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 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)

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.

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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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.)

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)

Sections of the dummies furnished by Radiological Engineering, Process Reactor Development Operation were machined into 1/4 inch diameter by 1 inch long cylinders and irradiated in the Quickie Facility at F area. The pieces were discharged directly into a holder one foot from the Beckman chamber. The transient time from in-pile to the chamber is approximately 30 seconds. The readings were taken using a Beckman chamber, Beckman Micro-Micro Ammeter and Recorder. This system has been calibrated with Co{sup 60} sources obtained from the Oak Ridge National Laboratory. We are including data taken from a sample of 61-S and 99.998 per cent aluminum which may be of interest.

Results of four experimentul runs in the Fluorox fluidized bed reactor system are reported. The engineering feasibility of UF/sub 6/ production from UF/ sub 4/ by use of dry air of O/sub 2/, 2UF/sub 4/ + O/sub 2/ = UF/sub 6/+ UO/sub 2/ F/sub 2/, in an Inconel fluidized bed reactor at 800 to 850 deg C was demonstrated in two experimental tests in which greater than 90% of the theoretical amount of UF/sub 6/ was collected or measured. Two runs made with crude UF/sub 4/ (produced from unpurified mill concentrate) as the feed material, showed thnt UF/sub 6/ could be produced at 700 to 725 deg C but corrosion on Inconel was prohibitive. (auth)

The proposed operating conditions for fuel elements to be charged into the NPR require the fuel to be of an extended surface geometry and maintain adequate strength and corrosion resistance in 300 C water. A contract was let to Nuclear Metals Inc. to produce by co-extrusion lengths of fuel rod containing both natural and 1.6% enriched uranium of irradiation quality for fabrication into fuel elements. The fuel rods used in the irradiation test represent the first enriched uranium rods coextruded in 0.030 inches of Zircaloy-2 to be irradiated and examined at Hanford. The rods used for this test were fabricated into four, 4 rod cluster fuel elements thus allowing adequate space between individual rods for expansion in the case of a fuel rod failure. This rod was of particular interest since it contained an irregular uranium-Zircaloy-2 interface. The purpose of the irradiation was to determine the dimensional stability of coextruded fuel rods and to determine whether the irregularity in the bond interface had any effect upon the irradiation performance of the fuel. Fuel elements were irradiated in 200 C water in the KER Loop 2 facility to an exposure of 0.28 a/o burnup (2,200 MWD/T). Post irradiation examination showed that …

At the present time, plants in the 100 to 500 Mw size range are more numerous and carry the greatest portion (over 50%) of the total steam-electric plant load in the US utillties industry. The contribution of plants of over 1,000-Mw capacity is increasing more rapidly than any other size classification and at present represents about 10% of the total capacity. By 1962 the TVA will have six plants with capacities of over 1,000-Mw. The largest steam-electric plant in the US is the TVA plant at Kingston, Tenn., with a nameplate capacity of 1,440Mw. Turbine-generator wits are also following a trend of ever-increasing size. In present construction, the 150 to 200 Mw size range for units ls the most common and represents the greatest contribution to capacity. Two units of 500-Mw nameplate rating each, the largest in the US, are on order by the TVA, and an 80 Mw unit is contemplated. (auth)

By applying the relationships between lattice parameters and composition in the solid solution range, a method was presented to analyze boron in boron carbide. The purpose was to establish a means of determining the boron composition in a boron carbide solid solution in the presence of excess carbon or boron. The different unit-cell dimensions sometimes reported for a given solid mixture are usually attributed to the existence of solid solutions. If the crystal lattice parameters vary linearly with composition, a curve can be constructed to determine composition from lattice parameter measurements. Toward this end four samples of boron carbide of known nominal composition were selected. For each sample precision lattice parameter measurements and chemical analyses of boron and carbon content were obtained to establish relationship between lattice parameters and boron composition. (auth)

The Army Gas Cooled Reactor System Program includes water moderated heterogeneous reactor (Gas Cooled Reactor Experiment I), a graphite moderated homogeneous reactor (Gas Cooled Reactor Experiment II), a mobile gas cooled reactor (ML-1), and the co ordination of thc Gas Turbine Test Facility. The progress of each project, the associated tests and data evaluation, the applicabie design criteria, and the fabrication of reactor components are briefly summarized. (For preceding period see IDO-28538.) (W.D.M.)

