Sample records for decladding
-
Evaluation of Methods for Decladding LWR Fuel for a Pyroprocessing-Based Reprocessing Plant
1992-10-01
oAD-A275 326 ORN.rFM-1121o04 OAK RIDGE NATIONAL LABORATORY Evaluation of Methods for Decladding _LWR Fuel for a Pyroprocessing -Based Reprocessing...Dist. Category UC-526 EVALUATION OF METHODS FOR DECLADDING LWR FUEL FOR A PYROPROCESSING -BASED REPROCESSING PLANT W. D. Bond J. C. Mailen G. E...decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyroprocesses
-
Treatment of G1 Baskets at the CEA Marcoule Site - 12027
DOE Office of Scientific and Technical Information (OSTI.GOV)
Fourquet, Line; Boya, Didier
2012-07-01
In the dismantling program for the first-generation French reactors in accordance with the nonproliferation treaty, the CEA is in charge of cleanup and dismantling operations for the facilities at Marcoule, including the decladding units. The G1 decladding was built between 1955 and 1957 in order to de-clad spent fuel elements from the G1 plutonium-producing reactor and prepare them for dissolution. The facility was also used for interim storage of G1, G2 and G3 fuel dissolution baskets, which had been used during plant operation for transfer (from the decladding facility to the UP1 plant) and/or dissolution of spent fuel elements. Onemore » of the cleanup projects involves recovery of the baskets, which will be cut up, sorted, and conditioned in metal bins. The bins will be immobilized with cement grout, then transferred to the onsite solid waste conditioning facility (CDS) and to the repository operated by the French National Radioactive Waste Management Agency (ANDRA). The project is now in progress, after special safety permits were issued and measurement stations and dedicated tools were developed to handle all types of baskets (which differed according to their origin and use). The disposal of all the baskets is scheduled to last 2 years and will produce 55 metal waste bins. (authors)« less
-
Bharadwaj, Reshma; Sai, V V R; Thakare, Kamini; Dhawangale, Arvind; Kundu, Tapanendu; Titus, Susan; Verma, Pradeep Kumar; Mukherji, Soumyo
2011-03-15
A novel label-free technique for the detection of pathogens based on evanescent wave absorbance (EWA) changes at 280 nm from a U-bent optical fiber sensor is demonstrated. Bending a decladded fiber into a U-shaped structure enhances the penetration depth of evanescent waves and hence sensitivity of the probe. We show that the enhanced EWA response from such U-bent probes, caused by the inherent optical absorbance properties of bacterial cells or biomolecules specifically bound to the sensor surface, can be exploited for the detection of pathogens. A portable optical set-up with a UV light emitting diode, a spectrometer and U-bent fiber optic probe of 200 μm core diameter, 0.75 mm bend radius and effective probe length of 1cm demonstrated an ability to detect less than 1000 cfu/ml. Copyright © 2011. Published by Elsevier B.V.
-
Molten salt extraction of transuranic and reactive fission products from used uranium oxide fuel
Herrmann, Steven Douglas
2014-05-27
Used uranium oxide fuel is detoxified by extracting transuranic and reactive fission products into molten salt. By contacting declad and crushed used uranium oxide fuel with a molten halide salt containing a minor fraction of the respective uranium trihalide, transuranic and reactive fission products partition from the fuel to the molten salt phase, while uranium oxide and non-reactive, or noble metal, fission products remain in an insoluble solid phase. The salt is then separated from the fuel via draining and distillation. By this method, the bulk of the decay heat, fission poisoning capacity, and radiotoxicity are removed from the used fuel. The remaining radioactivity from the noble metal fission products in the detoxified fuel is primarily limited to soft beta emitters. The extracted transuranic and reactive fission products are amenable to existing technologies for group uranium/transuranic product recovery and fission product immobilization in engineered waste forms.
-
LARGE SCALE METHOD FOR THE PRODUCTION AND PURIFICATION OF CURIUM
Higgins, G.H.; Crane, W.W.T.
