75 Matching Results

Search Results

Advanced search parameters have been applied.

Remotely operated plasma torch: a tool for nuclear reactor dismantling

Description: From winter meeting of American Society of Mechanical Engineers; Detroit, Michigan, USA (11 Nov 1973). The Elk River Reactor facility is being dismantled in order to return the site to unrestricted use. The highly radioactive components -- the reactor internals, the pressure vessel, and the outer thermal shield -- have been cut up and shipped to a burial ground. Approximately 10,000 Ci of radioactive metal was removed without significant release of activity and without any overexposure to personnel. The dismantling was accomplished with a remotely operated plasma torch system. The design of the system, the results obtained, and evaluation of this technology are given. (6 references) (auth)
Date: January 1, 1973
Creator: Beckers, R.M.; Blumberg, R. & Wodtke, C.H.
Partner: UNT Libraries Government Documents Department

Behavior of irradiated LWR fuel pellets during thermal transients

Description: Prediction of the behavior of LWR fuel rods and fission products under off-normal and accident conditions requires a physically realistic description of fuel swelling and fission-product release that currently does not exist. To satisfy this need, a program was initiated at ANL approximately a year ago with the prime objective of developing a comprehensive computer-base model that describes the release of fission products as a function of thermal transients anticipated in hypothetical accident situations. This model will be incorporated into ANC's FRAP accident-analysis code system. The analytical effort is supported by data developed from characterization of irradiated LWR fuel and from out-of-reactor transient heating tests of irradiated LWR fuel under conditions that simulate hypothetical LWR accidents. (auth)
Date: January 1, 1975
Creator: Kelman, L.R.; Rest, J.; Seitz, M.G. & Gehl, S.M.
Partner: UNT Libraries Government Documents Department

Heavy Section Steel Technology Program. Part II. Intermediate vessel testing

Description: The testing of the intermediate pressure vessels is a major activity under the Heavy Section Steel Technology Program. A primary objective of these tests is to develop or verify methods of fracture prediction, through the testing of selected structures and materials, in order that a valid basis can be established for evaluating the serviceability and safety of light-water reactor pressure vessels. These vessel tests were planned with sufficiently specific objectives that substantial quantitative weight could be given to the results. Each set of testing conditions was chosen so as to provide specific data by which analytical methods of predicting flaw growth, and in some cases crack arrest, could be evaluated. Every practical effort was made to assure that results would be relevant to some aspect of real reactor pressure vessel performance through careful control of material properties, selection of test temperatures, and design of prepared flaws. 5 references (auth)
Date: January 1, 1975
Creator: Whitman, G.D.
Partner: UNT Libraries Government Documents Department

Core thermal model development and experiments. Progress report, 1 July 1974--31 December 1974

Description: Analytical and experimental studies are being performed at the Pacific Northwest Laboratory to improve methods of thermal-hydraulic analysis for reactor cores. The work is primarily concerned with developing computational methods that can consider multidimensional two-phase hydrodynamic phenomena in reactor cores during postulated accidents. The basic approach being taken in the analytical part of this program is to adapt the subchannel analysis concept to analysis of a nuclear core. This is being done by representing the core as a matrix of control volumes. Transport equations are written for two-phase flow, and are then solved by a digital computer. This provides a multidimensional representation of the core from which it may be possible to define simpler analysis methods: i.e., define the minimum amount of detail required for core thermal analysis. The experimental part of this program is concerned with validating the computational model. Experiments are presently concerned with flow distributions in rod bundles containing flow blockages and with the evaluation of laser-Doppler techniques to measure two-phase hydrodynamic phenomena. Major efforts have been made to improve the computational capability of the COBRA code for reactor safety analysis; compare the code with experimental data; evaluate the code for core analysis; and experimentally measure velocity in rod bundles containing flow blockage. (auth)
Date: unknown
Creator: Rowe, D.S.
Partner: UNT Libraries Government Documents Department

Aerial radiological measuring survey of the area surrounding the La Crosse Boiling Water Reactor, Genoa, Wisconsin, July 1968

Description: The Aerial Radiological Measuring System (ARMS) was used to survey the area surrounding the La Crosse Boiling Water Reactor during July 1968. The survey measured terrestrial gamma radiation. A high-sensitivity detection system collected gamma spectral and gross-count data. The data were then computer processed into a map of a 625 sq. mile area showing isoexposure contours 3 ft above the ground. Exposure rates and isotopes identiiied are consistent with normal background radiation. (auth)
Date: October 1, 1973
Partner: UNT Libraries Government Documents Department

Aerial radiological measuring survey of the area surrounding the Monticello Nuclear Generating Plant, Monticello, Minnesota, August 1970

Description: The Aerial Radiological Measuring System (ARMS) was used to survey the Monticello Nuclear Generating Plant and surrounding area during Aug. 1970. The survey measured terrestrial background gamma radiation. A high-sensitivity detection system collected gamma spectral and gross-count data. The data were then computer processed into a map of a 625 sq. mile area showing isoexposure contours 3 ft above the ground. Results indicated the presence of isotopes normally found in the background radiation throughout the United States. (auth)
Date: September 1, 1973
Partner: UNT Libraries Government Documents Department

Aerial radiological measuring survey of the area surrounding the Vermont Yankee Generating Station and the Yankee Nuclear Power Station, September 18, 1970

Description: The Aerial Radiological Measuring System (ARMS) was used to survey the area surrounding the Vermont Yankee Generating Station and the Yankee Nuclear Power Station during September 1970. The survey measured terrestrial gamma radiation. A high-sensitivity detection system collected spectral and grosscount data for the flyable portions of the 625 sq. mile survey area. Isotopes identified and exposure rates are consistent with normal background radiation. (auth)
Date: October 1, 1973
Partner: UNT Libraries Government Documents Department

Use of vertical slip flow and flooding models in LOCA analysis

Description: Vertical slip flow and flooding models, which have been incorporated in a version of the RELAP4 computer code by Aerojet Nuclear Company have led to significant improvements in modeling nuclear reactor coolant system phenomena during postulated large and small break loss-of-coolant accidents. The vertical slip flow model computes the separated fluid component velocities and directions at vertical flow junctions. Use of the slip model allows the energy transfer between volumes to be based on individual liquid and vapor component flows rather than on the net junction flow. Continuity and momentum equations are unaffected by the addition of slip. The vertical flow slip model logic is based on the assumption that gravity forces dominate causing slip between phases. 7 references (auth)
Date: January 1, 1975
Creator: Fischer, S.R.
Partner: UNT Libraries Government Documents Department