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.

Several decisions regarding basic project philosophy must be made in order to proceed with scope design and the preparation of equipment procurement specifcations. The purpose of this document is to present as much pertinent data as possible to allow the project representatives to become familiar with the problems involved. A meeting of Representatives is planned for the near future after receipt of project authorization to discuss the scope of this project and its relationship to CG-775. Emergency flow requirements of the K reactors for planned future power levels is approximately 32,000 gpm within 68 sec. A detailed study of the existing high-pressure cross-tie line reveals that a duplicate cross-tie line and five low lift pump operation would be required to provide this flow. The existing emergency generation capacity is not adequate to supply five low lift pumps and all other necessary emergency electrical loads. A possible solution to adequate emergency flows is to connect the proposed steam turbine pump directly to the risers and to consider the turbine pump as the last ditch system. If it is determined that this does not meet the criteria of separate systems, then an alternate solution must be found.

This document presents a nomograph from which the relationship between reactor operating parameters, tube power, and outlet temperature can be correlated with rupture rate. The index indicates the severity of the reactor climate during irradiation and does not include the metal quality parameters defined in the rupture rate equation. The general form of the Operational Severity Index Equation is OSI=P{sup 3.3}/1000{times}t{sub 0}{sup 8.7}/100, where OSI, is the unitless Operational Severity Index, P is the tube power in kW, and t{sub 0} is the tube outlet temperature, in degrees C.

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.

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.

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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)

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 …

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)

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)

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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.)

A 3-inch-diameter flame tower for the conversion of uranosic oxide to uranium hexafluoride with elemental fluorine was tested for possible use in the fluorination step of the present uranium recovery process. The oxide was fed from a hopper to the tower by a screw feeder. The fluorine and the oxide entered at the top and flowed concurrently down through the tower. The unreacted or partially reacted oxide was collected in an ash receiver at the bottom. Fine solid particles were removed from the gas stream by an electrostatic precipitator and a tune-type filter. The uranium hexafluoride was collected in cold traps. Twenty-five experimental runs were conducted with average oxide feed rates from 3.73 to 19.38 lb/hr. The average fluorine flow rates were from 7.5% below to 44% above the stoichiometric amount of fluorine required. The best operating conditions were at a feed rate of 15 lb of oxide per hour with a minimum fluorine excess of 75% 110.6 lb of fluorine per hr). The material collected in the tower ash receiver represented between 6.0 and 10.0 percent of the total amount of uranium fed during the run. The ash, combined with an equal weight of oxide, can be fed back …

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)

Materials of construction were evaluated for use in critical areas of head-end processes for the aqueous reprocessing of spent nuclear fuel elements. The SulfexThorex, Darex-Thorex, Darex, Zirflex, and Zircex processes were considered. The effect of varying heat treatments on the resistance of the materials was also evaluated. Dissolution of unirradiated fuel pins was carried out in vessels of promising materials. The corrosion rate of Ni-o-nel was about 5 mils per month during actual fuel-pin dissolution by the Sulfex-Thorex process. Stabilization and heat treatment are necessary to prevent intergranular attack at welds. Carpenter 20 Cb is subject to stress-corrosion cracking by the Sulfex decladding solution and Illium R behaves similarly to Ni-o-nel in Thorex solutions. Titanium shows promise as a construction material for a Darex-Thorex dissolver. However, several questions remain concerning a vapor-phase attack observed around certain weldments. Carpenter 20 Cb, Ni-o-nel, and Types 309 and 309S Cb stainless steel appeared worthy of further study for the Zirflex dissolver. Preliminary evaluations show that at least Ni-o-nel and Carpenter 20 Cb should be studied further as possible construction materials for a single vessel for Zirflex and Sulfex-Thorex processes. Illium R, Hastellcy C, and nickel were not attacked by hydrcchlorination conditions of the …

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.)

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)

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 …

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.)

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Volatility Pilot Plant Mark III Fluorinator is a doublechamber type vessel, each chamber 2 1/2 ft by 16 in. outside diameter separated by a 5-in. pipe 15 in. long. ASME flanged and dished heads are used for the chamber tops and conical sections with a 60 deg apex angle for the chamber bottoms. A new furnace designed to maintain the complete lower chamber (molten salt+ freeboard) above melt temperature is to eliminate past experiences of salt solidification on the wall, heads, and in or on the internal process lines. External pipe runs are autoresistance heated to allow melting and drain back of salt plugs. The upper chamber serves as a gas de-entrainment and solids precipitation device to retain most of the entrained salt and condensable fluorides in the 100 to 400 deg C range. (auth)

Neutron multiplication measurements were performed on two identical finite Pu-metal slab assemblies separated and reflected by plexiglas. (auth)

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)

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