1959-05-19
A large-scale process for production and purification of Cm/sup 242/ is described. Aluminum slugs containing Am are irradiated and declad in a NaOH-- NaHO/sub 3/ solution at 85 to 100 deg C. The resulting slurry filtered and washed with NaOH, NH/sub 4/OH, and H/sub 2/O. Recovery of Cm from filtrate and washings is effected by an Fe(OH)/sub 3/ precipitation. The precipitates are then combined and dissolved ln HCl and refractory oxides centrifuged out. These oxides are then fused with Na/sub 2/CO/sub 3/ and dissolved in HCl. The solution is evaporated and LiCl solution added. The Cm, rare earths, and anionic impurities are adsorbed on a strong-base anfon exchange resin. Impurities are eluted with LiCl--HCl solution, rare earths and Cm are eluted by HCl. Other ion exchange steps further purify the Cm. The Cm is then precipitated as fluoride and used in this form or further purified and processed. (T.R.H.)
-
NASA Astrophysics Data System (ADS)
Halkare, Pallavi; Punjabi, Nirmal; Wangchuk, Jigme; Kondabagil, Kiran; Mukherji, Soumyo
2016-04-01
Hollow gold nanostructures (HGNS) have been used in variety of optical biosensors due to their inherent advantage of operating at near infra red (NIR) wavelength, large extinction coefficient and high dielectric sensitivity. The absorption wavelength of these nanostructures can be modulated by changing the ratio of hollow region to the core shell thickness. The aim of the present study is to incorporate the properties of HGNS, to develop LSPR based U-bent fiber optic sensor for detection of pathogens. The detection was carried out using an experimental set up consisting of a white light source, 200 μm diameter optical fiber having bend diameter of 1.6 mm +/- 0. 2 mm and a spectrometer. The HGNS were immobilized on the decladded portion of the fiber optic probe by chemisorptions. The effective plasmon penetration depth of the HGNS immobilized fiber optic sensor was approximated by using alternating layers of positively and negatively charged polyelectrolytes. The HGNS immobilized U-bent fiber optic sensor was used for detection of E.coli B40 strain using bacteriophage T4. The preliminary experiments were carried out with 104 cfu/ml of E.coli B40 and the change in absorbance obtained was approx. 0.042 +/- 0.0045 abs. units (n = 3). The response of this sensor was found to be better than spherical gold nanoparticle immobilized sensing platforms.
-
Graphite Waste Tank Cleanup and Decontamination under the Marcoule UP1 D and D Program - 13166
DOE Office of Scientific and Technical Information (OSTI.GOV)
Thomasset, Philippe; Chabeuf, Jean-Michel; Thiebaut, Valerie
2013-07-01
The UP1 plant in Marcoule reprocessed nearly 20,000 tons of used natural uranium gas cooled reactor fuel coming from the first generation of civil nuclear reactors in France. During more than 40 years, the decladding operations produced thousands of tons of processed waste, mainly magnesium and graphite fragments. In the absence of a French repository for the graphite waste, the graphite sludge content of the storage pits had to be retrieved and transferred into a newer and safer pit. After an extensive R and D program, the equipment and process necessary for retrieval operations were designed, built and tested. Themore » innovative process is mainly based on the use of two pumps (one to capture and the other one to transfer the sludge) working one after the other and a robotic arm mounted on a telescopic mast. A dedicated process was also set up for the removal of the biggest fragments. The retrieval of the most irradiating fragments was a challenge. Today, the first pit is totally empty and its stainless steel walls have been decontaminated using gels. In the second pit, the sludge retrieval and transfer operations have been almost completed. Most of the non-pumpable graphite fragments has been removed and transferred to a new storage pit. After more than 6 years of operations in sludge retrieval, a lot of experience was acquired from which important 'lessons learned' could be shared. (authors)« less
-
Establishment of the roadmap for chlorination process development for zirconium recovery and recycle
DOE Office of Scientific and Technical Information (OSTI.GOV)
Collins, E.D.; Del Cul, G.D.; Spencer, B.B.
