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PROBLEMS IN ACCOUNTABILITY MEASUREMENTS ASSOCIATED WITH THE INTERIM CHEMICAL PROCESSING PROGRAM

Description: Available knowledge of precision limits in S.S. accountability measurements and/or calculations by reactor and chemical processing groups is surveyed and summarized. Experienee in comparisons of reactor (production and research) calculations vs. chemical plant accountability measurements is also reported. A general tentative conclusion is that available precisions ( plus or minus 0.54 to plus or minus 0.78%) in chemical plant measurements (bulk and analytical) for fissionable material accountability is superior to the variable precision ( plus or minus 1.0 to 1l.0%) possible by calculations (nuclear and/or engineering) of power reactor systems; however, with operation and empirical experience (e.g., after two or three core loadings), it is believed that calculations for given reactors can attain acceptable precisions, e,g., less than plus or minus 1.0%. It may be proposed that fuel payments be made as follows: 90% of fuel value based on reactor calculations, an additional 5% based on dissolver analyses, and final settlement based on chemical plant material balance (product plus loss analyses). (auth)
Date: May 28, 1959
Creator: Arnold, E D & Gresky, A T
Partner: UNT Libraries Government Documents Department

AN EVALUATION OF THE PROPERTIES AND BEHAVIOR OF ZIRCONIUM-URANIUM ALLOYS

Description: Data from a survey of the literature and other available information on zirconium--uranium alloys have been reviewed for the purpose of obtaining a coherent picture of current knowledge about the properties and behavior of zirconium--uranium alloys. The results of the survey were used to revise and extend the presentation of material gathered earlier in a similar study and reported in BMI1030 in August 1955. The constitution of zirconium-uranium alloys is discussed, and a constitutional diagram for the system is presented. The effects of oxygen and nitrogen, which are present in these alloys as contaminants, on alloy constitution ars shownin the form of ternary diagrams and in terms of their quantitative effects on the phases present. The transformation kinetics and the nature of the transformation of the high-temperature body- centered-cubic gamma phase to the phases stable at room temperature are described. Two regions are discussed: in the 20 to 70 wi. % uranium composition range, gamma, which is retained on quenching, transforms isothermally to the intermediate epsilon-phase structure by a diffusion-controlled nucleation-and- growth process; in alloys containing less than 20 wt.% uranium, gamma transforms martensitically to a strained alpha-zirconium structure on quenching, with the diffusion-controlled transformation of gamma to either epsilon or alpha proceeding above the M/sub s/ temperature. Typical zirconium -- uranium alloy microstructures produced by working and by various heat treatments are shown, and a correlation of the structure with the manner in which it was produced and the resultant mechanical properties is attempted. Hardness, tensile, and creep data for zirconium--uranium alloys are tabulated and evaluated in terms of alloy constitution and transformation-kinetics behavior. Physical-property data for zirconium-uranium alloys are presented in tabular and graphical form. The best fabrication techniques for forging, hot and cold rolling, extruding, and swaging are detailed. The corrosion resistance of zirconium-uranium alloys in wafer ...
Date: September 28, 1959
Creator: Bauer, A.A. ed.
Partner: UNT Libraries Government Documents Department

ISOTOPE SEPARATION AND ISOTOPE EXCHANGE. A Bibliography with Abstracts

Description: The unclassified literature covering 2498 reports from 1907 through 1957 has been searched for isotopic exchange and isotepic separation reactions involving U and the lighter elements of the periodic chart through atomic number 30. From 1953 to 1957, all elements were included Numerous references to isotope properties, isotopic ratios, and kinetic isotope effects were included. This is a complete revision of TID-3036 (Revised) issued June 4, 1954. An author index is included. (auth)
Date: October 28, 1959
Creator: Begun, G.M.
Partner: UNT Libraries Government Documents Department

Waste Treatment and Disposal Problems of the Future Nuclear Power Industry

Description: The elements of waste treatment and disposal are assessed which are expected to become important in the development of the nuclear power industry of the future. Growth of the nuclear power economy is considered along with composition and quantities of anticipated waste. In addition, the economic implications of waste disposal are considered. It is concluded that research should be concentrated on decontaminating off-gases and on conversion of wastes to a more suitable form than liquid for storage. (J.R.D.)
Date: January 28, 1959
Creator: Bruce, F.R.
Partner: UNT Libraries Government Documents Department

Increased production from deliberate discharge cycling

Description: Considerable production gains might be attained if each reactor discharged its entire flattened region during one scheduled outage instead of utilizing several outages for this purpose. Several of the older reactors are now discharging a high percentage of their flattened zones in a single outage and could be put into this type of operation with relatively little difficulty. Production gains may be possible through better flattening efficiency, a more favorable rupture rate effect, fewer non-equilibrium losses, higher conversion ratio, and more efficient usage of outage work. Since this document is written Primarily from the Operational Physics standpoint, some gains and pitfalls which must be evaluated by other affected groups will only be mentioned here as possibilities. The purpose of this document is simply to point out the potential gains in flattening efficiency from this method. Potential gains from improved fuel performance have been described in another document.
Date: May 28, 1959
Creator: Carter, R. D.
Partner: UNT Libraries Government Documents Department

