Relevant bibliographies by topics / Argonne National Laboratory-West

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Academic literature on the topic 'Argonne National Laboratory-West'

Author: Grafiati
Published: 4 June 2021
Last updated: 2 February 2022

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Journal articles on the topic "Argonne National Laboratory-West"

1

Mariani, Robert D., Robert W. Benedict, Richard M. Lell, Ronald B. Turski, and Edward K. Fujita. "Criticality Safety Strategy and Analysis Summary for the Fuel Cycle Facility Electrorefiner at Argonne National Laboratory West." Nuclear Technology 114, no. 2 (May 1996): 224–34. http://dx.doi.org/10.13182/nt96-a35251.

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2

Esh, D. W., and R. W. Benedict. "Thermal/Hydrological Modeling of the Radioactive Scrap and Waste Facility (RSWF) with the Tough2 (Transport of Unsaturated Groundwater and Heat) Code." MRS Proceedings 465 (1996). http://dx.doi.org/10.1557/proc-465-1109.

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ABSTRACTThermal/hydrological modeling of the Radioactive Scrap and Waste Facility (RSWF) has been completed with the TOUGH2 (Transport of Unsaturated Groundwater and Heat) Code.[1] The RSWF will be utilized as an interim storage facility for ceramic and metallic waste forms developed from the electrometallurgical treatment of spent nuclear fuel. The RSWF is an array of 1,350 carbon steel liners located at grade near Argonne National Laboratory-West on the Idaho National Engineering Laboratory (INEL).The primary driving force for this modeling research was to assess thermal capacity limits for RSWF liners so that heat generating materials can be safely stored. Maximum wasteform temperatures will be governed by both the amount of heat the system can dissipate and the orientation and characteristics of the wasteform. The focus of this report is on the amount of heat the interim storage system can safely dissipate. The effect of the temporal variation of soil moisture on the performance of the RSWF is assessed. The facility was analyzed to determine the maximum allowable thermal loading of the RSWF liners.
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Books on the topic "Argonne National Laboratory-West"

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U.S. Dept. of Energy. Report of investigation into allegations of retaliation for raising safety and quality of work issues regarding Argonne National Laboratory's integral fast reactor project. Washington, DC: U.S. Dept. of Energy, Office of Nuclear Safety, 1991.

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2

U.S. Dept. of Energy. Report of investigation into allegations of retaliation for raising safety and quality of work issues regarding Argonne National Laboratory's integral fast reactor project. Washington, DC: U.S. Dept. of Energy, Office of Nuclear Safety, 1991.

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3

1933-, Lanphere Marvin A., and Geological Survey (U.S.), eds. Age and paleomagnetism of basaltic lava flows in corehole ANL-OBS-AQ-014 at Argonne National Laboratory-West, Idaho National Engineering and Environmental Laboratory. [Menlo Park, CA]: U.S. Dept. of the Interior, U.S. Geological Survey, 1997.

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Conference papers on the topic "Argonne National Laboratory-West"

1

McDermott, Mary D., Charles D. Griffin, Daniel K. Baird, Carl E. Baily, John A. Michelbacher, Kenneth E. Rosenberg, and S. Paul Henslee. "Completion of Experimental Breeder Reactor-II Sodium Processing at Argonne National Laboratory." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22485.

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The Experimental Breeder Reactor - II (EBR-II) at Argonne National Laboratory - West (ANL-W) was shutdown in September 1994 as mandated by the United States Department of Energy. Located in eastern Idaho, this sodium-cooled reactor had been in service since 1964, and was a test facility for fuels development, materials irradiation, system and control theory tests, and hardware development. The EBR-II termination activities began in October 1994, with the reactor being maintained in an industrially and radiologically safe condition for decommissioning. With the shutdown of EBR-II, its sodium coolant became a waste necessitating its reaction to a disposal form. A Sodium Process Facility (SPF), designed to convert sodium to 50 wt% sodium hydroxide, existed at the ANL-W site, but had never been operated. The SPF was upgraded to current standards and codes, and then modified in 1998 to convert the sodium to 70 wt% sodium hydroxide, a substance that solidifies at 65°C (150°F) and is acceptable for burial as low level radioactive waste in Idaho. In December 1998, the SPF began operations. Working with sodium and highly concentrated sodium hydroxide presented some unique operating and maintenance conditions. Several lessons were learned throughout the operating period. Processing of the 330 m3 (87,000 gallons) of EBR-II primary sodium, 50 m3 (13,000 gallons) of EBR-II secondary sodium, and 290 m3 (77,000 gallons) of Fermi-1 primary sodium was successfully completed in March 2001, ahead of schedule and within budget.
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2

