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Estimate Of Potential Fuel Reprocessing, Revision #30 Part A

Description: The power and estimated reprocessing load are tabulated for existing and proposed for United States and United States-built reactors of 10 Kw or greater thermal power.
Date: March 15, 1960
Creator: Ullmann, J. W.
Partner: UNT Libraries Government Documents Department
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High Flux Isotope Reactor - A General Description

Description: This report is issued primarily to make available a general description which reflects the current status of the High Flux Isotope Reactor which is being planned for construction at Oak Ridge National Laboratory.
Date: March 15, 1960
Creator: Cole, T. E.
Partner: UNT Libraries Government Documents Department
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VARI Solution of Simultaneous, First-Order, Ordinary, Differential Equations

Description: VARI solves on the IBM-650 a system of simultaneous, first-order, ordinary, differential equations. The program was written so that a large number of calculations could be done in a reasonable length of time. The program permits the consideration of the production of the isotope by absorption and/or decay of one or more of any of the other isotopes in the chain.
Date: March 15, 1960
Creator: Kerr, B. A.
Partner: UNT Libraries Government Documents Department
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Problems of a Small Leak Between the Flow Monitor and Heated Section of a PRTR Process Tube

Description: The result of a leak in a PRTR process tube between the flow monitor and the heated section would be to increase the flow through the monitor, but to decrease the flow through the heated section. The concern for the case of small leaks is whether the increase in flow through the flow monitor is sufficient to cause a high flow tip and a reactor scram for the condition where the flow through the heated section is reduced to the point to cause excessive fuel element temperatures.
Date: March 15, 1960
Creator: Hesson, G. M.
Partner: UNT Libraries Government Documents Department
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Unique Fabrication Processes Applied to Fuel Cladding Materials

Description: The fabrication processes applied to nuclear fuels are subject to severe limitations because of the conditions imposed by the reactor environment. The combined problems of neutrons fluxes, high heat fluxes, corrosion by the coolant , and embrittlement by hydriding or similar reactions may be minimized through establishing rigorous materials and fabrication specifications for fuel and cladding.
Date: March 15, 1960
Creator: Bush, S. H.
Partner: UNT Libraries Government Documents Department
open access

Zero Power Experiments for the SM-1 Core II : Task XV

Description: Abstract: An element by element reactivity check for SM-1 Core II fuel elements and control rod absorber sections was performed and the burnable nuclear poison loading in the SM-1 Core II stationary fuel elements was established. An approach to criticality of the SM-1 Core II was performed by the inverse multiplication method and the critical rod bank position obtained as a function of fuel loading up to the full SM-1 Core II loading. Maximum and minimum core reactivity measurements were obtai… more
Date: March 15, 1960
Creator: Robinson, R. A.; Weiss, S. H.; McCool, W. J. & Schrader, E. W.
Partner: UNT Libraries Government Documents Department
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Hanford Laboratories Operation monthly activities report, February 1960

Description: This is the monthly report for the Hanford Laboratories Operation, February, 1960. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.
Date: March 15, 1960
Partner: UNT Libraries Government Documents Department
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PURIFICATION OF Pm$sup 147$ FROM FISSION-PRODUCED RARE EARTHS

Description: Promethium is purified from inactive and radioactive long-lived fission- produced rare earths by the use of Dowex 50 or Nalcite HCR cation exchangers and organic eluants. Americium and promethium, having hydrated ionic radii of the same size, are contained in the same fraction of the eluate. Promethium is purified from americium by adsorbing both elements on Dowex 1 (thiocyanate form) and eluting promethium from the resin with ammonium thiocyanate solution. Equllibrium studies were made in orde… more
Date: March 15, 1960
Creator: Pressly, R.S.
Partner: UNT Libraries Government Documents Department
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An IBM-704 Code for a Harmonics Method Applied to Two-Region Spherical Reactors

Description: An IBM-704 computer code for the harmonics method of criticality calculation for two-region spherical reactors is described. In the harmonics method, the criticality condition corresponds to the vanishing of a certain infiniteorder determinant; in practice, this condition is replaced by equating a finite-order approximating determinant to zero. By hand, the calculations can be performed conveniently only for second-order approximating determinants. The approximating determinant with the describ… more
Date: March 15, 1960
Creator: Chalkley, R.; Nestor, C. W. Jr. & Tobias, M. L.
Partner: UNT Libraries Government Documents Department
open access

SELECTION OF THE PIQUA OMR FUEL ELEMENT

Description: Two types of aluminum-clad uranium alloy fuel elements, a square (parallel flat plate) and a circular (concentric cylindrical shell) were investigated to determine their relative suitability for use in the Piqua Reactor. Nuclear, thermal, and mechanical data are given, and considerations leading to selection of the circular element are presented. Design dimensions are listed and reactor thermal design and operating conditions are given for the proposed element. (auth)
Date: March 15, 1960
Creator: Baumeister, E.B. & Wilde, J.D.
Partner: UNT Libraries Government Documents Department
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PATHFINDER ATOMIC POWER PLANT TECHNICAL PROGRESS REPORT FOR OCTOBER 1959- DECEMBER 1959

