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Mechanical and Thermal Problems of Water-Cooled Nuclear Power Reactors

Description: Some of the principal problems faced in the mechanical and thermal design of the Shippingport Pressorized Water Reactor core are discussed. The interplay of these problems with the requirements of other technologies is discussed, and areas which need more work are outlined. (M.H.R.)
Date: January 1, 1955
Creator: Palladino, N. J. & Sherman, J.
Partner: UNT Libraries Government Documents Department

Reactor Development Quarterly Progress Report

Description: The design and construction program of the Bolling Experimental Reactor is reviewed. A number of preliminary experinnents were performed with Borax-II at pressures between 75 and 300 psi. The most corrosion-resistant U-Zr--Nb alloy developed so far is produced by heating in a vacuum to the gamma phase, quenching, and aging for 2 hr at 400 deg C. Special attention is given to the removal of H/sub 2/ from the material. Unclad and unirradiated samples of U--Nb and U--Nb--Zr alloys were corrosion tested in H/sub 2/O. Corrosion rates were also measured under irradiation conditions in CP-5. Elongation measurements of irradiated wrought and cast U--Zr material suggested no way for treating the wrought fuel so that stability comparable to the cast material could be obtained. Natural circulation boiling density tests at 600 psia were made in order to determine the effects of channel cross section and subcooling on the steam void fraction. Results of autoclave and dynamic corrosion studies of 2-S Al in H/sub 2/O are reported. These results include the testing of Ni-clad samples. A large number of criticality calculations were performed for the EBR-Il and the PBR. The solubility of ThO/sub 2/ pellets containing various concentrations of U/sub 3/ O/sub 8/ was tested in water at 316 deg C for periods of 672 to 744 hr. None of the samples disintegrated, although at least one sample developed cracks. Solutions of reactor kinetic equations were attempted for the purpose of studying transients in reactors with lifetimes of 7 x 10/sup -9/ 10/sup -7/, and 6 x 10/ sup -5/ sec. Ignition experiments were performed on Th, Cu, Al, Fe, Mg, Zr, and fluorothene when contacted with fluorides. Except for Zr and fluorothene, the materials did not ignite. (C.H.)
Date: January 31, 1955
Partner: UNT Libraries Government Documents Department

PRELIMINARY DESIGN AND COST ESTIMATE FOR THE PRODUCTION OF CENTRAL STATION POWER FROM AN AQUEOUS HOMOGENEOUS REACTOR UTILIZING THORIUM-URANIUM-233

Description: The design and economics of the Aqueous Homogeneous Reactor as basically under development at the Oak Ridge National Laboratory are presented. The reactor system utilizes thorium-U-233 fuel. Conditions accompanying reactor systems generating up to l080 mw of net electrical energy are covered. The study indicates that a generating station, with a net thermal efficiency of 28.l%, might be constructed for approximately 0/kw and 0/kw at the l80 mw and l080 mw electrical levels, respectively. These values result in capital expenses of approximately 4.72 and 2.86 milis/kwh. A major part of fuel cost is the expense of chemical processing. It is therefore advantageous 10 schedule fuel through a relatively large processing system since fixed charges are insensitive to chemical plant size. By handling fuel through a plant large enough for processing 200 kg of thorium per day, total fuel costa of about 1 mill/kwh result. This cost for fuel processing appears applicable to generating stations up to abeut 540 mw in size, decreasing to about 0.6 mills/kwh at the l080 mw level. Operating and maintenance expense, including heavy water cost on a lease basis, varies between l.34 and 0.89 mills/kwh for l80 and l080 megawatts respectively. If the purchase of heavy water is required, 0.3 to 0.4 mills/kwh must be added. It is concluded that the Aqueous Homogeneous Reactor may produce electrical power competitive with conventional generating systems when the remaining technical problems are solved. It is felt ihat the research and development now programed by the Oak Ridge National Laboratory will solve these problems and affect costs favorably. (auth)
Date: February 1, 1955
Creator: Carson, H.G. & Landrum, L.H. eds.
Partner: UNT Libraries Government Documents Department

A study to determine the economical tank size for radioactive waste disposal

Description: Purpose of this report is to determine optimum tank size from evaluating the quantities of principal construction materials with prevailing unit costs for various tank sizes. The materials were concrete in-place, reinforcing steel in-place, wood framework, 3/8 in. C steel plate liner in-place, earthwork excavation and backfill (engineering, overhead, piping, condenser, vapor manifold costs not included). Costs of optimum tank are distributed as follows: dome 25%, walls 28%, foundation 6%, floor 2%, steel liner 20%, earthwork 19%. For a given tank capacity, there is a definite optimum tank size; as the capacity increases, the diameter increases, and the height increases but at a lesser rate. Each diameter has an optimum height, which is that height at which unit cost of storage space is minimum for a given tank diameter. Optimum unit cost is $0.136/gallons for diameters 75--130 ft; for diameters<75 ft, the optimum unit increases. Tank forms 241-S, 241-SX, and 241-A were used in this study; storage cost of the analyzed tank was $0.121 compared to $0.136/gallon for 241-SX and 241-A, and $0.152/gallon for 241-S. Assumed unit costs for concrete and steel plate tank liner were 10% less than those of 241-SX and 241-A, causing the lower unit costs. Tanks used for 241-SX and 241-A are close to optimum tank size. Although a small savings in storage unit cost will result from optimum dimensions, greater savings could be realized by increasing the tank capacity 20--25% without changing tank diameter or storage unit cost. Curves in the appendix can be used to select optimum tanks for new tank farms.
Date: February 11, 1955
Creator: Stivers, H. W.
Partner: UNT Libraries Government Documents Department