In early July 1958, it appeared that Columbia River temperatures at HAPO would be near 24--50{degree}C by the end of August. River temperatures were averaging 40 to 50{degree}C above 1957 figures and were 3{degree} to 4{degree} above the ten year highs. It seemed desirable to examine the problem to determine if any corrective measure could be taken, since it was apparent that production losses were imminent. The large storage of cold water behind Grand Coulee Dam, normally untapped, was a source of possible relief. A plan for use was proposed for the peak high temperature period and agreed to by the Bureau of Reclamation.

The temperatures of the Columbia River was reduced 1 to 30 Centigrade with beneficial effects at HAPO. It is reasonable to expect that future benefits may be possible. It is desirable that the temperature of the river be controlled each year to the maximum extent possible. Instrumentation improvements requested to effect optimum savings. Records of river temperatures and flows should continue to be maintained by IPD as a necessary part of temperature optimization. Where possible, the coincident use of the river cooling technique should be made for the benefit of anadromous fish.

The design of internally cooled electrical coils for the production of high intensity magnetic fields presents many new aspects and combinations of the familiar modes of heat transfer. However, the customary methodology appears to be sufficient for preliminary analysis and understanding of those problems. This methodology comprises the derivation of a qualitative, approximate equation expressing the relative performance of the various parts of a system, followed by an examination of this equation in order to locate the limiting features of the system. These features are then investigated by more powerful methods, which in turn provide guidance for development research in the laboratory. (auth)

An elementary discussion is given of the reactor parameters which determine the signal-to-noise ratio obtainable with a pile oscillator. (auth)

This document details activities of the Fuels Preparation Department during the month of April 1959. (FI)

The purpose of this production test is to provide a standard method of obtaining scram transient reactivity information at the eight reactors, under conditions conducive to valid data. These conditions include the bypassing of the Panellit system at a low power level for a short, controlled period of time during May 1959.

The basic theory of operation of the triggered spark gap is established, and qualitative and quantitative engineering design data are given. From the basic twoelectrode gap, a three-electrode or triggered gap model is established with its static and dynamic triggering characteristics shown. Several geometry conditions such as gap spacings trigger electrode hole sizes and insulator effects are discussed, showing their influence upon the triggering mechanism. A suggested trigger mechanism is given based on that proposed by Sletten and Lewis for the trigatron and modified to fit the present analysis. (auth)

Production of Pu from separations plants was only 58% of April commitment because of Purex difficulties. UO{sub 3} production, shipments met schedules. Pu shape production and shipments exceeded forecast by 14%. Purex HS column, repaired Oct 1958, developed another leak and was bypassed April 18, resulting in Pu and U that required reprocessing. A Palm recovery run at Purex with all- reducing flowsheet, resulted in 87% recovery and excellent decontamination of product. The prototype dual-pass silver reactor in Purex C-cell plugged with offgases. Processing of unclarified feed through Purex solvent extraction continued. Redox dissolver batch sizes for E-metal processing were increased from 1.75 to 2.0 tons. Testing of first extraction cycle acidic flowsheet at Redox continued, with Np losses to HAW being below detection limit. Ru in 1AFS stream increased 10-fold F.P. activity but was removed in acid deficient U cycles. A sulfamic acid process is being explored for dissolving Pu metal. Scope design of Redox dissolver and RMA line replacement at Z plant was completed. Shielding effectiveness of medium and high density x-ray lead glass was compared.

This document details activities of the irradiation processing department during the month of April, 1959. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and engineering operations; production and reactor operations; facilities engineering operation; employee relations operation; and financial operation.

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