience with composite refractory carbide targets a procedure for the fabrication of UC2/C targets was developed. It includes the preparation of UC2 by carbothermal reduction of UO2, the slip-casting of fine-grained UC2/C slurry on graphite foil under inert gas atmosphere and the cutting of composite target discs which are stacked up to a lamellar structure. The thermal properties of such an arrangement are adequate to withstand the high power deposition of an intense, continuous proton beam and also beneficial for the fast release of short-lived radioactive isotopes. Molecular structure, particle size and the impact of sintering of the target discs were investigated via XRD and SEM. Thickness and mass distribution were measured with position-sensitive LIII-edge densitometry. The results confirm that the properties of the UC2/C target material are well suited for RIB production at TRIUMF while there is still room for improvement with regard to uniformity of mass distribution in target disc thickness.

Systemic metabolic signaling in acute and chronic gastrointestinal inflammation of inflammatory bowel diseases.

PubMed

Karrasch, T; Obermeier, F; Straub, R H

2014-06-01

Acute and chronic intestinal inflammation stimulates innate and adaptive immune systems, thereby increasing energy demand of activated immune cells. Energy regulation by systemically released mediators is of critical importance for homeostasis. We wanted to find out how systemic metabolic mediators are affected during intestinal inflammation. A total of 123 patients suffering from Crohn's disease (CD), 76 patients with ulcerative colitis (UC), and 21 healthy controls were recruited. Patients receiving systemic steroids or therapy regimens including biologicals (anti-TNF) were excluded from the study. Serum levels of IL-6, CRP, insulin, glucose, free fatty acid, and RBP-4 were measured by ELISA and RIA. Intestinal inflammation was accompanied by elevated systemic inflammatory para-meters such as IL-6 and CRP in UC and CD and, concomitantly, with elevated insulin levels and increased insulin/glucose ratio in patients with UC. This indicates insulin resistance in liver, muscle, and fat. In addition, intestinal inflammation was associated with elevated levels of circulating free fatty acids in UC and CD, indicating an activation of the organism's appeal for energy-rich substrates (energy appeal reaction). RBP-4 serum levels were also high in acute and chronic intestinal inflammation in UC and CD, which can support insulin resistance. The organism's "energy appeal reaction" in response to acute and chronic inflammation provides free energy in the circulation, which is needed by inflammatory cells. A major mechanism of the redirection program is insulin resistance. New therapeutic strategies might be developed in the future, directly impacting on the storage and utilization of energy-rich fuels. © Georg Thieme Verlag KG Stuttgart · New York.

Light atom quantum oscillations in UC and US

DOE PAGES

Yiu, Yuen; Aczel, Adam A.; Granroth, Garrett E.; ...

2016-01-19

High energy vibrational scattering in the binary systems UC and US is measured using time-of-flight inelastic neutron scattering. A clear set of well-defined peaks equally separated in energy is observed in UC, corresponding to harmonic oscillations of the light C atoms in a cage of heavy U atoms. The scattering is much weaker in US and only a few oscillator peaks are visible. We show how the difference between the materials can be understood by considering the neutron scattering lengths and masses of the lighter atoms. Monte Carlo ray tracing is used to simulate the scattering, with near quantitative agreementmore » with the data in UC, and some differences with US. The possibility of observing anharmonicity and anisotropy in the potentials of the light atoms is investigated in UC. Lastly, the observed data is well accounted for by considering each light atom as a single atom isotropic quantum harmonic oscillator.« less

Mapping of a disease susceptibility locus in chromosome 6p in Japanese patients with ulcerative colitis.

PubMed

Nomura, E; Kinouchi, Y; Negoro, K; Kojima, Y; Oomori, S; Sugimura, M; Hiroki, M; Takagi, S; Aihara, H; Takahashi, S; Hiwatashi, N; Shimosegawa, T

2004-09-01

Ulcerative colitis (UC) is a multifactorial disorder with both genetic and environmental factors. HLA-B*52 and DRB1*1502 are reported to be strongly associated with UC in Japan. However, the actual susceptible gene has not been identified yet. In this study, to map precisely the susceptible locus for UC, we performed association mapping in the chromosome 6p using 24 microsatellite markers distributed over 16 Mb. A total of 183 patients with UC and 186 healthy controls (HC) were included in this study. In all, 15 markers around the human leukocyte antigen (HLA) region showed statistical significance in the genotypic differentiation test concerned with the allelic distribution between the UC and HC. Especially, the markers between the centromeric region of HLA class I and the telomeric region of class III showed remarkably low P-values and the allele239 of C2-4-4 in class I marker showed the strongest association (Pc=2.9 x 10(-9): OR=3.74, 95% CI=2.50-5.60). Furthermore, we found strong linkage disequilibrium (LD) between the allele239 of C2-4-4 and HLA-B*52 in haplotype analysis. These results provide evidence that, in Japanese, important determinants of disease susceptibility to UC may exist in HLA, especially between the centromeric region of class I and the telomeric region of class III, under the strong LD with HLA-B*52.

Development of Online Spectroscopic pH Monitoring for Nuclear Fuel Reprocessing Plants: Weak Acid Schemes.

PubMed

Casella, Amanda J; Ahlers, Laura R H; Campbell, Emily L; Levitskaia, Tatiana G; Peterson, James M; Smith, Frances N; Bryan, Samuel A

2015-05-19

In nuclear fuel reprocessing, separating trivalent minor actinides and lanthanide fission products is extremely challenging and often necessitates tight pH control in TALSPEAK (Trivalent Actinide-Lanthanide Separation by Phosphorus reagent Extraction from Aqueous Komplexes) separations. In TALSPEAK and similar advanced processes, aqueous pH is one of the most important factors governing the partitioning of lanthanides and actinides between an aqueous phase containing a polyaminopolycarboxylate complexing agent and a weak carboxylic acid buffer and an organic phase containing an acidic organophosphorus extractant. Real-time pH monitoring would significantly increase confidence in the separation performance. Our research is focused on developing a general method for online determination of the pH of aqueous solutions through chemometric analysis of Raman spectra. Spectroscopic process-monitoring capabilities, incorporated in a counter-current centrifugal contactor bank, provide a pathway for online, real-time measurement of solution pH. The spectroscopic techniques are process-friendly and can be easily configured for online applications, whereas classic potentiometric pH measurements require frequent calibration/maintenance and have poor long-term stability in aggressive chemical and radiation environments. Raman spectroscopy discriminates between the protonated and deprotonated forms of the carboxylic acid buffer, and the chemometric processing of the Raman spectral data with PLS (partial least-squares) regression provides a means to quantify their respective abundances and therefore determine the solution pH. Interpretive quantitative models have been developed and validated under a range of chemical composition and pH conditions using a lactic acid/lactate buffer system. The developed model was applied to new spectra obtained from online spectral measurements during a solvent extraction experiment using a counter-current centrifugal contactor bank. The model predicted the pH of this validation data set within 11% for pH > 2, thus demonstrating that this technique could provide the capability of monitoring pH online in applications such as nuclear fuel reprocessing.

Argon Shrouded Plasma Spraying of Tantalum over Titanium for Corrosion Protection in Fluorinated Nitric Acid Media

NASA Astrophysics Data System (ADS)

Vetrivendan, E.; Jayaraj, J.; Ningshen, S.; Mallika, C.; Kamachi Mudali, U.

2018-02-01

Argon shrouded plasma spraying (ASPS) was used to deposit a Ta coating on commercially pure Ti (CP-Ti) under inert argon, for dissolver vessel application in the aqueous spent fuels reprocessing plant with high plutonium content. Oxidation during plasma spraying was minimized by shrouding argon system. Porosity and oxide content were controlled by optimizing the spraying parameters, to obtain a uniform and dense Ta coating. The Ta particle temperature and velocity were optimized by judiciously controlling the spray parameters, using a spray diagnostic charge-coupled device camera. The corrosion resistance of the Ta coatings developed by ASPS was investigated by electrochemical studies in 11.5 M HNO3 and 11.5 M HNO3 + 0.05 M NaF. Similarly, the durability of the ASPS Ta coating/substrate was evaluated as per ASTM A262 Practice-C test in boiling nitric acid and fluorinated nitric acid for 240 h. The ASPS Ta coating exhibited higher corrosion resistance than the CP-Ti substrate, as evident from electrochemical studies, and low corrosion rate with excellent coating stability in boiling nitric, and fluorinated nitric acid. The results of the present study revealed that tantalum coating by ASPS is a promising strategy for improving the corrosion resistance in the highly corrosive reprocessing environment.

Industrial research for transmutation scenarios

NASA Astrophysics Data System (ADS)

Camarcat, Noel; Garzenne, Claude; Le Mer, Joël; Leroyer, Hadrien; Desroches, Estelle; Delbecq, Jean-Michel

2011-04-01

This article presents the results of research scenarios for americium transmutation in a 22nd century French nuclear fleet, using sodium fast breeder reactors. We benchmark the americium transmutation benefits and drawbacks with a reference case consisting of a hypothetical 60 GWe fleet of pure plutonium breeders. The fluxes in the various parts of the cycle (reactors, fabrication plants, reprocessing plants and underground disposals) are calculated using EDF's suite of codes, comparable in capabilities to those of other research facilities. We study underground thermal heat load reduction due to americium partitioning and repository area minimization. We endeavor to estimate the increased technical complexity of surface facilities to handle the americium fluxes in special fuel fabrication plants, americium fast burners, special reprocessing shops, handling equipments and transport casks between those facilities.

Yttrium and rare earth stabilized fast reactor metal fuel

DOEpatents

Guon, Jerold; Grantham, LeRoy F.; Specht, Eugene R.

1992-01-01

To increase the operating temperature of a reactor, the melting point and mechanical properties of the fuel must be increased. For an actinide-rich fuel, yttrium, lanthanum and/or rare earth elements can be added, as stabilizers, to uranium and plutonium and/or a mixture of other actinides to raise the melting point of the fuel and improve its mechanical properties. Since only about 1% of the actinide fuel may be yttrium, lanthanum, or a rare earth element, the neutron penalty is low, the reactor core size can be reduced, the fuel can be burned efficiently, reprocessing requirements are reduced, and the nuclear waste disposal volumes reduced. A further advantage occurs when yttrium, lanthanum, and/or other rare earth elements are exposed to radiation in a reactor, they produce only short half life radioisotopes, which reduce nuclear waste disposal problems through much shorter assured-isolation requirements.

DLA Information Systems Technology Integration Guide for the Fiscal Year of 1991

DTIC Science & Technology

1991-01-01

systems which currently exist. In addition, the institution of full and open - opetition for acquisitions for our information systems has driven us...00 U,~c U U C )C/) III 0. -- -------- U)) 0 i U) U)C )z~ w nU U )c < ID a oE co 0 )C c ) 0 cu cuwwwwww Pag E0 I-- I - 0 < 0~0 I- a)) -Jc oo Page...0 -- co Ea -- LL . 0 E c 0) E- I-- w > G, 0 4A 0 .9 Page 1 0u 0 Q. x cts C0 m 0o > n0 ) 0-6 V (t) co .0=)E: Go c ’V x 0- 0 V) E Cc~ E C) ~ < o a

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

Shepherd, James; Fairweather, Michael; Hanson, Bruce C.

The oxidation of spent uranium carbide fuel, a candidate fuel for Generation IV nuclear reactors, is an important process in its potential reprocessing cycle. However, the oxidation of uranium carbide in air is highly exothermic. A model has therefore been developed to predict the temperature rise, as well as other useful information such as reaction completion times, under different reaction conditions in order to help in deriving safe oxidation conditions. Finite difference-methods are used to model the heat and mass transfer processes occurring during the reaction in two dimensions and are coupled to kinetics found in the literature.

Photon Upconversion and Molecular Solar Energy Storage by Maximizing the Potential of Molecular Self-Assembly.

PubMed

Kimizuka, Nobuo; Yanai, Nobuhiro; Morikawa, Masa-Aki

2016-11-29

The self-assembly of functional molecules into ordered molecular assemblies and the fulfillment of potentials unique to their nanotomesoscopic structures have been one of the central challenges in chemistry. This Feature Article provides an overview of recent progress in the field of molecular self-assembly with the focus on the triplet-triplet annihilation-based photon upconversion (TTA-UC) and supramolecular storage of photon energy. On the basis of the integration of molecular self-assembly and photon energy harvesting, triplet energy migration-based TTA-UC has been achieved in varied molecular systems. Interestingly, some molecular self-assemblies dispersed in solution or organogels revealed oxygen barrier properties, which allowed TTA-UC even under aerated conditions. The elements of molecular self-assembly were also introduced to the field of molecular solar thermal fuel, where reversible photoliquefaction of ionic crystals to ionic liquids was found to double the molecular storage capacity with the simultaneous pursuit of switching ionic conductivity. A future prospect in terms of innovating molecular self-assembly toward molecular systems chemistry is also discussed.

