11 Matching Results

Search Results

CORE LEVITATION IN THE EGCR IN CASE OF MAIN COOLANT PIPE FAILURE

Description: Results of an analysis to determine the extent of displacement of the EGCR core due to blowdown in case of several postulated hot main gas coolant pipe failures are summarized. Results show that the core will be damaged for ary hot pipe double-ended failure. Excepting the improbable case of no coolant flow existing prior to the break, the core will be damaged for any hot pipe fracture exposing a total flow area to the atmosphere equal to that of one pipe. Smaller breaks will probably be safe in this respect. (auth)
Date: August 4, 1959
Creator: Fontana, M.H.
Partner: UNT Libraries Government Documents Department

STATUS REPORT ON OXIDATION ANALYSES FOR THE EGCR

Description: Work done on the EGCR graphite combustion problem that may follow from the maximum credible accident, that is the rapid loss of pressure from the primaryreactor coolant system is summarized. The solution to date is to protect the fuel support sleeves with a siliconized siliconcarbide coating and to allow the moderator surfaces to oxidize. The moderator surfaces have available only 6.5% of the total core flow which places an upper limit on the rate of oxidation. The rate of heat removal through the sleeve to the main coolant flow is sufficient to cause a decrease in temperature throughout the reactor and subsequent quenching of the oxidation. This method depends only on continued coolant flow from one blower. Problems attendant with this and other schemes of controlling the fire are discussed. (auth)
Date: June 15, 1961
Creator: Fontana, M.H.
Partner: UNT Libraries Government Documents Department

Incentives and techniques for increasing the capacity of the geologic repository

Description: Estimates of the materials potentially destined for emplacement in Yucca Mountain exceed the statutory repository capacity limit of 70,000 metric tons initial heavy metal. Removal and subsequent burning of the actinides in these materials can dramatically increase the repository capacity, postponing or perhaps eliminating the need for a second repository. The detailed calculations described herein verify portions of a promising actinide removal and waste emplacement concept, HEWEC. Results from heat transfer calculations indicate that more than 2.5 times the material may be emplaced using a combination of optimum geometry and actinide recycle. This optimum geometry includes additional drifts and closer borehole spacing within the drifts. Future work will quantify the additional benefits that may be derived from drift ventilation and staggered emplacement strategies.
Date: September 1, 1994
Creator: Cowell, B.S.; Fontana, M.H. & Michaels, G.E.
Partner: UNT Libraries Government Documents Department

The HGCR-1, a Design Study of a Nuclear Power Station Employing a High-Temperature Gas-Cooled Reactor with Graphite-UO₂ Fuel Elements

Description: "The preliminary design of a 3095-Mw(thermal), helium-cooled, graphite- moderated reactor employing sign conditions, 1500 deg F reactor outlet gas would be circulated to eight steam generators to produce 1050 deg F, 1450-psi steam which would be converted to electrical power in eight 157-Mw(electrical) turbine- generators. The over-all efficiency of this nuclear power station is 36.5%. The significant activities released from the unclad graphite-UO₂ fuel appear to be less than 0.2% of those produced and would be equivalent to 0.002 curie/ cm/ sup 3/ in the primary helium circuit. The maintenance problems associated with this contamination level are discussed. A cost analysis indicates that the capital cost of this nuclear station per electrical kilowatt would be around 0, and that the production cost of electrical power would be 7.8 mills/kwhr" (p. iii).
Date: 1959
Creator: Cottrell, William B.; Copenhaver, C. M.; Culver, H. N.; Fontana, M. H.; Kelleghan, V. J. & Samuels, G.
Partner: UNT Libraries Government Documents Department

ACTIVITY RELEASE FROM THE N.S. SAVANNAH IN THE MAXIMUM CREDIBLE ACCIDENT

Description: The release of fission products that would occur following the maximum credible accident aboard the N.S. Savannah has been examined. Four significantly different, but realistic, operating histories were considered. The rate of release of noble gases and of iodine isotopes as a function of time after the accident was determined for each operating history and for both normal and emergency reactor-compartment ventilation systems. The influence of radioactive decay and of the time delay in release and transport of activity through the containment system was investigated. Most of the results are expressed in terms of activity release and resultant individual exposures, although some consideration is given to population exposures and to the interpretation of these results in the light of stationary reactor site criteria. (auth)
Date: October 16, 1963
Creator: Anderson, T.D.; Buchanan, J.R.; Cottrell, W.B.; Fontana, M.H.; Klepper, O.H. & McCurdy, H.C.
Partner: UNT Libraries Government Documents Department

THE HGCR-1, A DESIGN STUDY OF A NUCLEAR POWER STATION EMPLOYING A HIGH- TEMPERATURE GAS-COOLED REACTOR WITH GRAPHITE-UO$sub 2$ FUEL ELEMENTS

Description: The preliminary design of a 3095-Mw(thermal), helium-cooled, graphite- moderated reactor employing sign conditions, 1500 deg F reactor outlet gas would be circulated to eight steam generators to produce 1050 deg F, 1450-psi steam which would be converted to electrical power in eight 157-Mw(electrical) turbine- generators. The over-all efficiency of this nuclear power station is 36.5%. The significant activities released from the unclad graphite-UO/sub 2/ fuel appear to be less than 0.2% of those produced and would be equivalent to 0.002 curie/ cm/ sup 3/ in the primary helium circuit. The maintenance problems associated with this contamination level are discussed. A cost analysis indicates that the capital cost of this nuclear station per electrical kilowatt would be around 0, and that the production cost of electrical power would be 7.8 mills/kwhr. (auth)
Date: July 28, 1959
Creator: Cottrell, W.B.; Copenhaver, C.M.; Culver, H.N.; Fontana, M.H.; Kelleghan, V.J. & Samuels, G.
Partner: UNT Libraries Government Documents Department

Accelerator-based conversion (ABC) of weapons plutonium: Plant layout study and related design issues

Description: In preparation for and in support of a detailed R and D Plan for the Accelerator-Based Conversion (ABC) of weapons plutonium, an ABC Plant Layout Study was conducted at the level of a pre-conceptual engineering design. The plant layout is based on an adaptation of the Molten-Salt Breeder Reactor (MSBR) detailed conceptual design that was completed in the early 1070s. Although the ABC Plant Layout Study included the Accelerator Equipment as an essential element, the engineering assessment focused primarily on the Target; Primary System (blanket and all systems containing plutonium-bearing fuel salt); the Heat-Removal System (secondary-coolant-salt and supercritical-steam systems); Chemical Processing; Operation and Maintenance; Containment and Safety; and Instrumentation and Control systems. Although constrained primarily to a reflection of an accelerator-driven (subcritical) variant of MSBR system, unique features and added flexibilities of the ABC suggest improved or alternative approaches to each of the above-listed subsystems; these, along with the key technical issues in need of resolution through a detailed R&D plan for ABC are described on the bases of the ``strawman`` or ``point-of-departure`` plant layout that resulted from this study.
Date: April 1995
Creator: Cowell, B. S.; Fontana, M. H.; Krakowski, R. A.; Beard, C. A.; Buksa, J.J.; Davidson, J. W. et al.
Partner: UNT Libraries Government Documents Department