Toy Top Plasma Injector

Description: Introduction: "It is the purpose of this note to describe the construction and operation of the plasma injectors used in the magnetic high compression experiments in progress at the Lawrence Radiation Laboratory at Livermore. As the investigations of these injections is still in progress, remarks concerning their operation or the characteristics of the injected plasma are of a tentative nature."
Date: May 28, 1959
Creator: Coensgen, F. H.; Cummins, W. & Sherman, A.
Partner: UNT Libraries Government Documents Department

Toy Top Plasma Injector

Description: The construction and operation of the plasma injector, Toy Top, used ia the magnetic high compression experements in progess at the Lawrence Radiation Jab. at Livemore are described The essential part of the injector consists of a stack of deuterated titanium washers 3/4 in. O.D. and/2 in. I.D. Details of the construction are sbown (W.D.M.)
Date: May 28, 1959
Creator: Coensgen, F.; Cummins, W. & Sherman, A.
Partner: UNT Libraries Government Documents Department

Analysis of 100-K emergency water requirements after CGI-844 pump failure

Description: The demand plot has a 5-set, modified pump decay curve; it shows that 20,000 gpm emergency flow would be required within 80 seconds of complete pump power failure. Bases for the demand curve are constant bulk inlet temperature of 2 C, constant bulk outlet temperature of 95 C, K-3 I&E fuel elements, and initial reactor flow of 188,000 gpm.
Date: May 28, 1959
Creator: Corlett, R. F.
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

STATUS REVIEW OF THE KEWB PROGRAM

Description: jectives, the accomplishments, and a summary of the work outstanding. The obtectives of the experimental and analytical studies were to investigate and reach an understanding of the kinetic behavior of aqueous homogeneous reactors. Information produced by the program, experiments on the spherical core, capsule experiments, and the remaining work schedule are discussed. (W.D.M.)
Date: January 28, 1959
Creator: Flora, J. W.
Partner: UNT Libraries Government Documents Department

HIGH TEMPERATURE CORROSION STUDY INTERIM REPORT FOR THE PERIOD NOVEMBER 1958 THROUGH MAY 1959

Description: Samples of grade A Monel snd grade A nickel were subjected statically in a single reactor to an undiluted atmosphere of gaseous fluorine at pressures up to one atmosphere absolute and temperatures up to 1500 deg F. The grade A Monel was conservatively estimated to have consumed at least 40 times as much fluorine as grade A nickel during the entire period of the investigation. Samples of fused alpha Al/sub 2/O/sub 3/, alpha -Al/sub 2/O/sub 3/- MgO spinel, and alpha -Al/sub 2/O/sub 3/-NiO--nickel cermet were exposed to undiluted fluorine at one atinosphere absolute pressure at temperatures of 1340 and 1500 deg F. Results indicated that the alpha -Al/sub 2/O/sub 3/ is as good as the Ni in the region of 1300 deg F. Grade A nickel samples coated with nickel fluoride filins of 37,000 and 74,000 A, respectively, were exposed to an absolute pressure of gaseous UF/sub 6/ of 12 cm of Hg at temperatures of 1000 and 1800 deg F. (W.L.H.)
Date: July 28, 1959
Creator: Hale, C.F.; Barber, E.J.; Bernhardt, H.A. & Rapp, K.E.
Partner: UNT Libraries Government Documents Department

Design of production test IP-280-A-FP: Irradiation of alloyed dingot uranium fuel elements

Description: Objective of this test is to authorize irradiation of alloyed, low hydrogen dingot uranium fuel elements on a pilot scale, and to monitor their performance. Initially, 25 tons per month of alloyed, low hydrogen dingot material will be charged for two months. Measured charges will be loaded with the initial 25 tons to monitor the stability of this material. Following a two-month delay in the monitor charging, and if the dingot meets all specifications, routine charging of quantities up to 60 tons/ month may proceed for six months and, assuming continued favorable performance, up to 150 tons/month may be accepted to complete large scale evaluation of dingot uranium, and on a continuing basis thereafter.
Date: August 28, 1959
Creator: Hall, R. E. & Hodgson, W. H.
Partner: UNT Libraries Government Documents Department

PRELIMINARY EVALUATION OF A PROPOSED FUEL MATERIAL FOR HIGH TEMPERATURE REACTORS

Description: Results are reported for preliminary experiments to determine the stability of solid solutions of UO/sub 2/ and ThO/sub 2/ in air at temperatures of 2500 deg F and above. The results are compared with those obtained by workers at Argonne during development work on the Borax IV experimental breeder reactor. It is concluded that further evaluation of the system at temperatures above 2500 deg F is required. (auth)
Date: October 28, 1959
Creator: Juenke, E.F.
Partner: UNT Libraries Government Documents Department