Radionuclide speciation in effluent from La Hague reprocessing plant in France.

PubMed

Salbu, B; Skipperud, L; Germain, P; Guéguéniat, P; Strand, P; Lind, O C; Christensen, G

2003-09-01

Effluent from the La Hague nuclear fuel reprocessing plant was mixed with seawater in order to investigate the fate of the various radionuclides. Thus, a major objective of the present work is to characterize the effluent from La Hague reprocessing plant and to study how the radionuclide speciation changes with time when discharged into the marine environment. Discharges from the La Hague nuclear reprocessing plant represent an important source of artificially produced radionuclides to the North Sea. The transport, distribution, and biological uptake of radionuclides in the marine environment depends, however, on the physicochemical forms of radionuclides in the discharged effluents and on transformation processes that occur after entering the coastal waters. Information of these processes is needed to understand the transport and long-term distribution of the radionuclides. In the present work, a weekly discharged effluent from the nuclear fuel reprocessing plant at Cap La Hague in France was mixed with coastal water and fractionated with respect to particle size and charged species using ultra centrifugation and hollow fiber ultrafiltration with on line ion exchange. The size distribution pattern of gamma-emitting radionuclides was followed during a 62-h period after mixing the effluent with seawater. 54Mn was present as particulate material in the effluent, while other investigated radionuclides were discharged in a more mobile form or were mobilized after mixing with sea water (e.g., 60Co) and can be transported long distances in the sea. Sediments can act as a sink for less mobile discharged radionuclides (Skipperud et al. 2000). A kinetic model experiment was performed to provide information of the time-dependent distribution coefficients, Kd (t). The retention of the effluent radionuclides in sediments was surprisingly low (Kd 20-50), and the sediments acted as a poor sink for the released radionuclides. Due to the presence of non-reacting radionuclide species in the effluent, a major fraction of the radionuclides, such as Cs-isotopes, 106Ru and 125Sb, in the effluent will be subjected to marine transport to the Northern Seas (i.e., the North Sea, Norwegian Sea and the Barents Sea). The La Hague effluent may, therefore, contribute to enriched levels of radionuclides found in the English Channel, including 90Sr, 60Co and Pu-isotopes, and also 106Ru and 125Sb.

Significance of and prospects for fuel recycle in Japan

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

Otsuka, K.; Ikeda, K.

Japan's nuclear power plant capacity ranks fourth in the world at around 20 GW. But nuclear fuel cycle industries (enrichment, reprocessing and radioactive waste management) are still in their infancy compared with the size and stage of the power plants. Thus it is a matter of urgency to establish a nuclear fuel cycle in Japan which can promote nuclear energy as a quasi-indigenous energy source. Some moves toward establishing a nuclear fuel cycle have been observed recently. As a case in point, in July 1984, the Federation of Electric Power Companies has formally requested Aomori Prefecture to locate nuclear fuelmore » cycle facilities in the Shimokita Peninsula region. Plutonium recovered from spent fuel will be utilized in LWR, ATR, and FBR. Research and development activities on these technologies are in progress.« less

Decreased levels of serum omentin-1 in patients with inflammatory bowel disease.

PubMed

Yin, Jian; Hou, Peng; Wu, Zhiqiang; Nie, Yanxiao

2015-01-10

Inflammation is involved in the mechanism of inflammatory bowel disease (IBD). Omentin, a newly discovered adipokine, is thought to play an anti-inflammatory role. This study aimed to determine whether serum levels of omentin-1 are associated with the presence and disease activity of IBD. This study consisted of 192 patients with IBD: 100 with Crohn's disease [CD], 92 with ulcerative colitis [UC], and 104 healthy subjects. Serum levels of omentin-1 were measured using enzyme-linked immunosorbent assay (ELISA). Serum omentin-1 levels were significantly decreased in CD and UC patients compared with healthy controls. Multivariable logistic regression analysis revealed that serum omentin-1 levels were inversely associated with the presence of CD and UC. Active CD and UC patients both had significantly decreased levels of serum omentin-1 compared with inactive CD and UC patients. In both CD and UC patients, serum omentin-1 levels were significantly associated with decreased levels of body mass index (BMI) and C-reactive protein (CRP). Decreased serum omentin-1 levels could be considered as an independent predicting marker of the presence and disease activity of IBD.

Healing Acceleration of Acetic Acid-induced Colitis by Marigold (Calendula officinalis) in Male Rats

PubMed Central

Tanideh, Nader; Jamshidzadeh, Akram; Sepehrimanesh, Masood; Hosseinzadeh, Masood; Koohi-Hosseinabadi, Omid; Najibi, Asma; Raam, Mozhdeh; Daneshi, Sajad; Asadi-Yousefabad, Seyedeh-Leili

2016-01-01

Background/Aim: Ulcerative colitis (UC) is a type of chronic inflammatory bowel disease with unknown etiology. Several therapeutic strategies such as consumption of medicinal plants have been used for its treatment. The aim of this study was to evaluate healing effects of Calendula officinalis hydroalcoholic extract in experimentally induced UC in rat. Materials and Methods: Ninety-six rats, weighing 200 ± 20 g, were randomly divided into eight equal groups. UC induced by 3% acetic acid and oral doses of C. officinalis extract, 1500 and 3000 mg/kg, and enema (gel 10% and 20%) were given. Two groups as positive controls were given asacol (enema) and oral mesalamine. Negative control groups were given normal saline and base gel. On days 3 and 7, intestinal histopathology and weight changes, plus oxidative stress indices including malondialdehyde (MDA) level and myeloperoxidase (MPO) activity were assayed. Results: A significant increase in the body weight of rats was seen in the group given C. officinalis extract 3000 mg/kg orally, oral mesalamine, and 20% intracolonic gel form of marigold extract compared with negative control and base gel groups during the experimental period. Acute inflammation and granular atrophy after UC induction were resolved completely completely by both 20% intracolonic gel and 3000 mg/kg orally. An increase in MPO activity and a decrease in MDA level in response to oral and intracolonic gel form of C. officinalis were observed 3 and and 7 days after treatment (P < 0.05). Conclusion: Our results indicate that oral and enema forms of hydroalcoholic extract of C. officinalis can be offered as are potential therapeutic agents for UC induced in rats. PMID:26831607

Meta-analysis of the association between appendiceal orifice inflammation and appendectomy and ulcerative colitis.

PubMed

Deng, Peng; Wu, Junchao

2016-07-01

This study aimed to investigate the relationship between appendiceal orifice inflammation (AOI) and appendectomy and ulcerative colitis (UC) by a meta-analysis. Databases were thoroughly searched for studies on AOI and UC up to January 2016. Three comparisons were performed: a) whether the previous appendectomy was a risk factor of UC; b) influence of appendectomy on UC courses; c) influence of AOI on UC severity. Odds ratios (ORs) and 95% confidence intervals (CIs) were the effects sizes. The merging of results and publication bias assessment were performed by using RevMan 5.3. Sensitivity analysis was conducted using Stata 12.0. Nineteen studies were selected in the present study. Results of comparison I showed that appendectomy was a protective factor of UC (OR = 0.44; 95% CI [0.30, 0.64]). Comparison II indicated appendectomy had no significant influence in the courses of UC (proctitis: OR = 1.03, 95% CI [0.74, 1.42]; left-sided colitis: OR = 1.01, 95% CI [0.73, 1.39]; pancolitis: OR = 0.92, 95% CI [0.59, 1.43]; colectomy: OR = 1.38, 95% CI [0.62, 3.04]). Comparison III indicated UC combined with AOI did not affect the courses of UC (proctitis: OR = 1.15, 95% CI [0.67, 1.98]; left-sided colitis: OR = 1.14, 95% CI [0.24, 5.42]; colectomy: OR = 0.36, 95% CI [0.10, 1.23]). Sensitivity analysis confirmed the robust of the results in the present study. In conclusion, this meta-analysis indicated appendectomy can reduce the risk of UC. But appendectomy or AOI had no influence on the severity of the disease and the effect of surgical treatment.

A proliferation of nuclear waste for the Southeast.

PubMed

Alvarez, Robert; Smith, Stephen

2007-12-01

The U.S. Department of Energy's (DOE) Global Nuclear Energy Partnership (GNEP) is being promoted as a program to bring about the expansion of worldwide nuclear energy. Here in the U.S. much of this proposed nuclear power expansion is slated to happen in the Southeast, including here in South Carolina. Under the GNEP plan, the United States and its nuclear partners would sell nuclear power plants to developing nations that agree not to pursue technologies that would aid nuclear weapons production, notably reprocessing and uranium enrichment. As part of the deal, the United States would take highly radioactive spent ("used") fuel rods to a reprocessing center in this country. Upon analysis of the proposal, it is clear that DOE lacks a credible plan for the safe management and disposal of radioactive wastes stemming from the GNEP program and that the high costs and possible public health and environmental impacts from the program pose significant risks, especially to this region. Given past failures to address waste problems before they were created, DOE's rush to invest major public funds for deployment of reprocessing should be suspended.

Canister arrangement for storing radioactive waste

DOEpatents

Lorenzo, D.K.; Van Cleve, J.E. Jr.

1980-04-23

The subject invention relates to a canister arrangement for jointly storing high level radioactive chemical waste and metallic waste resulting from the reprocessing of nuclear reactor fuel elements. A cylindrical steel canister is provided with an elongated centrally disposed billet of the metallic waste and the chemical waste in vitreous form is disposed in the annulus surrounding the billet.

Canister arrangement for storing radioactive waste

DOEpatents

Lorenzo, Donald K.; Van Cleve, Jr., John E.

1982-01-01

The subject invention relates to a canister arrangement for jointly storing high level radioactive chemical waste and metallic waste resulting from the reprocessing of nuclear reactor fuel elements. A cylindrical steel canister is provided with an elongated centrally disposed billet of the metallic waste and the chemical waste in vitreous form is disposed in the annulus surrounding the billet.

A change in temperature modulates defence to yellow (stripe) rust in wheat line UC1041 independently of resistance gene Yr36.

PubMed

Bryant, Ruth R M; McGrann, Graham R D; Mitchell, Alice R; Schoonbeek, Henk-Jan; Boyd, Lesley A; Uauy, Cristobal; Dorling, Steve; Ridout, Christopher J

2014-01-08

Rust diseases are of major importance in wheat production worldwide. With the constant evolution of new rust strains and their adaptation to higher temperatures, consistent and durable disease resistance is a key challenge. Environmental conditions affect resistance gene performance, but the basis for this is poorly understood. Here we show that a change in day temperature affects wheat resistance to Puccinia striiformis f. sp tritici (Pst), the causal agent of yellow (or stripe) rust. Using adult plants of near-isogenic lines UC1041 +/- Yr36, there was no significant difference between Pst percentage uredia coverage in plants grown at day temperatures of 18°C or 25°C in adult UC1041 + Yr36 plants. However, when plants were transferred to the lower day temperature at the time of Pst inoculation, infection increased up to two fold. Interestingly, this response was independent of Yr36, which has previously been reported as a temperature-responsive resistance gene as Pst development in adult UC1041 -Yr36 plants was similarly affected by the plants experiencing a temperature reduction. In addition, UC1041 -Yr36 plants grown at the lower temperature then transferred to the higher temperature were effectively resistant and a temperature change in either direction was shown to affect Pst development up to 8 days prior to inoculation. Results for seedlings were similar, but more variable compared to adult plants. Enhanced resistance to Pst was observed in seedlings of UC1041 and the cultivar Shamrock when transferred to the higher temperature. Resistance was not affected in seedlings of cultivar Solstice by a temperature change in either direction. Yr36 is effective at 18°C, refining the lower range of temperature at which resistance against Pst is conferred compared to previous studies. Results reveal previously uncharacterised defence temperature sensitivity in the UC1041 background which is caused by a change in temperature and independently of Yr36. This novel phenotype is present in some cultivars but absent in others, suggesting that Pst defence may be more stable in some cultivars than others when plants are exposed to varying temperatures.

Undernutrition and serum and urinary urea nitrogen of white-tailed deer during winter

USGS Publications Warehouse

DelGiudice, G.D.; Mech, L.D.; Seal, U.S.