Process development studies are being conducted to recover, purify, and reuse the zirconium (about 98.5% by mass) in used nuclear fuel (UNF) zirconium alloy cladding. Feasibility studies began in FY 2010 using empty cladding hulls that were left after fuel dissolution or after oxidation to a finely divided oxide powder (voloxidation). In FY 2012, two industrial teams (AREVA and Shaw-Westinghouse) were contracted by the Department of Energy Office of Nuclear Energy (NE) to provide technical assistance to the project. In FY 2013, the NE Fuel Cycle Research and Development Program requested development of a technology development roadmap to guide futuremore » work. The first step in the roadmap development was to assess the starting point, that is, the current state of the technology and the end goal. Based on previous test results, future work is to be focused on first using chlorine as the chlorinating agent and secondly on the use of a process design that utilizes a chlorination reactor and dual ZrCl{sub 4} product salt condensers. The likely need for a secondary purification step was recognized. Completion of feasibility testing required an experiment on the chemical decladding flowsheet option. This was done during April 2013. The roadmap for process development will continue through process chemistry optimization studies, the chlorinated reactor design configuration, product salt condensers, and the off-gas trapping of tritium or other volatile fission products from the off-gas stream. (authors)« less
-
Top Ten Reasons for DEOX as a Front End to Pyroprocessing
DOE Office of Scientific and Technical Information (OSTI.GOV)
B.R. Westphal; K.J. Bateman; S.D. Herrmann
A front end step is being considered to augment chopping during the treatment of spent oxide fuel by pyroprocessing. The front end step, termed DEOX for its emphasis on decladding via oxidation, employs high temperatures to promote the oxidation of UO2 to U3O8 via an oxygen carrier gas. During oxidation, the spent fuel experiences a 30% increase in lattice structure volume resulting in the separation of fuel from cladding with a reduced particle size. A potential added benefit of DEOX is the removal of fission products, either via direct release from the broken fuel structure or via oxidation and volatilizationmore » by the high temperature process. Fuel element chopping is the baseline operation to prepare spent oxide fuel for an electrolytic reduction step. Typical chopping lengths range from 1 to 5 mm for both individual elements and entire assemblies. During electrolytic reduction, uranium oxide is reduced to metallic uranium via a lithium molten salt. An electrorefining step is then performed to separate a majority of the fission products from the recoverable uranium. Although DEOX is based on a low temperature oxidation cycle near 500oC, additional conditions have been tested to distinguish their effects on the process.[1] Both oxygen and air have been utilized during the oxidation portion followed by vacuum conditions to temperatures as high as 1200oC. In addition, the effects of cladding on fission product removal have also been investigated with released fuel to temperatures greater than 500oC.« less
-
DOE Office of Scientific and Technical Information (OSTI.GOV)
Jantzen, Carol M.; Lee, William E.; Ojovan, Michael I.
The main immobilization technologies that are available commercially and have been demonstrated to be viable are cementation, bituminization, and vitrification. Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. Most of the nations that have generated HLW are immobilizing in either alkali borosilicate glass or alkali aluminophosphate glass. The exact compositions of nuclear waste glasses are tailored for easy preparation and melting, avoidance of glass-in-glass phase separation, avoidance of uncontrolled crystallization, and acceptable chemical durability, e.g., leach resistance. Glass has also been used to stabilize a variety of lowmore » level wastes (LLW) and mixed (radioactive and hazardous) low level wastes (MLLW) from other sources such as fuel rod cladding/decladding processes, chemical separations, radioactive sources, radioactive mill tailings, contaminated soils, medical research applications, and other commercial processes. The sources of radioactive waste generation are captured in other chapters in this book regarding the individual practices in various countries (legacy wastes, currently generated wastes, and future waste generation). Future waste generation is primarily driven by interest in sources of clean energy and this has led to an increased interest in advanced nuclear power production. The development of advanced wasteforms is a necessary component of the new nuclear power plant (NPP) flowsheets. Therefore, advanced nuclear wasteforms are being designed for robust disposal strategies. A brief summary is given of existing and advanced wasteforms: glass, glass-ceramics, glass composite materials (GCM’s), and crystalline ceramic (mineral) wasteforms that chemically incorporate radionuclides and hazardous species atomically in their structure. Cementitious, geopolymer, bitumen, and other encapsulant wasteforms and composites that atomically bond and encapsulate wastes are also discussed. The various processing technologies are cross-referenced to the various types of wasteforms since often a particular type of wasteform can be made by a variety of different processing technologies.« less
-
The MaNGA integral field unit fiber feed system for the Sloan 2.5 m telescope
DOE Office of Scientific and Technical Information (OSTI.GOV)
Drory, N.; MacDonald, N.; Byler, N.