Einziger, R. E., H. C. Tsai, M. C. Billone, and B. A. Hilton. "Examination of Spent PWR Fuel Rods After 15 Years in Dry Storage." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22456.

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Virginia Power Surry Nuclear Station Pressurized Water Reactor (PWR) fuel was stored in a dry inert atmosphere Castor V/21 cask at the Idaho National Environmental and Engineering Laboratory (INEEL) for 15 years at peak cladding temperatures decreasing from about 350 to 150°C. Prior to the storage, the loaded cask was subjected to extensive thermal benchmark tests. The cask was opened to examine the fuel for degradation and to determine if it was suitable for extended storage. No rod breaches had occurred and no visible degradation or crud/oxide spallation were observed. Twelve rods were removed from the center of the T11 assembly and shipped from INEEL to the Argonne-West HFEF for profilometric scans. Four of these rods were punctured to determine the fission gas release from the fuel matrix and internal pressure in the rods. Three of the four rods were cut into five segments each, then shipped to the Argonne-East AGHCF for detailed examination. The test plan calls for metallographic examination of six samples from two of the rods, microhardness and hydrogen content measurements at or near the six metallographic sample locations, tensile testing of six samples from the two rods, and thermal creep testing of eight samples from the two rods to determine the extent of residual creep life. The results from the profilometry (12 rods), gas release measurements (4 rods), metallographic examinations (2 samples from 1 rod), and microhardness and hydrogen content characterization (2 samples from 1 rod) are reported here. The tensile and creep studies are just starting and will be reported at a later date, along with the additional characterization work to be performed. Although only limited prestorage characterization is available, a number of preliminary conclusions can be drawn based on comparison with characterization of Florida Power Turkey Point rods of a similar vintage. Based on this comparison, it appears that little or no cladding thermal creep and fission gas release from the fuel pellets occurred during the thermal benchmark tests or storage. Measurements of the cladding outer-diameter, oxide thickness and wall thickness are in the expected range for cladding of the Surry exposure. The measured hydrogen content is consistent with the oxide thickness. The volume of hydrides varies azimuthally around the cladding, but there is little variation across the thickness, of the cladding. It is most significant that all of the hydrides appear to have retained the circumferential orientation typical of prestorage PWR fuel rods.
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3

De Luca, Domenico, Alessandro Petruzzi, Marco Cherubini, and Valeria Parrinello. "RELAP5-3D Analysis of EBR-II Shutdown Heat Removal Test SHRT-17." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60629.