A STUDY OF PROBLEMS ASSOCIATED WITH RELEASE OF FISSION PRODUCTS FROM CERAMIC FUELS IN GAS-COOLED REACTORS

Description: The diffusion of fission products out of the fuel elements leads to increased shielding requirements, a greater hazard due to their possible release to the surroundings, and more difficult maintenance problems. Continuous processing of the contaminated coolant may alleviate the hazard and maintenance problems; however, extensive in-pile loop experiments are needed for a quantitative evaluation of methods. By proper design of major components such as heat exchangers and blowers, direct maintenance of contaminated equipment may be possible, with or without premaintenance decontamination Such an aporoach is to be preferred to that of providing remote maintenance facilities which, in the case of the reactors considered added from 0.7 to 1.8 mills/kwhr to the cost of power. (auth)
Date: October 28, 1959
Creator: Lane, J.A.; Bennett, L.L.; Culver, H.N.; King, L.J.; Sanders, J.P.; Scott, J.L. et al.
Partner: UNT Libraries Government Documents Department

DEFUELING THE S2G REACTOR

Description: The defueling of the S2G Reactor which was conducted at the Electric Boat Division, General Dynamics Corporation Groton Connecticut during January 1959, is reported from the viewpoint of the participating personnel from Knolls Atomic Power Laboratory. The sequence of events is outlined, difficulties encountered during the operation are described, and conclusions of possible interest to other naval nuclear reactors are given (auth)
Date: May 28, 1959
Creator: Moore, C.V.
Partner: UNT Libraries Government Documents Department

GAS-PRESSURE BONDING OF ZIRCALOY-CLAD FLAT-PLATE URANIUM DIOXIDE FUEL ELEMENTS

Description: A solid-state bonding technique involving the use of gas pressure at elevated temperatures was investigated for the preparation of compartmented Zircaloy-clad flat-plate uranium dioxide fuel elements. These investigations involved development of methods for the surface preparation and assembly of fuel- element components for bonding, determination of optimum bonding parameters, development of barrier coatings for uranium dioxide to prevent reaction with Zircaloy, and extensive testing and evaluation of the bonded fuel elements. During the course of this work, the process was continually modified and refined in an effort to improve the quality of the bonded element and decrease the cost of fabrication. The surface-preparation studies indicated that satisfactory bonding could be obtained consistently with both machined and belt-abraded components. Belt abrasion is more economical and was used as the standard technique in the development phases of the program. Initially the elements were assembled into a stainless steel or Ti-Namel envelope which was evacuated and sealed prior to bonding. Later studies showed that the quality of bonded elements could be improved and process costs decreased by edge welding the Zircaloy components to form a gastight assembly that was then bonded without use of a protective envelope. Further cost reductions were incorporated into the process by the use of piece Zircaloy components to form the picture frame. Optimum bending with a minimum core-to-cladding reaction was achieved by pressure bonding at 1500 to 1550 deg F for 4 hr using a helium gas pressure of 10,000 psi. A postbonding heat treatment for 5 min at 1850 deg F in a salt bath promoted additional grain growth at the bond interface during the alpha-to-beta transformation. Barrier layers of graphite. chronaium, iron. molybdenum, nickel, niobium, palladium, and various oxides were investigated to prevent reaction between the UO/sub 2/ core and Zircaloy cladding. Graphite, in the form ...
Date: August 28, 1959
Creator: Paprocki, S.J.; Hodge, E.S.; Carmichael, D.C. & Gripshover, P.J.
Partner: UNT Libraries Government Documents Department

PROPERTIES OF URANIUM DIOXIDE-STAINLESS STEEL DISPERSION FUEL PLATES

Description: The physical and mechanical properties of GCRE-type fuel elements were determined from room temperature to 1650 deg F. The fuel elements were prepared by cladding Type 318 stainless steel sheet to a core containing 15 to 35 wt.% UO/ sub 2/ in either prealloyed Type 318 stainless steel or elemental iron-18 wt.% chromium-14 wt. % nickel-2.5 wt. % molybdenum. The tensile strength in the direction perpendicular to the rolling plane decreased from 24,600 psi at room temperature to 9,200 psi at 1650 deg F for the reference fuel plate, whose core contained 25 wt.% UO/sub 2/ in the elemental alloy. The tensile strength in the longitudinal direction for this fuel element ranged from 54,800 psi at room temperature to 14,200 psi at 1650 deg F, with elongation in 2 in. ranging from 8 to 13 per cent. The extrapolated stress for 1000hr rupture life at 1650 deg F was 1800 psi, and a 1.4T bend was withstood without cracking. The mean linear thermal coefficient of expansion was 11.0 x 10/sup -6/ per deg F for the range 68 to 1700 deg F. (auth)
Date: April 28, 1959
Creator: Paprocki, S.J.; Keller, D.L. & Fackelmann, J.M.
Partner: UNT Libraries Government Documents Department