A Sodium-Graphite Reactor Steam-Electric Station for 75 Megawatts Net Generation

Description: The major design features, nuclear characteristics and performance data for a nuclear fueled central station power plant of 75,000 kw net capacity are presented. The heat source is a Na cooled graphite moderated reactor. The design of the reactor takes full advantage of the experience gained to date on the Sodium Reactor Experiment (SRE); the plant described here is a straightforward extension of the smaller experimental SRE, which is now under construction. The fuel elements are made up of rod clusters and the moderator is in the form of Zr canned graphite elements. The performance of the reactor has been based on conservative temperatures and coolant flow velocities which result in a plant with "built-in reserve." Thus, as experience is gained and anticipated improvements in reactor fuel elements and construction materials are proven, the performance of the plant can be increased accordingly. Two reactor designs are described, one for operation with slightly enriched U fuel elements and the other for operation with Th--U fuel elements. The associated heat exchangers, pumps, steam, and electrical generating equipment are identical for either reactor design. An analysis of turbine cycles describes the particular cycle chosen for initial operation and discusses a method by which modern central station performance can be initially obtained. The design and performance data which are required to enable reliable estimates of the plant construction and operating costs to be made are established. (auth)
Date: March 22, 1955
Creator: Weisner, E. F. & Sybert, W. M.
Partner: UNT Libraries Government Documents Department

HEAT TRANSFER ANALYSIS AND DESIGN OF A PLUGGING INDICATOR SYSTEM FOR SRE

Description: The analysis was performed on a system comprising a counterflow, concentric-pipe economizer, heat exchanger, flowmeter, plug, and connecting pipe. The system was assumed to be at some initial temperature equal to the inlet sodium temperature and suddenly loses heat to a medium in the heat exchanger. Design and operating data are presented. A cooling rate curve is given where the nitrogen flow rate is decreased when the plug temperature reaches 400 deg F. The time variation of minimum temperatures is given for various values of thermal capacitance with constant equilibrium temperature, and the economizer parameter with constant equilibrium temperatures and thermal capacitance. The variation in heat exchanger parameter with economizer parameter for a constant equilibrium minimum temperature of 250 deg F, and a constant inlet temperature of 750 deg F is indicated. (B.O.G.)
Date: April 1, 1955
Creator: Sletten, H. L.
Partner: UNT Libraries Government Documents Department

Subcontract Work by Thompson Products, Inc.

Description: In April, 1954, Thompson Products Inc. formally began work on a program leading to the fabrication of two test radiators. The work done to date is summarized in report number PWAC-127, First Technical Report by Thompson Products Inc. on Liquid-Metal-to-Air Radiators and in PWAC-128, Second Technical Report by Thompson Products Inc. on Liquid-Metal-to-Air Radiators. The first report describes possible radiator design concepts meeting the required specifications and presents an analytical method which was derived to optimize the designs. Thompson Products has decided that a core with cast NaK passages in combination with the sheet-metal, ribbon fins on the airside offers maximum reliability. The second report formulates the analysis of the final radiator design and presents the fabrication and evaluation studies used to select the design. To evaluate the permeability of cast NaK passages, Thompson Products is planning to build a hot NaK flow rig. A parallel investigation on wrought materials will be carried out as an alternate NaK passage type. The next phase of the program, the final design of the test radiator, has started.
Date: April 20, 1955
Creator: Shaw, R.H.
Partner: UNT Libraries Government Documents Department

DESIGN STUDY OF A NUCLEAR POWER PLANT FOR 100-KW ELECTRIC AND 400-KW HEAT CAPACITY

Description: A conceptional design study was made of a lowpower ''package'' reactor plant for the production of 100 kw of electrical power and 400 kw of heat at remote Arctic installations. The power plant steam generator is proposed to be an unmanned, heterogeneous, boilingtype reactor capable of continuous operation for extended periods. The design is based on data derived from experiments with boiling-type reactors conducted by Argorne at the Reactor Testing Station, Arco, Idaho.
Date: May 1, 1955
Creator: Treshow, M.; Snider, A.R. & Shaftman, D.H.
Partner: UNT Libraries Government Documents Department