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Coordinated Research Project (CRP) on EBR-II Shutdown Heat Removal Tests (SHRT) was established by International Atomic Energy Agency (IAEA). The objective of the project is to support and to improve validation of simulation tools and projects for Sodium-cooled Fast Reactors (SFR). The Experimental Breeder Reactor II (EBR-II) plant was a uranium metal-alloy-fuelled liquid-metal-cooled fast reactor designed and operated by Argonne National Laboratory (ANL) for the U.S. Department of Energy at the Argonne-West site. In the frame of this project, benchmark analysis of one of the EBR-II shutdown heat removal tests, protected loss-of-flow transient (SHRT-17), has been performed. The aim of this paper is to present modeling of EBR-II reactor design using RELAP-3D, to show the results of the transient analysis of SHRT-17, and to discuss the results of application of the Fast Fourier Transform Based Method (FFTBM) to perform a quantitative accuracy evaluation of the model developed. Complete nodalization of the reactor was made from the beginning. Model is divided in primary side that contains core, pumps, reactor pool and, for this kind of reactor specific, Z pipe, and intermediate side that contains Intermediate Heat Exchanger (IHX). After achievement of acceptable steady-state results, transient analysis was performed. Starting from full power and flow, both the primary loop and intermediate loop coolant pumps were simultaneously tripped and the reactor was scrammed to simulate a protected loss-of-flow accident. In addition, the primary system auxiliary coolant pump, that normally had an emergency battery power supply, was turned off. Despite early rise of the temperature in the reactor, the natural circulation characteristics managed to keep it at acceptable levels and cooled the reactor down safely at decay heat power levels. Thermal-hydraulics characteristics and plant behavior was focused on prediction of natural convection cooling by evaluating the reactor core flow and temperatures and their comparison with experimental data that were provided by ANL. Finally, the process of qualification of a system thermal-hydraulic code calculation was applied. It consists of three steps: 1) the geometrical fidelity of the nodalization, related with the evaluation and comparison of the geometrical data of the hardware respect to the estimated numerical values implemented in the nodalization; 2) the steady state level qualification, dealing with the capability of the nodalization to reproduce the steady state qualified conditions of the system; 3) the “on-transient” qualification, necessary to demonstrate the capability of the code and of the developed nodalization to reproduce the relevant thermal-hydraulic phenomena expected during the transient. The latter is a very complex step which foreseen different phases following our methodology of qualification (SCCRED, Standardized and Consolidated Calculated & Reference Experimental Database methodology). In the framework of the benchmark, the focus was only on the so called “Quantitative Accuracy Evaluation” that is performed by the FFTBM.
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4

Michelbacher, John A., Carl E. Baily, Daniel K. Baird, S. Paul Henslee, Collin J. Knight, and Kenneth E. Rosenberg. "Shutdown and Closure of the Experimental Breeder Reactor–II." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22462.

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The Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to maintain the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The EBR-II is a pool-type reactor. The primary system contained approximately 325 m3 (86,000 gallons) of sodium and the secondary system contained 50 m3 (13,000 gallons). In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility was built to react the sodium to a solid sodium hydroxide monolith for burial as a low level waste in a land disposal facility. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in circuits and components must be passivated, inerted, or removed to preclude future concerns with sodium-air reactions that could generate potentially explosive mixtures of hydrogen and leave corrosive compounds. The passivation process being implemented utilizes a moist carbon dioxide gas that generates a passive layer of sodium carbonate/sodium bicarbonate over any quantities of residual sodium. Tests being conducted will determine the maximum depths of sodium that can be reacted using this method, defining the amount that must be dealt with later to achieve RCRA clean closure. Deactivation of the EBR-II complex is on schedule for a March, 2002, completion. Each system associated with EBR-II has an associated layup plan defining the system end state, as well as instructions for achieving the layup condition. A goal of system-by-system layup is to minimize surveillance and maintenance requirements during the interim period between deactivation and decommissioning. The plans also establish document archival of not only all the closure documents, but also the key plant documents (P&IDs, design bases, characterization data, etc.) in a convenient location to assure the appropriate knowledge base is available for decommissioning, which could occur decades in the future.
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Reports on the topic "Argonne National Laboratory-West"

1

Jensen, B. A., J. Sanders, T. Wenz, and R. Buchheit. Results of active well coincidence counter cross-calibration measurements at Argonne National Laboratory-West. Office of Scientific and Technical Information (OSTI), October 2002. http://dx.doi.org/10.2172/805262.

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2

Goff, K. M., R. D. Mariani, and N. L. Bonomo. Depleted uranium start-up of spent fuel treatment operation at Argonne National Laboratory-West. Office of Scientific and Technical Information (OSTI), December 1995. http://dx.doi.org/10.2172/202337.

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Author, Not Given. Enhancement of Argonne National Laboratory -- West (ANL-W) to Examine LWR Fuel and Control Rods. Office of Scientific and Technical Information (OSTI), August 2006. http://dx.doi.org/10.2172/942160.

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4

Routine environmental reaudit of the Argonne National Laboratory - West. Office of Scientific and Technical Information (OSTI), April 1996. http://dx.doi.org/10.2172/216273.

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