1994-01-01

Direct, practical means of assessing undernutrition in deer (Odocoileus spp.) and other ungulates during winter are needed in areas of research and management. We examined the relationship between mass loss and serum urea nitrogen (SUN) and urinary urea nitrogen:creatinine (U:C) in captive white-tailed deer (O. virginianus). During 4 February-5 May 1988, we maintained 7 adult white-tailed deer on various feeding regimes to simulate natural nutritional restriction during winter. Mass loss was greater (P = 0.037) in deer (17.0-32.2%) fed restricted amounts of a low protein low energy diet versus control deer (7.0-17.4%) fed the same diet ad libitum. Serum triiodothyronine (T3) concentrations did not differ (P = 0.191) between groups, but declined (P = 0.001) as nutrition declined. Slopes of percent mass lossSUN and urinary U:C relationships were positive (P = 0.008 and 0.055) in 7 and 6 deer, respectively. Mean U:C was directly related (r2 = 0.52, P = 0.040) to mean cumulative mass loss, whereas mean SUN was not (r2 = 0.29, P = 0.125). Data presented support the potential of urinary U:C as an index of winter nutritional condition of white-tailed deer; however, additional research is required to provide a complete understanding of this index's utility under field conditions.

Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report December 2014

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

Renae Soelberg

2014-12-01

• PNNL has completed sectioning of the U.C. Berkeley hydride fuel rodlet 1 (highest burn-up) and is currently polishing samples in preparation for optical metallography. • A disk was successfully sectioned from rodlet 1 at the location of the internal thermocouple tip as desired. The transition from annular pellet to solid pellet is verified by the eutectic-filled inner cavity located on the back face of this disk (top left) and the solid front face (bottom left). Preliminary low-resolution images indicate interesting sample characteristics in the eutectic surrounding the rodlet at the location of the outer thermocouple tip (right). This samplemore » has been potted and is currently being polished for high-resolution optical microscopy and subsequent SEM analysis. (See images.)« less

Interface induced high temperature superconductivity in single unit-cell FeSe on SrTiO{sub 3}(110)

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

Zhou, Guanyu; Zhang, Ding; Liu, Chong

2016-05-16

We report high temperature superconductivity in one unit-cell (1-UC) FeSe films grown on SrTiO{sub 3} (STO)(110) substrate by molecular beam epitaxy. By in-situ scanning tunneling microscopy measurement, we observe a superconducting gap as large as 17 meV on the 1-UC FeSe films. Transport measurements on 1-UC FeSe/STO(110) capped with FeTe layers reveal superconductivity with an onset transition temperature (T{sub C}) of 31.6 K and an upper critical magnetic field of 30.2 T. We also find that T{sub C} can be further increased by external electric field although the effect is weaker than that on STO(001) substrate.

Combination of Azathioprine and Aminosalicylate Treatment Prevent Risk of Cardiovascular Disease in Women with Ulcerative Colitis by Reducing Inflammation.

PubMed

dos Santos, Lana Claudinez; Costa, Aline Villela; Lopes, Lorrayne Gonçalves; Leonel, Alda Jusceline; Aguilar, Edenil Costa; Noviello, Maria de Lourdes Meirelles; Ferrari, Maria de Lourdes de Abreu; Alvarez-Leite, Jacqueline I

2015-08-07

Ulcerative colitis (UC) is a chronic inflammatory bowel disease with involvement of the immune system. Chronic inflammatory diseases have been associated with increased risk of cardiovascular disease (CVD) but few studies have assessed this risk in patients with UC and the influence of drug treatment. Thus, we evaluated the risk of development of CVD in women with UC in clinical remission, considering the drug treatment. Twenty-one women with UC participated in this study: 12 used aminosalicylates (ASA group) and 9 used azathioprine added to aminosalicylates (AZA+ASA group). The healthy control group was matched for age. We evaluated blood pressure, body composition, and biochemical and immunological parameters. Compared to the respective control group, the UC groups showed expansion of body fat and less lean body mass. Blood pressure, pro-inflammatory cytokines, nitric oxide, C reactive protein, erythrocyte sedimentation rate (ESR), and anti-oxidized LDL antibodies were higher in UC groups. Only AZA+ASA group showed increased anti-inflammatory cytokines (IL-10 and TGF-β). Framingham scores showed higher risk of CVD in UC groups. UC groups were compared and women treated with azathioprine showed reduction of total protein, globulin, ESR, and lymphocytes, with increased IL-6, TNF, IL-10, and TGF-β. Our data suggest that women with UC in clinical remission have a higher risk for development of atherosclerosis and CVD when compared to the control group, while women treated with azathioprine seem more protected than those treated only with aminosalicylates, due to better regulation of the inflammatory process.

Department of the Air Force Supporting Data for Fiscal Year 1995 Research, Development, Test and Evaluation Descriptive Summaries

DTIC Science & Technology

1994-02-01

c CLIOU 0 V ONj 00 . ca % 4 .) en 4...AD-A279 096 oci 00 o Io lef 1-4 00 .00 >~-4 C ) .tU E4.~ -4; ’.) .0 4-- -. Eu ca--- CU. >0 LU .e >. 060 U.U> m W CA *0E c00 &MU)%~ z ~0 0 U co o w) 0...0O o E~ ’= C E . M u-L od \\cu~ CU-L 4w) uC~ >0 oo CQ O 0 t ~ C /) ~ C MI 0 U *-’-’ cu. Cu~ 0%oU .E O 4uC). 0 0cl .20 0ý 0uU (U ~ U) toU 00 -Z 0 a 0

Assessing the impacts of ethanol and isobutanol on gaseous and particulate emissions from flexible fuel vehicles.

PubMed

Karavalakis, Georgios; Short, Daniel; Russell, Robert L; Jung, Heejung; Johnson, Kent C; Asa-Awuku, Akua; Durbin, Thomas D

2014-12-02

This study investigated the effects of higher ethanol blends and an isobutanol blend on the criteria emissions, fuel economy, gaseous toxic pollutants, and particulate emissions from two flexible-fuel vehicles equipped with spark ignition engines, with one wall-guided direct injection and one port fuel injection configuration. Both vehicles were tested over triplicate Federal Test Procedure (FTP) and Unified Cycles (UC) using a chassis dynamometer. Emissions of nonmethane hydrocarbons (NMHC) and carbon monoxide (CO) showed some statistically significant reductions with higher alcohol fuels, while total hydrocarbons (THC) and nitrogen oxides (NOx) did not show strong fuel effects. Acetaldehyde emissions exhibited sharp increases with higher ethanol blends for both vehicles, whereas butyraldehyde emissions showed higher emissions for the butanol blend relative to the ethanol blends at a statistically significant level. Particulate matter (PM) mass, number, and soot mass emissions showed strong reductions with increasing alcohol content in gasoline. Particulate emissions were found to be clearly influenced by certain fuel parameters including oxygen content, hydrogen content, and aromatics content.

Demand driven salt clean-up in a molten salt fast reactor - Defining a priority list.

PubMed

Merk, B; Litskevich, D; Gregg, R; Mount, A R

2018-01-01

The PUREX technology based on aqueous processes is currently the leading reprocessing technology in nuclear energy systems. It seems to be the most developed and established process for light water reactor fuel and the use of solid fuel. However, demand driven development of the nuclear system opens the way to liquid fuelled reactors, and disruptive technology development through the application of an integrated fuel cycle with a direct link to reactor operation. The possibilities of this new concept for innovative reprocessing technology development are analysed, the boundary conditions are discussed, and the economic as well as the neutron physical optimization parameters of the process are elucidated. Reactor physical knowledge of the influence of different elements on the neutron economy of the reactor is required. Using an innovative study approach, an element priority list for the salt clean-up is developed, which indicates that separation of Neodymium and Caesium is desirable, as they contribute almost 50% to the loss of criticality. Separating Zirconium and Samarium in addition from the fuel salt would remove nearly 80% of the loss of criticality due to fission products. The theoretical study is followed by a qualitative discussion of the different, demand driven optimization strategies which could satisfy the conflicting interests of sustainable reactor operation, efficient chemical processing for the salt clean-up, and the related economic as well as chemical engineering consequences. A new, innovative approach of balancing the throughput through salt processing based on a low number of separation process steps is developed. Next steps for the development of an economically viable salt clean-up process are identified.

20 CFR 603.3 - What is the purpose and scope of this subpart?

Code of Federal Regulations, 2010 CFR

2010-04-01

... FEDERAL-STATE UNEMPLOYMENT COMPENSATION (UC) PROGRAM; CONFIDENTIALITY AND DISCLOSURE OF STATE UC INFORMATION Confidentiality and Disclosure Requirements § 603.3 What is the purpose and scope of this subpart... the disclosure requirements of Sections 303(a)(7), (c)(1), (d), (e), (h), and (i), SSA, and Section...

The {sup 129}Iodine bomb pulse recorded in Mississippi River Delta sediments: Results from isotopes of I, Pu, Cs, Pb, and C

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

Oktay, S.D.; Santschi, P.H.; Moran, J.E.

2000-03-01

Anthropogenic sources from nuclear reprocessing discharges and bomb test fallout have completely overwhelmed the natural signal on the surface of the earth in the last 50 years. However, the transfer functions in and out of environmental compartments are not well known due to temporal variations in the sources of {sup 129}I and to a lack of knowledge regarding the forms of iodine. From a vertical profile of {sup 129}I/{sup 127}I ratios in sediments located in the Mississippi Delta region in approximately 60 meters water depth, the {sup 129}I input function to this region was reconstructed. Dates in the core weremore » assigned based on the plutonium peak at 20 cm depth (assumed to have been deposited in 1963) and the excess {sup 210}Pb profile in the same depth interval, and below that, based on the steadily decreasing {sup 240}Pu/{sup 239}Pu ratios from a ratio of 0.18 at 22 cm to 0.05 at 57 cm depth, the 1953 horizon. Atom ratios of {sup 129}I/{sup 137}I Cs, decay corrected to 1962, the year of maximum radionuclide production, are about 0.3, very close to the production ratios of about 0.2 during atomic bomb tests. This evidence, combined with other observations, strongly suggests that {sup 129}I in Mississippi River Delta sediments originates from atomic bomb fallout eroded from soils of the Mississippi River drainage basin, with little alteration of the isotopic ratios during transport from watershed to coastal deposits. Based on these observations and on laboratory evidence, the authors propose a conceptual model which explains this correspondence and the low {sup 129}I/{sup 127}I ratios. Differences in mobilities of the different chemical forms of {sup 129}I and {sup 127}I, as well as the variances in chemical forms of {sup 129}I from nuclear bomb fallout versus nuclear fuel reprocessing, are proposed to have created such a correspondence between I-isotope ratios and bomb fallout nuclides, without revealing recent inputs from nuclear fuel reprocessing releases to the northern hemisphere observed in watersheds of the USA and Europe.« less

An all sulfur analogue of the smallest subunit of F420-non-reducing hydrogenase from Methanococcus voltae--metal binding and structure.

PubMed

Pfeiffer, M; Klein, A; Steinert, P; Schomburg, D

The 25 amino acid long subunit VhuU of the F420-non-reducing hydrogenase from Methanococcus voltae contains selenocysteine within the consensus sequence of known [NiFe] hydrogenases DP(C or U)CxxCxxH (U = selenocysteine). The sulfur-analogue VhuUc was chemically synthesized, purified and its metal binding capability, the catalytic properties, and structural features were investigated. The polypeptide was able to bind nickel, but did not catalyse the heterolytic activation of H2. 2D-NMR spectroscopy revealed an alpha-helical secondary structure for the 15 N-terminal amino acids in 50% TFE. Nickel only binds to the C-terminus, which contains the conserved amino acid motif. Structures derived from the NMR data are compatible with the participation of both sulfur atoms from the conserved cysteine residues in a metal ion binding. Structures obtained from the data sets for Ni.VhuUc as well as Zn.VhuUc showed no further ligands. The informational value for Ni.VhuUc was low due to paramagnetism.

Thermal reactions of uranium metal, UO 2, U 3O 8, UF 4, and UO 2F 2 with NF 3 to produce UF 6

NASA Astrophysics Data System (ADS)

McNamara, Bruce; Scheele, Randall; Kozelisky, Anne; Edwards, Matthew

2009-11-01

This paper demonstrates that NF 3 fluorinates uranium metal, UO 2, UF 4, UO 3, U 3O 8, and UO 2F 2·2H 2O to produce the volatile UF 6 at temperatures between 100 and 550 °C. Thermogravimetric and differential thermal analysis reaction profiles are described that reflect changes in the uranium fluorination/oxidation state, physiochemical effects, and instances of discrete chemical speciation. Large differences in the onset temperatures for each system investigated implicate changes in mode of the NF 3 gas-solid surface interaction. These studies also demonstrate that NF 3 is a potential replacement fluorinating agent in the existing nuclear fuel cycle and in actinide volatility reprocessing.

