60 research outputs found
Internal tides and tidal cycles of vertical mixing in western Long Island Sound
In estuaries, tidal period variations in the rate of vertical mixing have been observed to result from various causes: in Liverpool Bay and the York River, they have been attributed to tidal straining of the along-channel density gradient modulating stratification; in the Hudson River they arise from tidal modulation of the height of the tidal current bottom boundary layer (BBL). Along continental shelves, tidal period fluctuations in mixing have been observed to result from the dissipation of internal waves (IWs). Western Long Island Sound (WLIS) moored instrument records indicate that large near-bottom increases in dissolved oxygen (DO) and heat and a decrease in salt occur during the middle of the flood tide: an analysis of water mass signatures indicates that the transport involved is vertical and not horizontal. Temperature data from a vertical thermistor array deployed in the WLIS for 16 days in August 2009 clearly show a tidal cycle of IW activity creating a mean thermocline depression at midflood of approximately 25% of the water depth with individual IW thermocline depressions of as much as 50% of the water depth. Contemporaneous ADCP measurements show increases in shear due to IWs during the flood. Near-bottom internal wave activity is maximal at and after midflood and is correlated with near-bottom temperature and DO tendencies at both tidal and subtidal scales. We conclude that internal tides are an important vertical mixing mechanism in the WLIS through both increased shear from IWs and displacement of the pycnocline into the region of high shear in the BBL
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REACTIVITY ACCIDENT TEST RESULTS AND ANALYSES FOR THE SPERT III E-CORE: A SMALL, OXIDE-FUELED, PRESSURIZED-WATER REACTOR.
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Report of the SPERT I Destructive Test Program on an Aluminum, Plate-Type, Water-Moderated Reactor
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Results of the first nuclear blowdown test on single fuel rods (LOC-11 Series in PBF)
This paper presents results of the first nuclear blowdown tests (LOC-11A, LOC-11B, LOC-11C) ever conducted. The Loss-of-Coolant Accident (LOCA) Test Series is being conducted in the Power Burst Facility (PBF) reactor at the Idaho National Engineering Laboratory, near Idaho Falls, Idaho, for the Nuclear Regulatory Commission. The objective of the LOC-11 tests was to obtain data on the behavior of pressurized and unpressurized rods when exposed to a blowdown similar to that expected in a pressurized water reactor (PWR) during a hypothesized double-ended cold-leg break. The data are being used for the development and verification of analytical models that are used to predict coolant and fuel rod pressure during a LOCA in a PWR
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Fuel Summary Report: Shippingport Light Water Breeder Reactor
The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749
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Film boiling behavior in a nine rod cluster
Film boiling in a cluster geometry was investigated during a recent test conducted in the Power Burst Facility. Cluster information is necessary to assess the applicability of single rod experimental data used in developing light water reactor fuel rod behavior models. The nine-rod test, part of the Power-Cooling-Mismatch (PCM) Test Series, was designed to investigate the film boiling behavior of a central fuel rod surrounded by fuel rods that were also in the film boiling heat transfer regime. Information obtained provides insight into film boiling fuel rod behavior, in an environment believed to be representative of a power reactor rod during a period of over-power or low-flow operation. The nine test rods were arranged in a 3 x 3 lattice with spacing typical of a PWR cluster. Except for fuel enrichment and overall length, the rods were of PWR design
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