Description: >Technical progress on the research and development program being performed in connection with the design of the Sioux Falls Power Reactor is reported. Areas covered in detail include fuel element research and development, reactor mechanical studies, nuclear analysis, chemistry, instrumentation and control, plant safety analysis, feasibility studies, and steam plant and reactor auxiliary systems design. (For preceding period see ACNP-5924.) (W.D.M.)
Date: March 15, 1960
Partner: UNT Libraries Government Documents Department
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Viewing Equipment for Use in the HRT Core and Blanket Vessels

Description: S>Several viewing devices were built or purchased and various viewing aids were developed to permit examination of the core vessel of the Homogeneous Reactor Test. The hole in the core vessel was viewed and photographed from both the inside and the outside. A number of simple periscopes were designed and fabricated. A small television camera was purchased, and manipulators were built to insent it in the core or in the blanket. A radiation-resistant 21-ft-long periscope was procured which is … more
Date: March 15, 1960
Creator: Culver, J. S.
Partner: UNT Libraries Government Documents Department
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STUDY OF FUEL TEMPERATURE AND FLOW EFFECTS OF PLUGGING IN SRE FUEL CHANNELS

Description: The fuel surface temperatures and flow rate effects which might be produced by plugs in SRE fuel channels were investigated. It was found that plugs which locally insulate more than 20 to 30% of the surface of a fuel rod are dangerous. Heat transfer through the mcderator would not significantly affect the sensitivity of the outlet thermocouple to change in mass flow through a fuel channel, provided the reactor power is above 2 Mw and the channel is not completely plugged. If a 5% or greater inc… more
Date: March 15, 1960
Creator: Noyes, R.C.
Partner: UNT Libraries Government Documents Department
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AN ADVANCED SODIUM-GRAPHITE REACTOR NUCLEAR POWER PLANT

Description: An advanced sodium-cooled, graphite-moderated nuclear power plant is described which utilizes high-pressure, high-temperature steam to generate electricity at a high thermal efficiency. Steam is generated at 2400 psig, superheated to 1050 deg F and, after partial expansion in the turbine, reheated to 1000 deg F. Net thermal efficiency of the plant is 42.3%. In a plant sized to produce a net electrical output of 256 Mw, the estimated cost is 8232/kw. Estimated cost of power generation is 6.7 mil… more
Date: March 15, 1960
Creator: Churchill, J. R. & Renard, J.
Partner: UNT Libraries Government Documents Department
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Interim Report on Safety Procedures for the Task 2 Thermoelectric Generator

Description: Operational hazards associated with the use of a radioisotope-fueled auxiliary power unit for a satellite mission are evaluated. The entire fabrication-to-flight and/or retrieval and disposal sequence is examined and safe handling procedures suggested. The design and operation of the Task 2 thermoelectric generator is discussed. (C.J.G.)
Date: March 15, 1960
Creator: Klein, L. T.
Partner: UNT Libraries Government Documents Department
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An Advanced Sodium-Graphite Reactor Nuclear Power Plant

Description: Abstract: This report describes an advanced sodium-cooled, graphite-moderated nuclear power plant which utilizes high-pressure, high-temperature steam to generate electricity at a high thermal efficiency.
Date: March 15, 1960
Creator: Churchill, J. R. & Renard, J.
Partner: UNT Libraries Government Documents Department
open access

Thermal Cycling Tests of Porous Uranium

Description: Abstract: It has been proposed that a fuel element containing porous uranium be used for moderately high temperature applications. In an effort to determine the structural integrity of this type of fuel body, thermal cycling tests have been conducted on porous uranium.
Date: March 15, 1960
Creator: Wilkinson, L. E.
Partner: UNT Libraries Government Documents Department
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Graphite burnout, interim report on IP-25-A (PT-105-532-E)

Description: Graphite reacts with such gases as CO{sub 2}, O{sub 2}, or water vapor to form gaseous oxides of carbon. In the case of CO{sub 2}-graphite interaction, the reaction rate is not significant until about 550 C. Water oxidizes graphite, very roughly, three times faster than CO{sub 2}. Air will oxidize graphite appreciably at temperatures below 500 C. Graphite removal from Hanford reactors is very important, since graphite is used both as a structural support and a moderator for neutrons. Griggs has… more
Date: March 15, 1960
Creator: Ryan, B. A. & de Halas, D. R.
Partner: UNT Libraries Government Documents Department
open access

High Flux Isotope Reactor: a General Description

Description: Current status of the High Flux Isotope Reactor which is being planned for construction at Oak Ridge.
Date: March 15, 1960
Creator: Cole, T. E.
Partner: UNT Libraries Government Documents Department
open access

Specifications for Swaged UO2 19-Rod Cluster, PRTR Fuel Element Mark 1

Description: Specifications including detail dimension, materials, fabrication steps, acceptance criteria, and final assembly steps for the swaged uranium dioxide, 19-rod cluster Plutonium Recycle Test Reactor fuel element.
Date: March 15, 1960
Creator: Millhollen, M. K.
Partner: UNT Libraries Government Documents Department
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