The UC2-x - Carbon eutectic: A laser heating study

NASA Astrophysics Data System (ADS)

Manara, D.; Boboridis, K.; Morel, S.; De Bruycker, F.

2015-11-01

The UC2-x - carbon eutectic has been studied by laser heating and fast multi-wavelength pyrometry under inert atmosphere. The study has been carried out on three compositions, two of which close to the phase boundary of the UC2-x - C miscibility gap (with C/U atomic ratios 2 and 2.1), and one, more crucial, with a large excess of carbon (C/U = 2.82). The first two compositions were synthesised by arc-melting. This synthesis method could not be applied to the last composition, which was therefore completed directly by laser irradiation. The U - C - O composition of the samples was checked by using a combustion method in an ELTRA® analyser. The eutectic temperature, established to be 2737 K ± 20 K, was used as a radiance reference together with the cubic - tetragonal (α → β) solid state transition, fixed at 2050 K ± 20 K. The normal spectral emissivity of the carbon-richer compounds increases up to 0.7, whereas the value 0.53 was established for pure hypostoichiometric uranium dicarbide at the limit of the eutectic region. This increase is analysed in the light of the demixing of excess carbon, and used for the determination of the liquidus temperature (3220 K ± 50 K for UC2.82). Due to fast solid state diffusion, also fostered by the cubic - tetragonal transition, no obvious signs of a lamellar eutectic structure could be observed after quenching to room temperature. The eutectic surface C/UC2-x composition could be qualitatively, but consistently, followed during the cooling process with the help of the recorded radiance spectra. Whereas the external liquid surface is almost entirely constituted by uranium dicarbide, it gets rapidly enriched in demixed carbon upon freezing. Demixed carbon seems to quickly migrate towards the inner bulk during further cooling. At the α → β transition, uranium dicarbide covers again the almost entire external surface.

Legal, institutional, and political issues in transportation of nuclear materials at the back end of the LWR nuclear fuel cycle

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

Lippek, H.E.; Schuller, C.R.

1979-03-01

A study was conducted to identify major legal and institutional problems and issues in the transportation of spent fuel and associated processing wastes at the back end of the LWR nuclear fuel cycle. (Most of the discussion centers on the transportation of spent fuel, since this activity will involve virtually all of the legal and institutional problems likely to be encountered in moving waste materials, as well.) Actions or approaches that might be pursued to resolve the problems identified in the analysis are suggested. Two scenarios for the industrial-scale transportation of spent fuel and radioactive wastes, taken together, high-light mostmore » of the major problems and issues of a legal and institutional nature that are likely to arise: (1) utilizing the Allied General Nuclear Services (AGNS) facility at Barnwell, SC, as a temporary storage facility for spent fuel; and (2) utilizing AGNS for full-scale commercial reprocessing of spent LWR fuel.« less

Accumulation of Sellafield-derived radiocarbon ((14)C) in Irish Sea and West of Scotland intertidal shells and sediments.

PubMed

Tierney, Kieran M; Muir, Graham K P; Cook, Gordon T; MacKinnon, Gillian; Howe, John A; Heymans, Johanna J; Xu, Sheng

2016-01-01

The nuclear energy industry produces radioactive waste at various stages of the fuel cycle. In the United Kingdom, spent fuel is reprocessed at the Sellafield facility in Cumbria on the North West coast of England. Waste generated at the site comprises a wide range of radionuclides including radiocarbon ((14)C) which is disposed of in various forms including highly soluble inorganic carbon within the low level liquid radioactive effluent, via pipelines into the Irish Sea. This (14)C is rapidly incorporated into the dissolved inorganic carbon (DIC) reservoir and marine calcifying organisms, e.g. molluscs, readily utilise DIC for shell formation. This study investigated a number of sites located in Irish Sea and West of Scotland intertidal zones. Results indicate (14)C enrichment above ambient background levels in shell material at least as far as Port Appin, 265 km north of Sellafield. Of the commonly found species (blue mussel (Mytilus edulis), common cockle (Cerastoderma edule) and common periwinkle (Littorina littorea)), mussels were found to be the most highly enriched in (14)C due to the surface environment they inhabit and their feeding behaviour. Whole mussel shell activities appear to have been decreasing in response to reduced discharge activities since the early 2000s but in contrast, there is evidence of continuing enrichment of the carbonate sediment component due to in-situ shell erosion, as well as indications of particle transport of fine (14)C-enriched material close to Sellafield. Copyright © 2015 The Authors. Published by Elsevier Ltd.. All rights reserved.

Genetic polymorphisms in the IL-18 gene and ulcerative colitis risk: a meta-analysis.

PubMed

Wang, Ying; Tong, Jing; Chang, Bing; Wang, Bai-Fang; Zhang, Dai; Wang, Bing-Yuan

2014-07-01

This meta-analysis was performed to evaluate the relationships between genetic polymorphisms in the IL-18 gene and ulcerative colitis (UC) risk. The PubMed, CISCOM, CINAHL, Web of Science, Google Scholar, EBSCO, Cochrane Library, and CBM databases were searched for relevant articles published before November 1st, 2013 without any language restrictions. Meta-analysis was conducted using the STATA 12.0 software. Crude odds ratios (ORs) with their 95% confidence intervals (95% CI) were calculated. Eight case-control studies with a total of 1000 UC cases and 1392 healthy subjects met the inclusion criteria. Six common polymorphisms in the IL-18 gene were evaluated, including rs1946518 A>C, rs187238 G>C, rs917997 G>A, Codon35, rs1946519 C>A, and rs360718 A>C. The results of our meta-analysis suggest that the IL-18 rs1946518 (allele model: OR=1.22, 95% CI: 1.01-1.48, p=0.039; dominant model: OR=1.44, 95% CI: 1.01-2.06, p=0.045; respectively), rs187238 (allele model: OR=1.38, 95% CI: 1.19-1.61, p<0.001; dominant model: OR=1.50, 95% CI: 1.03-2.19, p=0.034; respectively), and rs360718 (allele model: OR=2.18, 95% CI: 1.22-3.90, p=0.008) polymorphisms might be strongly correlated with an increased risk of UC. A subgroup analysis was conducted to investigate the effect of ethnicity on an individual's risk of UC. Our results revealed positive significant correlations between IL-18 genetic polymorphisms and an increased risk of UC among Asians (allele model: OR=1.36, 95% CI: 1.16-1.60, p<0.001; dominant model: OR=1.50, 95% CI: 1.14-1.98, p=0.004; respectively) and Africans (allele model: OR=1.45, 95% CI: 1.03-2.05, p=0.034), but not among Caucasians (all p>0.05). Our findings provide convincing evidence that IL-18 genetic polymorphisms may contribute to susceptibility to UC, especially the rs1946518, rs187238, and rs360718 polymorphisms among Asians and Africans.

The Breast Cancer DNA Interactome

DTIC Science & Technology

2013-10-01

in Gheldof et al. with minor modifications [62]. HMEC, MCF7 and MDA-MB-231 cells (26107) were fixed in 2% formaldehyde in fresh medium for 10 min at... digested with 1500 U of HindIII (New England Biolabs Ipswich, MA) overnight at 37uC with shaking at 950 rpm. 200 ml of digested nuclei were removed for...assessing digestion efficiency by qPCR. The restriction enzyme was inactivated by the addition of 1.6% SDS and was incubated at 65uC for 20 min. The

Recruitment, Advancement and Retention of Women in the Physical Sciences at U.C. Irvine

NASA Astrophysics Data System (ADS)

Druffel, E. R.; Smecker-Hane, T.; Kehoe, P.; Bryant, S. V.

2004-12-01

Strategies for the recruitment, retention and advancement of women in the physical sciences at U.C. Irvine are presented. The NSF-funded ADVANCE Program has implemented several new initiatives. Among these are new requirements for recruitment committees, participation by school equity advisors, personalized mentoring programs and establishment of senior chairs. Progress towards our goals are reviewed and evaluated. Issues such as dual career couples and the balance between family/personal time and work are also addressed.

Uranium carbide dissolution in nitric solution: Sonication vs. silent conditions

NASA Astrophysics Data System (ADS)

Virot, Matthieu; Szenknect, Stéphanie; Chave, Tony; Dacheux, Nicolas; Moisy, Philippe; Nikitenko, Sergey I.

2013-10-01

The dissolution of uranium carbide (UC) in nitric acid media is considered by means of power ultrasound (sonication) or magnetic stirring. The induction period required to initiate UC dissolution was found to be dramatically shortened when sonicating a 3 M nitric solution (Ar, 20 kHz, 18 W cm-2, 20 °C). At higher acidity, magnetic stirring offers faster dissolution kinetics compared to sonication. Ultrasound-assisted UC dissolution is found to be passivated after ∼60% dissolution and remains incomplete whatever the acidity which is confirmed by ICP-AES, LECO and SEM-EDX analyses. In general, the kinetics of UC dissolution is linked to the in situ generation of nitrous acid in agreement with the general mechanism of UC dissolution; the nitrous acid formation is reported to be faster under ultrasound at low acidity due to the nitric acid sonolysis. The carbon balance shared between the gaseous, liquid, and solid phases is strongly influenced by the applied dissolution procedure and HNO3 concentration.

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

None

>Fundamental Alloying. Studies of crystal structures, reactions at metal surfaces, spectroscopy of molten salts, mechanical deformation, and alloy theory are reported. Long-Range Applied Metallurgy. A thermal comparator is described and the characteristic temperature of U0/sub 2/ determined. Sintering studies were carried out on ThO/sub 2/. The diffusion of fission products in fuel and of Al/sup 26/ and Mn/sup 54/ in Al and the reaction of Be with UC were studied. Transformation and oxidation data were obtained for a number of Zr alloys. Reactor Metallurgy. A large number of ceramic technology projects are described. Some corrosion data are given for metalsmore » exposed to impure He and molten fluorides. Studies were made of the fission-gas-retention Properties of ceramic fuel bodies. A large number of materials compatibility studies are described. The mechanical properties of some reactor materials were studied. Fabrication work was conducted to develop materials for application in low-, medium-, and high-temperature reactors or systems. A large number of new metallographic and nondestructive testing techniques are reported. Studies were carried out on the oxidation, carburization, and stability of alloys. Equipment for postirradiation examination is described. Preparation of some alloys and dispersion fuels by powder metallurgy methods was studied. The development of welding and brazing techniques for reactor materials is described. (D.L.C.)« less

Berberine ameliorates chronic relapsing dextran sulfate sodium-induced colitis in C57BL/6 mice by suppressing Th17 responses.

PubMed

Li, Yan-Hong; Xiao, Hai-Tao; Hu, Dong-Dong; Fatima, Sarwat; Lin, Cheng-Yuan; Mu, Huai-Xue; Lee, Nikki P; Bian, Zhao-Xiang

2016-08-01

Ulcerative colitis (UC) is an increasingly common condition particularly in developed countries. The lack of satisfactory treatment has fueled the search for alternative therapeutic strategies. In recent studies, berberine, a plant alkaloid with a long history of medicinal use in Chinese medicine, has shown beneficial effects against animal models of acute UC. However, UC usually presents as a chronic condition with frequent relapse in patients. How berberine will act on chronic UC remains unclear. In the present study, we adopted dextran sulfate sodium (DSS)-induced chronic relapsing colitis model to assess the ameliorating activity of berberine. Colitis was induced by two cycles of 2.0% DSS for five days followed by 14days of drinking water plus a third cycle consisting of DSS only for five days. The colitis mice were orally administered 20mg/kg berberine from day 13 onward for 30days and monitored daily. The body weight, stool consistency, and stool bleeding were recorded for determination of the disease activity index (DAI). At the end of treatment, animals were sacrificed and samples were collected and subjected to histological, RT-qPCR, Western blot, and LC-MS analyses. Lymphocytes were isolated from spleens and mesenteric lymph nodes (MLN) and cultured for flow cytometry analysis of IL-17 secretion from CD4(+) cells and the Th17 cell differentiation. Results showed that berberine significantly ameliorated the DAI, colon shortening, colon tissue injury, and reduction of colonic expression of tight junction (TJ) protein ZO-1 and occludin of colitis mice. Notably, berberine treatment pronouncedly reduced DSS-upregulated Th17-related cytokine (IL-17 and ROR-γt) mRNAs in the colon. Furthermore, the mRNA expression of IL-6 and IL-23, and the phosphorylation of STAT3 in colon tissues from DSS-treated mice were pronouncedly inhibited by berberine. Moreover, the up-regulation of IL-17 secretion from CD4(+) cells of spleens and MLNs caused by DSS were significantly reversed by berberine treatment. Furthermore, Th17 cell differentiation from naive CD4(+) cells isolated from above DSS colitis mice were suppressed by berberine in a concentration-dependent manner. In summary, we demonstrated for the first time that berberine reduced the severity of chronic relapsing DSS-induced colitis by suppressing Th17 responses. The demonstration of activity in this mouse model supports the possibility of clinical efficacy of berberine in treating chronic UC. Copyright © 2016 Elsevier Ltd. All rights reserved.