2015-02-01
We describe the design, manufacture, and performance of bare-fiber integral field units (IFUs) for the SDSS-IV survey Mapping Nearby Galaxies at Apache Point Observatory (MaNGA) on the the Sloan 2.5 m telescope at Apache Point Observatory. MaNGA is a luminosity-selected integral-field spectroscopic survey of 10{sup 4} local galaxies covering 360–1030 nm at R∼2200. The IFUs have hexagonal dense packing of fibers with packing regularity of 3 μm (rms), and throughput of 96 ± 0.5% from 350 nm to 1 μm in the lab. Their sizes range from 19 to 127 fibers (3–7 hexagonal layers) using Polymicro FBP 120:132:150 μm core:clad:buffermore » fibers to reach a fill fraction of 56%. High throughput (and low focal-ratio degradation (FRD)) is achieved by maintaining the fiber cladding and buffer intact, ensuring excellent surface polish, and applying a multi-layer anti-reflection (AR) coating of the input and output surfaces. In operations on-sky, the IFUs show only an additional 2.3% FRD-related variability in throughput despite repeated mechanical stressing during plate plugging (however other losses are present). The IFUs achieve on-sky throughput 5% above the single-fiber feeds used in SDSS-III/BOSS, attributable to equivalent performance compared to single fibers and additional gains from the AR coating. The manufacturing process is geared toward mass-production of high-multiplex systems. The low-stress process involves a precision ferrule with a hexagonal inner shape designed to lead inserted fibers to settle in a dense hexagonal pattern. The ferrule ID is tapered at progressively shallower angles toward its tip and the final 2 mm are straight and only a few microns larger than necessary to hold the desired number of fibers. Our IFU manufacturing process scales easily to accommodate other fiber sizes and can produce IFUs with substantially larger fiber counts. To assure quality, automated testing in a simple and inexpensive system enables complete characterization of throughput and fiber metrology. Future applications include larger IFUs, higher fill factors with stripped buffer, de-cladding, and lenslet coupling.« less
-
The MaNGA Integral Field Unit Fiber Feed System for the Sloan 2.5 m Telescope
NASA Astrophysics Data System (ADS)
Drory, N.; MacDonald, N.; Bershady, M. A.; Bundy, K.; Gunn, J.; Law, D. R.; Smith, M.; Stoll, R.; Tremonti, C. A.; Wake, D. A.; Yan, R.; Weijmans, A. M.; Byler, N.; Cherinka, B.; Cope, F.; Eigenbrot, A.; Harding, P.; Holder, D.; Huehnerhoff, J.; Jaehnig, K.; Jansen, T. C.; Klaene, M.; Paat, A. M.; Percival, J.; Sayres, C.
2015-02-01
We describe the design, manufacture, and performance of bare-fiber integral field units (IFUs) for the SDSS-IV survey Mapping Nearby Galaxies at Apache Point Observatory (MaNGA) on the the Sloan 2.5 m telescope at Apache Point Observatory. MaNGA is a luminosity-selected integral-field spectroscopic survey of 104 local galaxies covering 360-1030 nm at R˜ 2200. The IFUs have hexagonal dense packing of fibers with packing regularity of 3 μm (rms), and throughput of 96 ± 0.5% from 350 nm to 1 μm in the lab. Their sizes range from 19 to 127 fibers (3-7 hexagonal layers) using Polymicro FBP 120:132:150 μm core:clad:buffer fibers to reach a fill fraction of 56%. High throughput (and low focal-ratio degradation (FRD)) is achieved by maintaining the fiber cladding and buffer intact, ensuring excellent surface polish, and applying a multi-layer anti-reflection (AR) coating of the input and output surfaces. In operations on-sky, the IFUs show only an additional 2.3% FRD-related variability in throughput despite repeated mechanical stressing during plate plugging (however other losses are present). The IFUs achieve on-sky throughput 5% above the single-fiber feeds used in SDSS-III/BOSS, attributable to equivalent performance compared to single fibers and additional gains from the AR coating. The manufacturing process is geared toward mass-production of high-multiplex systems. The low-stress process involves a precision ferrule with a hexagonal inner shape designed to lead inserted fibers to settle in a dense hexagonal pattern. The ferrule ID is tapered at progressively shallower angles toward its tip and the final 2 mm are straight and only a few microns larger than necessary to hold the desired number of fibers. Our IFU manufacturing process scales easily to accommodate other fiber sizes and can produce IFUs with substantially larger fiber counts. To assure quality, automated testing in a simple and inexpensive system enables complete characterization of throughput and fiber metrology. Future applications include larger IFUs, higher fill factors with stripped buffer, de-cladding, and lenslet coupling.