Creep behavior of uranium carbide-based alloys

NASA Technical Reports Server (NTRS)

Seltzer, M. S.; Wright, T. R.; Moak, D. P.

1975-01-01

The present work gives the results of experiments on the influence of zirconium carbide and tungsten on the creep properties of uranium carbide. The creep behavior of high-density UC samples follows the classical time-dependence pattern of (1) an instantaneous deformation, (2) a primary creep region, and (3) a period of steady-state creep. Creep rates for unalloyed UC-1.01 and UC-1.05 are several orders of magnitude greater than those measured for carbide alloys containing a Zr-C and/or W dispersoid. The difference in creep strength between alloyed and unalloyed materials varies with temperature and applied stress.

Evidence to Support Monitoring of Vedolizumab Trough Concentrations in Patients With Inflammatory Bowel Diseases.

PubMed

Dreesen, Erwin; Verstockt, Bram; Bian, Sumin; de Bruyn, Magali; Compernolle, Griet; Tops, Sophie; Noman, Maja; Van Assche, Gert; Ferrante, Marc; Gils, Ann; Vermeire, Séverine

2018-04-25

Trough concentrations of vedolizumab were found to correlate with clinical response in phase 3 studies of patients with ulcerative colitis (UC) or Crohn's disease (CD). Nevertheless, there are no solid data to support monitoring of vedolizumab trough concentrations in treated patients. We investigated the correlation between vedolizumab exposure and response in a real-world population and aimed to identify patient factors that affect exposure and response. We performed a retrospective cohort study of 179 consecutive patients (66 with UC and 113 with CD) who began vedolizumab therapy from September 1, 2015, through October 1, 2016, at University Hospitals Leuven, Belgium. Serum concentrations of vedolizumab were measured before all infusions up to week 30. Effectiveness endpoints included endoscopic healing (UC, Mayo endoscopic sub-score ≤1; CD, absence of ulcers), clinical response (physicians' global assessment), and biologic response or remission (based on level of C-reactive protein) and were assessed at week 14 (for patients with UC) and week 22 (for patients with CD). A stepwise forward addition-backward elimination modeling approach was performed to identify factors independently associated with vedolizumab exposure and response. Vedolizumab trough concentrations >30.0 μg/mL at week 2, >24.0 μg/mL at week 6, and >14.0 μg/mL during maintenance therapy associated with a higher probability of attaining the effectiveness endpoints for patients with UC or CD (P < .05). Higher body mass and more severe disease (based on high level of C-reactive protein and low level of albumin and/or hemoglobin) at the start of vedolizumab therapy associated with lower trough concentrations of vedolizumab over the 30-week period and a lower probability of achieving mucosal healing (P < .05). Mucosal healing was achieved in significantly more patients with UC than patients with CD, even though a diagnosis of UC was not an independent predictor of higher vedolizumab trough concentrations. In a retrospective study of 179 patients with CD or UC, we observed a correlation between vedolizumab exposure and response. These findings support monitoring of vedolizumab trough concentrations to predict patients' outcome. Copyright © 2018 AGA Institute. Published by Elsevier Inc. All rights reserved.

Interaction between BaCO{sub 3} and OPC/BFS composite cements at 20 {sup o}C and 60 {sup o}C

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

Utton, C.A., E-mail: c.utton@sheffield.ac.u; Gallucci, E.; Hill, J.

2011-03-15

A BaCO{sub 3} slurry, containing radioactive {sup 14}C, is produced during the reprocessing of spent nuclear fuel. This slurry is encapsulated in a Portland-blastfurnace slag composite cement. The effect of BaCO{sub 3} on the hydration of OPC and Portland-blastfurnace slag cements has been studied in this work. Samples containing a simulant BaCO{sub 3} slurry were cured for up to 720 days at 20 and 60 {sup o}C and analysed by XRD, SEM(EDX) and ICC. BaCO{sub 3} reacted with OPC to precipitate BaSO{sub 4} from a reaction between soluble sulfate and BaCO{sub 3}. Calcium monocarboaluminate subsequently formed from the carbonate released.more » The monocarboaluminate precipitated as crystals in voids formed during hydration. At 60 {sup o}C in OPC, it was not identified by XRD, suggesting the phase is unstable in this system around this temperature. In the Portland-blastfurnace slag cements containing BaCO{sub 3}, less monocarboaluminate and BaSO{sub 4} were formed, but the hydration of BFS was promoted and monocarboaluminate was stable up to 60 {sup o}C.« less

Nuclear physics research operation. Monthly report, November 1958

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

Faulkner, J.E.

1958-12-10

This report is a summary of projects worked on in support of the production reactors at Hanford. The projects include criticality studies, from tasks associated with fuel element reprocessing to shipments of slightly enriched uranium. They include studies of neutron cross sections for different reactions and neutron flux measurements in different reactor locations, as well as design studies for future reactor projects.

10 CFR Appendix B to Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Quality Assurance Criteria for Nuclear Power Plants and... LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Pt. 50, App. B Appendix B to Part 50—Quality Assurance... report a description of the quality assurance program to be applied to the design, fabrication...

10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

Code of Federal Regulations, 2010 CFR

2010-01-01

... transferred to a Federal repository no later than 10 years following separation of fission products from the.... Disposal of high-level radioactive fission product waste material will not be permitted on any land other... of the policy stated above with respect to high-level radioactive fission product wastes generated...

10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

Code of Federal Regulations, 2011 CFR

2011-01-01

... transferred to a Federal repository no later than 10 years following separation of fission products from the.... Disposal of high-level radioactive fission product waste material will not be permitted on any land other... of the policy stated above with respect to high-level radioactive fission product wastes generated...

No increased cancer risks from nuclear facilities

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

Not Available

1990-11-08

This article reports the results of a US Department of Health and Human Services (HHS) and National Cancer Institute (NCI) two-year survey that shows no increased risk of death from cancer for people living in counties containing or close to nuclear plants. 62 plants and their surrounding counties were included in the survey including commercial, US DOE and fuel reprocessing plants.

Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center

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

M. D. Staiger

2007-06-01

This report provides a quantitative inventory and composition (chemical and radioactivity) of calcined waste stored at the Idaho Nuclear Technology and Engineering Center. From December 1963 through May 2000, liquid radioactive wastes generated by spent nuclear fuel reprocessing were converted into a solid, granular form called calcine. This report also contains a description of the calcine storage bins.

Predictors of Outcome in Ulcerative Colitis.

PubMed

Waterman, Matti; Knight, Jo; Dinani, Amreen; Xu, Wei; Stempak, Joanne M; Croitoru, Kenneth; Nguyen, Geoffrey C; Cohen, Zane; McLeod, Robin S; Greenberg, Gordon R; Steinhart, A Hillary; Silverberg, Mark S

2015-09-01

Approximately 80% of patients with ulcerative colitis (UC) have intermittently active disease and up to 20% will require a colectomy, but little data available on predictors of poor disease course. The aim of this study was to identify clinical and genetic markers that can predict prognosis. Medical records of patients with UC with ≥5 years of follow-up and available DNA and serum were retrospectively assessed. Immunochip was used to genotype loci associated with immune mediated inflammatory disorders (IMIDs), inflammatory bowel diseases, and other single nucleotide polypmorphisms previously associated with disease severity. Serum levels of pANCA, ASCA, CBir1, and OmpC were also evaluated. Requirement for colectomy, medication, and hospitalization were used to group patients into 3 prognostic groups. Six hundred one patients with UC were classified as mild (n = 78), moderate (n = 273), or severe disease (n = 250). Proximal disease location frequencies at diagnosis were 13%, 21%, and 30% for mild, moderate, and severe UC, respectively (P = 0.001). Disease severity was associated with greater proximal extension rates on follow-up (P < 0.0001) and with shorter time to extension (P = 0.03) and to prednisone initiation (P = 0.0004). When comparing severe UC with mild and moderate UC together, diagnosis age >40 and proximal disease location were associated with severe UC (odds ratios = 1.94 and 2.12, respectively). None of the single nucleotide polypmorphisms or serum markers tested was associated with severe UC, proximal disease extension or colectomy. Older age and proximal disease location at diagnosis, but not genetic and serum markers, were associated with a more severe course. Further work is required to identify biomarkers that will predict outcomes in UC.

Flowsheets and source terms for radioactive waste projections

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

Forsberg, C.W.

1985-03-01

Flowsheets and source terms used to generate radioactive waste projections in the Integrated Data Base (IDB) Program are given. Volumes of each waste type generated per unit product throughput have been determined for the following facilities: uranium mining, UF/sub 6/ conversion, uranium enrichment, fuel fabrication, boiling-water reactors (BWRs), pressurized-water reactors (PWRs), and fuel reprocessing. Source terms for DOE/defense wastes have been developed. Expected wastes from typical decommissioning operations for each facility type have been determined. All wastes are also characterized by isotopic composition at time of generation and by general chemical composition. 70 references, 21 figures, 53 tables.

Liquidus temperature and chemical durability of selected glasses to immobilize rare earth oxides waste

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

Mohd Fadzil, Syazwani Binti; Hrma, Pavel R.; Schweiger, Michael J.

Pyroprocessing is a reprocessing method for managing and reusing used nuclear fuel (UNF) by dissolving it in an electrorefiner with a molten alkali or alkaline earth chloride salt mixture while avoiding wet reprocessing. Pyroprocessing UNF with a LiCl-KCl eutectic salt releases the fission products from the fuel and generates a variety of metallic and salt-based species, including rare earth (RE) chlorides. If the RE-chlorides are converted to oxides, borosilicate glass is a prime candidate for their immobilization because of its durability and ability to dissolve almost any RE waste component into the matrix at high loadings. Crystallization that occurs inmore » waste glasses as the waste loading increases may complicate glass processing and affect the product quality. This work compares three types of borosilicate glasses in terms of liquidus temperature (TL): the International Simple Glass designed by the International Working Group, sodium borosilicate glass developed by Korea Hydro and Nuclear Power, and the lanthanide aluminoborosilicate (LABS) glass established in the United States. The LABS glass allows the highest waste loadings (over 50 mass% RE2O3) while possessing an acceptable chemical durability.« less

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

Shinichi Aose; Takafumi Kitajima; Kouji Ogasawara

CPF (Chemical Processing Facility) was constructed at Nuclear Fuel Cycle Engineering Laboratories of JAEA (Japan Atomic Energy Agency) in 1980 as a basic research field where spent fuel pins from fast reactor (FR) and high level liquid waste can be dealt with. The renovation consists of remodeling of the CA-3 cell and the laboratory A, installation of globe boxes, hoods and analytical equipments to the laboratory C and the analytical laboratory. Also maintenance equipments in the CA-5 cell which had been out of order were repaired. The CA-3 cell is the main cell in which important equipments such as amore » dissolver, a clarifier and extractors are installed for carrying out the hot test using the irradiated FR fuel. Since the CPF had specialized originally in the research function for the Purex process, it was desired to execute the research and development of such new, various reprocessing processes. Formerly, equipments were arranged in wide space and connected with not only each other but also with utility supply system mainly by fixed stainless steel pipes. It caused shortage of operation space in flexibility for basic experimental study. Old equipments in the CA-3 cell including vessels and pipes were removed after successful decontamination, and new equipments were installed conformably to the new design. For the purpose of easy installation and rearranging the experimental equipments, equipments are basically connected by flexible pipes. Since dissolver is able to be easily replaced, various dissolution experiments is conducted. Insoluble residue generated by dissolution of spent fuel is clarified by centrifugal. This small apparatus is effective to space-saving. Mini mixer settlers or centrifugal contactors are put on to the prescribed limited space in front of the backside wall. Fresh reagents such as solvent, scrubbing and stripping solution are continuously fed from the laboratory A to the extractor by the reagent supply system with semi-automatic observation system. The in-cell crane in CA-5 was renovated to increase driving efficiency. At the renovation for the in-cell crane, full scale mockup test and 3D simulation test had been executed in advance. After the renovation, hot tests in the CPF had been resumed from JFY 2002. New equipments such as dissolver, extractor, electrolytic device, etc. were installed in CA-3 conformably to the new design laid out in order to ensure the function and space. Glove boxes in the analysis laboratory were renewed in order to let it have flexibility from the viewpoint of conducting basic experiments (ex. U crystallization). Glove boxes and hoods were newly installed in the laboratory A for basic research and analysis, especially on MA chemistries. One laboratory (the laboratory C) was established to research about dry reprocessing. The renovation of the CPF has been executed in order to contribute to the development on the advanced fast reactor fuel cycle system, which will give us many sort of technical subject and experimental theme to be solved in the 2. Generation of the CPF.« less

Plutonium: Advancing our Understanding to Support Sustainable Nuclear Fuel Cycles

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

Lines, Amanda M.; Adami, Susan R.; Casella, Amanda

With Global energy needs increasing, real energy solutions to meet demands now, are needed. Fossil fuels are not an ideal candidate to meet these needs because of their negative impact on the environment. Renewables such as wind and solar have huge potential, but still need major technological advancements (particularly in the area of battery storage) before they can effectively meet growing world needs. The best option for meeting large energy needs without a large carbon footprint is nuclear energy. Of course, nuclear energy can face a fair amount of opposition and concern. However, through modern engineering and science many ofmore » these concerns can now be addressed. Many safety concerns can be met by engineering advancements, but perhaps the biggest area of concern is what to do with the used nuclear fuel after it is removed from the reactor. Currently the United States (and several other countries) utilize an open fuel cycle, meaning fuel is only used once and then discarded. It should be noted that fuel coming out of a reactor has utilized approximately 1% of the total energy that could be produced by the uranium in the fuel rod. The answer here is to close the fuel cycle and recycle the nuclear materials. By reprocessing used nuclear fuel, all the U can be repurposed without requiring disposal. The various fission products can be removed and either discarded (hugely reduced waste volume) or more reasonably, utilized in specialty reactors to make more energy or needed research/medical isotopes. While reprocessing technology is currently advanced enough to meet energy needs, completing research to improve and better understand these techniques is still needed. Better understanding behavior of fission products is one area of important research. Despite it being discovered over 75 years ago, plutonium is still an exciting element to study because of the complex solution chemistry it exhibits. In aqueous solutions Pu can exist simultaneously in multiple oxidation states, including 3+, 4+, and 6+. It also readily forms a variety of metal-ligand complexes depending on solution pH and available ligands. Understanding of the behavior of Pu in solution remains an important area of research today, with relevance to developing sustainable nuclear fuel cycles, minimizing its impact on the environment, and detecting and preventing the spread of nuclear weapons technology.« less

IMPACT OF FISSION PRODUCTS IMPURITY ON THE PLUTONIUM CONTENT IN PWR MOX FUELS

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

Gilles Youinou; Andrea Alfonsi

2012-03-01

This report presents the results of a neutronics analysis done in response to the charter IFCA-SAT-2 entitled 'Fuel impurity physics calculations'. This charter specifies that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies (UOX SNF) is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate PWR MOX fuel assemblies. Only non-gaseous FP have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1). This mixture of Pu and FP is called PuFP. Note that, in this preliminary analysis,more » the FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.« less

Impact of firing temperature on multi-wavelength selective Stokes and anti-Stokes luminescent behavior by Gd2O2S:Er,Yb phosphor and its application in solar energy harvesting

NASA Astrophysics Data System (ADS)

Kataria, V.; Mehta, D. S.

2018-04-01

Erbium (Er3+)-ytterbium (Yb3+) doped gadolinium oxysulphide (Gd2O2S) phosphor has been developed via a facile method of solid-state flux fusion, and offers two-fold spectrum modification with highly intense Stokes and anti-Stokes shift. The effect of the firing cycle on the photoluminescent response and morphology of Gd2O2S:Er,Yb is scrutinized, wherein the firing temperature was varied (1000 °C-1250 °C), keeping firing time and all other parameters constant. Interestingly, the nanostructures fired below 1150 °C showed nanorods of diameter ~200 nm and length ~1-2 µm, whereas firing at 1150 °C and above rendered nanospheres with small diameter, ~350 nm. Highly bright upconversion (UC) emission was achieved even under an extremely low excitation power density of 800 µW cm-2 from a 980 nm laser, and was comfortably visible to the naked eye. The incident power dependent studies disclosed increase in UC-emission intensity with increasing excitation power and a quasi-linear dependence on excitation power density. Intense characteristic UC-emission of Er3+ excited states at 525 nm, 556 nm and 668 nm were observed, and the green emission band was found to be dominant over the red band in intensity. Concurrently, downconversion (DC) emission at 556 nm and 669 nm was also exhibited under ultraviolet excitation (285 nm and 380 nm), with the red band being more powerful than the green, unlike UC-emission. Firing temperature dependent studies divulged the dependence of luminescence intensity on the firing cycle of the luminophore and formation of the respective luminescent phase. The UC-emission intensity was found to be maximum for samples fired at 1150 °C, whereas samples fired at 1000 °C showed the highest DC-emission intensity. The excitation and emission profile of single Gd2O2S:Er,Yb phosphor lying in the desired spectral region and as a dual spectral converter marks its possible application for enhanced harvesting of sunlight.

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

Washiya, Tadahiro; Komaki, Jun; Funasaka, Hideyuki

Japan Atomic Energy Agency (JAEA) has been developing the new aqueous reprocessing system named 'NEXT' (New Extraction system for TRU recovery)1-2, which provides many advantages as waste volume reduction, cost savings by advanced components and simplification of process operation. Advanced head-end systems in the 'NEXT' process consist of fuel disassembly system, fuel shearing system and continuous dissolver system. We developed reliable fuel disassembly system with innovative procedure, and short-length shearing system and continuous dissolver system can be provided highly concentrated dissolution to adapt to the uranium crystallization process. We have carried out experimental studies, and fabrication of engineering-scale test devicesmore » to confirm the systems performance. In this paper, research and development of advanced head-end systems are described. (authors)« less

Scalable, cGMP-compatible purification of extracellular vesicles carrying bioactive human heterodimeric IL-15/lactadherin complexes

PubMed Central

Watson, Dionysios C.; Yung, Bryant C.; Bergamaschi, Cristina; Chowdhury, Bhabadeb; Bear, Jenifer; Stellas, Dimitris; Morales-Kastresana, Aizea; Jones, Jennifer C.; Felber, Barbara K.; Chen, Xiaoyuan; Pavlakis, George N.

2018-01-01

ABSTRACT The development of extracellular vesicles (EV) for therapeutic applications is contingent upon the establishment of reproducible, scalable, and high-throughput methods for the production and purification of clinical grade EV. Methods including ultracentrifugation (U/C), ultrafiltration, immunoprecipitation, and size-exclusion chromatography (SEC) have been employed to isolate EV, each facing limitations such as efficiency, particle purity, lengthy processing time, and/or sample volume. We developed a cGMP-compatible method for the scalable production, concentration, and isolation of EV through a strategy involving bioreactor culture, tangential flow filtration (TFF), and preparative SEC. We applied this purification method for the isolation of engineered EV carrying multiple complexes of a novel human immunostimulatory cytokine-fusion protein, heterodimeric IL-15 (hetIL-15)/lactadherin. HEK293 cells stably expressing the fusion cytokine were cultured in a hollow-fibre bioreactor. Conditioned medium was collected and EV were isolated comparing three procedures: U/C, SEC, or TFF + SEC. SEC demonstrated comparable particle recovery, size distribution, and hetIL-15 density as U/C purification. Relative to U/C, SEC preparations achieved a 100-fold reduction in ferritin concentration, a major protein-complex contaminant. Comparative proteomics suggested that SEC additionally decreased the abundance of cytoplasmic proteins not associated with EV. Combination of TFF and SEC allowed for bulk processing of large starting volumes, and resulted in bioactive EV, without significant loss in particle yield or changes in size, morphology, and hetIL-15/lactadherin density. Taken together, the combination of bioreactor culture with TFF + SEC comprises a scalable, efficient method for the production of highly purified, bioactive EV carrying hetIL-15/lactadherin, which may be useful in targeted cancer immunotherapy approaches. PMID:29535850

Simulation of ground-water flow near the nuclear-fuel reprocessing facility at the Western New York Nuclear Service Center, Cattaraugus County, New York

USGS Publications Warehouse

Yager, R.M.

1987-01-01

A two-dimensional finite-difference model was developed to simulate groundwater flow in a surficial sand and gravel deposit underlying the nuclear fuel reprocessing facility at Western New York Nuclear Service Center near West Valley, N.Y. The sand and gravel deposit overlies a till plateau that abuts an upland area of siltstone and shale on its west side, and is bounded on the other three sides by deeply incised stream channels that drain to Buttermilk Creek, a tributary to Cattaraugus Creek. Radioactive materials are stored within the reprocessing plant and are also buried within a till deposit at the facility. Tritiated water is stored in a lagoon system near the plant and released under permit to Franks Creek, a tributary to Buttermilk Creek. Groundwater levels predicted by steady-state simulations closely matched those measured in 23 observation wells, with an average error of 0.5 meter. Simulated groundwater discharges to two stream channels and a subsurface drain were within 5% of recorded values. Steady-state simulations used an average annual recharge rate of 46 cm/yr; predicted evapotranspiration loss from the ground was 20 cm/yr. The lateral range in hydraulic conductivity obtained through model calibration was 0.6 to 10 m/day. Model simulations indicated that 33% of the groundwater discharged from the sand and gravel unit (2.6 L/sec) is lost by evapotranspiration, 3% (3.0 L/sec) flows to seepage faces at the periphery of the plateau, 20% (1.6 L/sec) discharges to stream channels that drain a large wetland area near the center of the plateau, and the remaining 8% (0.6 L/sec) discharges to a subsurface french drain and to a wastewater treatment system. Groundwater levels computed by a transient-state simulation of an annual climatic cycle, including seasonal variation in recharge and evapotranspiration, closely matched water levels measured in eight observation wells. The model predicted that the subsurface drain and the stream channel that drains the wetland would intercept most of the recharge originating near the reprocessing plant. (Lantz-PTT)

Reprocessing of LiH in Molten Chlorides

NASA Astrophysics Data System (ADS)

Masset, Patrick J.; Gabriel, Armand; Poignet, Jean-Claude

2008-06-01

LiH was used as inactive material to stimulate the reprocessing of lithium tritiate in molten chlorides. The electrochemical properties (diffusion coefficients, apparent standard potentials) were measured by means of transient electrochemical techniques (cyclic voltammetry and chronopotentiometry). At 425 ºC the diffusion coefficient and the apparent standard potential were 2.5 · 10-5 cm2 s-1 and -1.8 V vs. Ag/AgCl, respectively. For the process design the LiH solubility was measured by means of DTA to optimize the LiH concentration in the molten phase. In addition electrolysis tests were carried out at 460 ºC with current densities up to 1 A cm-2 over 24 h. These results show that LiH may be reprocessed in molten chlorides consisting in the production of hydrogen gas at the anode and molten metallic lithium at the cathode.

TNFalpha and IL-10 gene polymorphisms in inflammatory bowel disease. Association of -1082 AA low producer IL-10 genotype with steroid dependency.

PubMed

Castro-Santos, Patricia; Suarez, Ana; López-Rivas, Laureano; Mozo, Lourdes; Gutierrez, Carmen

2006-05-01

An altered production of cytokines underlies inflammatory bowel disease (IBD) susceptibility. Various polymorphisms at the IL-10 and TNFalpha gene promoters control cytokine production levels. The influence of these polymorphisms on susceptibility to ulcerative colitis (UC) and Crohn's disease (CD) and their association with clinical features were analyzed. Genetic polymorphisms of TNFalpha (-308 G/A) and IL-10 (-1082 G/A, -812 C/T, and -592 C/A) were determined using the LightCycler system with hybridization probes matched with one sequence variant. The study population included 99 UC patients, 146 CD patients, and 343 matched controls. We did not find association between TNFalpha or IL-10 gene polymorphisms and UC or CD susceptibility, though a slight influence of -1082*G allele in UC appearance was observed. In a stratified analysis, a highly significant association between the -1082 AA IL-10 genotype and the steroid dependency was observed in IBD (p < 0.0001), contributing both UC (p = 0.004) and CD (p = 0.003) to this association. In contrast, TNFalpha genotypes did not influence steroid dependency in IBD. Further, the contribution of cytokine genotypes and of clinical features to the appearance of steroid-dependent status (dependent variable) was studied by multivariate analysis. The steroid-dependent phenotype correlated in UC with extensive disease (p = 0.010) and with the low producer -1082 AA IL-10 genotype (p = 0.002) and in CD with penetrating disease (p = 0.010), arthritis (p = 0.011), and the -1082 AA IL-10 genotype (p = 0.006). The main conclusion is that carriage of the -1082 AA IL-10 genotype (low producer) is a relevant risk factor for developing steroid-dependent IBD.

Potential link between Fusobacterium enrichment and DNA methylation accumulation in the inflammatory colonic mucosa in ulcerative colitis

PubMed Central

Tahara, Tomomitsu; Hirata, Ichiro; Nakano, Naoko; Tahara, Sayumi; Horiguchi, Noriyuki; Kawamura, Tomohiko; Okubo, Masaaki; Ishizuka, Takamitsu; Yamada, Hyuga; Yoshida, Dai; Ohmori, Takafumi; Maeda, Kohei; Komura, Naruomi; Ikuno, Hirokazu; Jodai, Yasutaka; Kamano, Toshiaki; Nagasaka, Mitsuo; Nakagawa, Yoshihito; Tuskamoto, Tetsuya; Urano, Makoto; Shibata, Tomoyuki; Kuroda, Makoto; Ohmiya, Naoki

2017-01-01

BACKGROUND AND AIM Fusobacterium enrichment has been associated with colorectal cancer development. Ulcerative colitis (UC) associated tumorigenesis is characterized as high degree of methylation accumulation through continuous colonic inflammation. The aim of this study was to investigate a potential link between Fusobacterium enrichment and DNA methylation accumulation in the inflammatory colonic mucosa in UC. METHODS In the candidate analysis, inflamed colonic mucosa from 86 UC patients were characterized the methylation status of colorectal a panel of cancer related 24 genes. In the genome-wide analysis, an Infinium HumanMethylation450 BeadChip array was utilized to characterize the methylation status of >450,000 CpG sites for fourteen UC patients. Results were correlated with Fusobacterium status. RESULTS UC with Fusobacterium enrichment (FB-high) was characterized as high degree of type C (for cancer-specific) methylation compared to other (FB-low/neg) samples (P<0.01). Genes hypermethylated in FB-high samples included well-known type C genes in colorectal cancer, such as MINT2 and 31, P16 and NEUROG1. Multivariate analysis demonstrated that the FB high status held an increased likelihood for methylation high as an independent factor (odds ratio: 16.18, 95% confidence interval: 1.94-135.2, P=0.01). Genome-wide methylation analysis demonstrated a unique methylome signature of FB-high cases irrespective of promoter, outside promoter, CpG and non-CpG sites. Group of promoter CpG sites that were exclusively hypermethylated in FB-high cases significantly codified the genes related to the catalytic activity (P=0.039). CONCLUSION Our findings suggest that Fusobacterium accelerates DNA methylation in specific groups of genes in the inflammatory colonic mucosa in UC. PMID:28977914

Direct comparison of two different mesalamine formulations for the induction of remission in patients with ulcerative colitis: A double-blind, randomized study

PubMed Central

Ito, Hiroaki; Iida, Mitsuo; Matsumoto, Takayuki; Suzuki, Yasuo; Sasaki, Hidetaka; Yoshida, Toyomitsu; Takano, Yuichi; Hibi, Toshifumi

2010-01-01

Background: Mesalamine is the first-line drug for the treatment of ulcerative colitis (UC). We directly compared the efficacy and safety of two mesalamine formulations for the induction of remission in patients with UC. Methods: In a multicenter, double-blind, randomized study, 229 patients with mild-to-moderate active UC were assigned to 4 groups: 66 and 65 received a pH-dependent release formulation of 2.4 g/day (pH-2.4 g) or 3.6 g/day (pH-3.6 g), respectively; 65 received a time-dependent release formulation of 2.25 g/day (Time-2.25 g), and 33 received placebo (Placebo). The drugs were administered three times daily for eight weeks. The primary endpoint was a decrease in the UC disease activity index (UC-DAI). Results: In the full analysis set (n = 225) the decrease in UC-DAI in each group was 1.5 in pH-2.4 g, 2.9 in pH-3.6 g, 1.3 in Time-2.25 g and 0.3 in Placebo, respectively. These results demonstrate the superiority of pH-3.6 g over Time-2.25 g (P = 0.003) and the noninferiority of pH-2.4 g to Time-2.25 g. Among the patients with proctitis-type UC, a significant decrease in UC-DAI was observed in pH-2.4 g and pH-3.6 g as compared to Placebo, but not in Time-2.25 g. No differences were observed in the safety profiles. Conclusions: Higher dose of the pH-dependent release formulation was more effective for induction of remission in patients with mild-to-moderate active UC. Additionally, the pH-dependent release formulation was preferable to the time-dependent release formulation for patients with proctitis-type UC (UMIN Clinical Trials Registry, no. C000000288). (Inflamm Bowel Dis 2010) PMID:20049950

Quantity and management of spent fuel from prototype and research reactors in Germany

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

Dorr, Sabine; Bollingerfehr, Wilhelm; Filbert, Wolfgang

Within the scope of an R and D project (project identification number FKZ 02 S 8679) sponsored by BMBF (Federal Ministry of Education and Research), the current state of storage and management of fuel elements from prototype and research reactors was established, and an approach for their future storage/management was developed. The spent fuels from prototype and research reactors in Germany that require disposal were specified and were described in regard to their repository-relevant characteristics. As there are currently no casks licensed for disposal in Germany, descriptions of casks that were considered to be suitable were provided. Based on themore » information provided on the spent fuel from prototype and research reactors and the potential casks, a technical disposal concept was developed. In this context, concepts to integrate the spent fuel from prototype and research reactors into existing disposal concepts for spent fuel from German nuclear power plants and for waste from reprocessing were developed for salt and clay formations. (authors)« less

Evolution of spent nuclear fuel in dry storage conditions for millennia and beyond

NASA Astrophysics Data System (ADS)

Wiss, Thierry; Hiernaut, Jean-Pol; Roudil, Danièle; Colle, Jean-Yves; Maugeri, Emilio; Talip, Zeynep; Janssen, Arne; Rondinella, Vincenzo; Konings, Rudy J. M.; Matzke, Hans-Joachim; Weber, William J.

2014-08-01

Significant amounts of spent uranium dioxide nuclear fuel are accumulating worldwide from decades of commercial nuclear power production. While such spent fuel is intended to be reprocessed or disposed in geologic repositories, out-of-reactor radiation damage from alpha decay can be detrimental to its structural stability. Here we report on an experimental study in which radiation damage in plutonium dioxide, uranium dioxide samples doped with short-lived alpha-emitters and urano-thorianite minerals have been characterized by XRD, transmission electron microscopy, thermal desorption spectrometry and hardness measurements to assess the long-term stability of spent nuclear fuel to substantial alpha-decay doses. Defect accumulation is predicted to result in swelling of the atomic structure and decrease in fracture toughness; whereas, the accumulation of helium will produce bubbles that result in much larger gaseous-induced swelling that substantially increases the stresses in the constrained spent fuel. Based on these results, the radiation-ageing of highly-aged spent nuclear fuel over more than 10,000 years is predicted.

Comparison of treatment outcomes according to output voltage during shockwave lithotripsy for ureteral calculi: a prospective randomized multicenter study.

PubMed

Park, Jinsung; Kim, Hong-Wook; Hong, Sungwoo; Yang, Hee Jo; Chung, Hong

2015-05-01

To investigate the effect of fixed versus escalating voltage during SWL on treatment outcomes in patients with ureteral calculi (UC). A prospective, randomized, multicenter trial was conducted on 120 patients who were diagnosed with a single radiopaque UC. The patients were randomized into group C (n = 60, constant 13 kV, 3,000 shock wave, 2 Hz) or group E (n = 60, 11.4-12.0-13 kV per 1,000 shock waves, 2 Hz). They were evaluated by plain abdominal radiography and urinalysis at 1 week after a single session of SWL, and repeat SWL was performed if needed. The primary endpoint was stone-free rate at 1 week (SFR1) after SWL. Secondary endpoints were post-SWL visual pain score (VPS), oral analgesic requirements during 1 week, and cumulative SFRs after the second and third sessions of SWL. Groups C and E were well balanced in terms of baseline patients and stone characteristics, including pre-SWL VPS, stone location, and stone size (6.24 ± 1.92 vs. 6.30 ± 2.13 mm). SFR1s were not significantly different between groups C and E (60.0 vs. 68.3%, p = 0.447). Analyses stratified by stone size (<6 vs. ≥6 mm) showed no difference in SFR1 (p = 0.148 vs. 0.808). In the analyses stratified by stone location, group E tended to be more effective in distal UC (81.0 vs. 50.0%, p = 0.052), whereas no difference was seen in proximal UC (p = 0.487). Secondary endpoints were also similar between the two groups. Our results suggest that voltage escalation during SWL in UC may not provide superior stone fragmentation compared to fixed voltage.

Effect of reprocessing cycles on the degradation of polypropylene copolymer filled with talc or montmorillonite during injection molding process

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

Demori, R.; Mauler, R. S., E-mail: raquel.mauler@ufrgs.br; Ashton, E.

Mechanical recycling of polymeric materials is a favorable technique resulting in economic and environmental benefits, especially in the case of polymers with a high production volume as the polypropylene copolymer (PP). However, recycling by reprocessing techniques can lead to thermal, mechanical or thermo-oxidative degradation that can affect the structure of the polymer and subsequently the material properties. PP filled with montmorillonite (MMT) or talc are widely produced and studied, however, its degradation reactions by reprocessing cycles are poorly studied so far. In this study, the effects of reprocessing cycles in the structure and in the properties of the PP/MMT andmore » PP/Talc were evaluated. The samples were mixed with 5% talc or MMT Cloisite C15A in a twin-screw extrusion. After extrusion, this filled material was submitted to five reprocessing cycles through an injection molding process. In order to evaluate the changes induced by reprocessing techniques, the samples were characterized by DSC, FT-IR, Izod impact and tensile strength tests. The study showed that Young modulus, elongation at brake and Izod impact were not affected by reprocessing cycles, except when using talc. In this case, the elongation at brake reduced until the fourth cycle, showing rigidity increase. The DSC results showed that melting and crystallization temperature were not affected. A comparison of FT-IR spectra of the reprocessed indicated that in both samples, between the first and the fifth cycle, no noticeable change has occurred. Thus, there is no evidence of thermo oxidative degradation. In general, these results suggest that PP reprocessing cycles using MMT or talc does not change the material properties until the fifth cycle.« less

Determination of the NPP Kr\\vsko spent fuel decay heat

NASA Astrophysics Data System (ADS)

Kromar, Marjan; Kurinčič, Bojan

2017-07-01

Nuclear fuel is designed to support fission process in a reactor core. Some of the isotopes, formed during the fission, decay and produce decay heat and radiation. Accurate knowledge of the nuclide inventory producing decay heat is important after reactor shut down, during the fuel storage and subsequent reprocessing or disposal. In this paper possibility to calculate the fuel isotopic composition and determination of the fuel decay heat with the Serpent code is investigated. Serpent is a well-known Monte Carlo code used primarily for the calculation of the neutron transport in a reactor. It has been validated for the burn-up calculations. In the calculation of the fuel decay heat different set of isotopes is important than in the neutron transport case. Comparison with the Origen code is performed to verify that the Serpent is taking into account all isotopes important to assess the fuel decay heat. After the code validation, a sensitivity study is carried out. Influence of several factors such as enrichment, fuel temperature, moderator temperature (density), soluble boron concentration, average power, burnable absorbers, and burnup is analyzed.

2005 Tri-Service Infrastructure Systems Conference and Exhibition. Volume 10, Track 12

DTIC Science & Technology

2005-08-04

Schmidt Track 12 Greenup L&D Miter Gate Repair and...j uv en ile fi sh p as sa ge o n ty pi ca l d es ig n cr ite ria R ob er t B uc hh ol z H yd ra ul ic d es ig n of ju ve ni le fis h pa ss...C hi ca go se w er sy st em s E rn es to G o T R A C K 1 8 C iv il M e ch a n ic a l S e ss io n 1 8 D N ew c oa tin g pr od uc ts fo

TRMM Data Improvement as Part of the GPM Data Processing

NASA Technical Reports Server (NTRS)

Stocker, Erich F.; Ji, Y.; Kwiatkowski, J.; Kelley, O.; Stout, J.; Woltz, L.

2016-01-01

NASA has a long standing commitment to the improvement of its mission datasets. Indeed, data reprocessing is always built into the plans, schedule and budget for the mission data processing system. However, in addition to these ongoing mission reprocessing, NASA also supports a final reprocessing of all the data for a mission upon its completion (known as Phase F). TRMM Phase F started with the end of the TRMM mission in June of 2015. This last reprocessing has two overall goals: improvement of the TRMM mission data products; incorporation of the 17+ years of TRMM data into the ongoing NASA/JAXA GPM data processing. The first goal guarantees that the latest algorithms used for precipitation retrievals will also be used in reprocessing the TRMM data. The second goal ensures that as GPM algorithms are improved, the entire TRMM data will always be reprocessed with each GPM reprocessing. In essence TRMM becomes another of the GPM constellation satellites. This paper will concentrate on presenting the improvements to TMI level 1 data including calibration, geolocation, and emissive antenna corrections. It will describe the format changes that will occur how the TMI level 1C product will be intercalibrated using GMI as the reference calibration. It will also provide an overview of changes in the precipitation radar products as well as the combined TMIPR product.

Effect of DECIDE (Decision-making Education for Choices In Diabetes Everyday) Program Delivery Modalities on Clinical and Behavioral Outcomes in Urban African Americans With Type 2 Diabetes: A Randomized Trial.

PubMed

Fitzpatrick, Stephanie L; Golden, Sherita Hill; Stewart, Kerry; Sutherland, June; DeGross, Sharie; Brown, Tina; Wang, Nae-Yuh; Allen, Jerilyn; Cooper, Lisa A; Hill-Briggs, Felicia

2016-12-01

To compare the effectiveness of three delivery modalities of Decision-making Education for Choices In Diabetes Everyday (DECIDE), a nine-module, literacy-adapted diabetes and cardiovascular disease (CVD) education and problem-solving training, compared with an enhanced usual care (UC), on clinical and behavioral outcomes among urban African Americans with type 2 diabetes. Eligible participants (n = 182) had a suboptimal CVD risk factor profile (A1C, blood pressure, and/or lipids). Participants were randomized to DECIDE Self-Study (n = 46), DECIDE Individual (n = 45), DECIDE Group (n = 46), or Enhanced UC (n = 45). Intervention duration was 18-20 weeks. Outcomes were A1C, blood pressure, lipids, problem-solving, disease knowledge, and self-care activities, all measured at baseline, 1 week, and 6 months after completion of the intervention. DECIDE modalities and Enhanced UC did not significantly differ in clinical outcomes at 6 months postintervention. In participants with A1C ≥7.5% (58 mmol/mol) at baseline, A1C declined in each DECIDE modality at 1 week postintervention (P < 0.05) and only in Self-Study at 6 months postintervention (b = -0.24, P < 0.05). There was significant reduction in systolic blood pressure in Self-Study (b = -4.04) and Group (b = -3.59) at 6 months postintervention. Self-Study, Individual, and Enhanced UC had significant declines in LDL and Self-Study had an increase in HDL (b = 1.76, P < 0.05) at 6 months postintervention. Self-Study and Individual had a higher increase in knowledge than Enhanced UC (P < 0.05), and all arms improved in problem-solving (P < 0.01) at 6 months postintervention. DECIDE modalities showed benefits after intervention. Self-Study demonstrated robust improvements across clinical and behavioral outcomes, suggesting program suitability for broader dissemination to populations with similar educational and literacy levels. © 2016 by the American Diabetes Association.

Oncogenic Smad3 signaling induced by chronic inflammation is an early event in ulcerative colitis-associated carcinogenesis.

PubMed

Kawamata, Seiji; Matsuzaki, Koichi; Murata, Miki; Seki, Toshihito; Matsuoka, Katsuyoshi; Iwao, Yasushi; Hibi, Toshifumi; Okazaki, Kazuichi

2011-03-01

Both chronic inflammation and somatic mutations likely contribute to the pathogenesis of ulcerative colitis (UC)-associated dysplasia and cancer. On the other hand, both tumor suppression and oncogenesis can result from transforming growth factor (TGF)-β signaling. TGF-β type I receptor (TβRI) and Ras-associated kinases differentially phosphorylate a mediator, Smad3, to become C-terminally phosphorylated Smad3 (pSmad3C), linker phosphorylated Smad3 (pSmad3L), and both C-terminally and linker phosphorylated Smad3 (pSmad3L/C). The pSmad3C/p21(WAF1) pathway transmits a cytostatic TGF-β signal, while pSmad3L and pSmad3L/C promote cell proliferation by upregulating c-Myc oncoprotein. The purpose of this study was to clarify the alteration of Smad3 signaling during UC-associated carcinogenesis. By immunostaining and immunofluorescence, we compared pSmad3C-, pSmad3L-, and pSmad3L/C-mediated signaling in colorectal specimens representing colitis, dysplasia, or cancer from eight UC patients with signaling in normal colonic crypts. We also investigated p53 expression and mutations of p53 and K-ras genes. We further sought functional meaning of the phosphorylated Smad3-mediated signaling in vitro. As enterocytes in normal crypts migrated upward toward the lumen, cytostatic pSmad3C/p21(WAF1) tended to increase, while pSmad3L/c-Myc shown by progenitor cells gradually decreased. Colitis specimens showed prominence of pSmad3L/C/c-Myc, mediated by TGF-β and tumor necrosis factor (TNF)-α, in all enterocyte nuclei throughout entire crypts. In proportion with increases in frequency of p53 and K-ras mutations during progression from dysplasia to cancer, the oncogenic pSmad3L/c-Myc pathway came to be dominant with suppression of the pSmad3C/p21(WAF1) pathway. Oncogenic Smad3 signaling, altered by chronic inflammation and eventually somatic mutations, promotes UC-associated neoplastic progression by upregulating growth-related protein. Copyright © 2010 Crohn's & Colitis Foundation of America, Inc.

College Readiness as a Graduation Requirement: An Assessment of San Diego's Challenges

ERIC Educational Resources Information Center

Betts, Julian R.; Zau, Andrew C.; Bachofer, Karen Volz

2013-01-01

To be considered for admission to the University of California (UC) or the California State University (CSU) system, high school students must complete all a-g courses with grades of C or higher. The a-g course sequence includes 30 semesters of UC-approved college preparatory coursework in seven subject areas, and completion indicates a high level…

Accelerator-driven Transmutation of Waste

NASA Astrophysics Data System (ADS)

Venneri, Francesco

1998-04-01

Nuclear waste from commercial power plants contains large quantities of plutonium, other fissionable actinides, and long-lived fission products that are potential proliferation concerns and create challenges for the long-term storage. Different strategies for dealing with nuclear waste are being followed by various countries because of their geologic situations and their views on nuclear energy, reprocessing and non-proliferation. The current United States policy is to store unprocessed spent reactor fuel in a geologic repository. Other countries are opting for treatment of nuclear waste, including partial utilization of the fissile material contained in the spent fuel, prior to geologic storage. Long-term uncertainties are hampering the acceptability and eventual licensing of a geologic repository for nuclear spent fuel in the US, and driving up its cost. The greatest concerns are with the potential for radiation release and exposure from the spent fuel for tens of thousands of years and the possible diversion and use of the actinides contained in the waste for weapons construction. Taking advantage of the recent breakthroughs in accelerator technology and of the natural flexibility of subcritical systems, the Accelerator-driven Transmutation of Waste (ATW) concept offers the United States and other countries the possibility to greatly reduce plutonium, higher actinides and environmentally hazardous fission products from the waste stream destined for permanent storage. ATW does not eliminate the need for, but instead enhances the viability of permanent waste repositories. Far from being limited to waste destruction, the ATW concept also brings to the table new technologies that could be relevant for next-generation power producing reactors. In the ATW concept, spent fuel would be shipped to the ATW site where the plutonium, transuranics and selected long-lived fission products would be destroyed by fission or transmutation in their first and only pass through the facility, using an accelerator-driven subcritical burner cooled by liquid lead/bismuth and limited pyrochemical treatment of the spent fuel and residual waste. This approach contrasts with the present-day practices of aqueous reprocessing (Europe and Japan), in which high purity plutonium is produced and used in the fabrication of fresh mixed oxide fuel (MOX) that is shipped off-site for use in light water reactors.

Experience with soluble neutron poisons for criticality control at ICPP

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

Wilson, R.E.; Mortimer, S.R.

1978-01-01

Soluble neutron poisons assure criticality control in two of the headend fuel reprocessing systems at the Idaho Chemical Processing Plant. Soluble poisons have been used successfully since 1964 and will be employed in the projected new headend processes. The use of soluble poisons (1) greatly increases the process output (2) allows versatility in the size of fuel assemblies processed and (3) allows the practical reprocessing of some fuels. The safety limit for all fluids entering the U-Zr alloy dissolver is 3.6 g/liter boron. To allow for possible deviations in the measurement systems and drift between analytical sampling periods, the standardmore » practice is to use 3.85 g/liter boron as the lower limit. This dissolver has had 4000 successful hours of operation using soluble poisons. The electrolytic dissolution process depends on soluble gadolinium for criticality safety. This system is used to process high enriched uranium clad in stainless steel. Electrolytic dissolution takes advantage of the anodic corrosion that occurs when a large electrical current is passed through the fuel elements in a corrosive environment. Three control methods are used on each headend system. First, the poison is mixed according to standard operating procedures and the measurements are affirmed by the operator's supervisor. Second, the poisoned solution is stirred, sampled, analyzed, and the analysis reported while still in the mix tank. Finally, a Nuclear Poison Detection System (NPDS) must show an acceptable poison concentration before the solution can be transferred. The major disadvantage of using soluble poisons is the need for very sophisticated control systems and procedures, which require extensive checkout. The need for a poisoned primary heating and cooling system means a secondary system is needed as well. Experience has shown, however, that production enhancement more than makes up for the problems.« less

Demand driven salt clean-up in a molten salt fast reactor – Defining a priority list

PubMed Central

Litskevich, D.; Gregg, R.; Mount, A. R.

2018-01-01

The PUREX technology based on aqueous processes is currently the leading reprocessing technology in nuclear energy systems. It seems to be the most developed and established process for light water reactor fuel and the use of solid fuel. However, demand driven development of the nuclear system opens the way to liquid fuelled reactors, and disruptive technology development through the application of an integrated fuel cycle with a direct link to reactor operation. The possibilities of this new concept for innovative reprocessing technology development are analysed, the boundary conditions are discussed, and the economic as well as the neutron physical optimization parameters of the process are elucidated. Reactor physical knowledge of the influence of different elements on the neutron economy of the reactor is required. Using an innovative study approach, an element priority list for the salt clean-up is developed, which indicates that separation of Neodymium and Caesium is desirable, as they contribute almost 50% to the loss of criticality. Separating Zirconium and Samarium in addition from the fuel salt would remove nearly 80% of the loss of criticality due to fission products. The theoretical study is followed by a qualitative discussion of the different, demand driven optimization strategies which could satisfy the conflicting interests of sustainable reactor operation, efficient chemical processing for the salt clean-up, and the related economic as well as chemical engineering consequences. A new, innovative approach of balancing the throughput through salt processing based on a low number of separation process steps is developed. Next steps for the development of an economically viable salt clean-up process are identified. PMID:29494604

Corrosion and Microstructure Correlation in Molten LiCl-KCl Medium

NASA Astrophysics Data System (ADS)

Ravi Shankar, A.; Mathiya, S.; Thyagarajan, K.; Kamachi Mudali, U.

2010-07-01

Pyrochemical reprocessing in molten chloride salt medium has been considered as one of the best options for the reprocessing of spent metallic fuels of future fast breeder reactors. The unit operations such as salt preparation, electrorefining, and cathode processing involve the presence of molten LiCl-KCl eutectic salt from 673 to 1373 K (400 to 1100 °C). The present work discusses the corrosion behavior of electroformed nickel (EF Ni) without and with nickel-tungsten (Ni-W) coating, 316L SS, and INCONEL 625 alloy in molten LiCl-KCl eutectic salt at 673 K, 773 K, and 873 K (400 °C, 500 °C, and 600 °C) in the presence of air. The weight percent loss of the exposed samples was determined by the weight loss method and surface morphology of the salt exposed, and product layers were examined by scanning electron microscopy (SEM). X-ray diffraction (XRD) and energy-dispersive X-ray (EDX) analysis were also carried out on the exposed and corrosion product layers to understand the phases present and the corrosion mechanism involved. The results of the present study indicated that INCONEL 625 alloy showed superior corrosion resistance compared to electroformed nickel (EF Ni), EF Ni with nickel-tungsten (Ni-W) coating (EF Ni-W), and 316L SS. The EF Ni with Ni-W coating exhibits better corrosion resistance than EF Ni without tungsten coating. Based on the surface morphology, XRD, and EDX analysis of corrosion product layers, the mechanism of corrosion of INCONEL 625 and 316L involves formation of chromium-rich compound at the surface and subsequent spallation. For the EF Ni, the porous thick NiO corrosion product allows the penetration of salt, thus accelerating the corrosion. Improved corrosion resistance of EF Ni-W was attributed to the W-rich NiO layer, while for INCONEL 625, the adherent and protective NiO layer improved the corrosion resistance. The article highlights the results of the present investigation.