Introduction: "This is the ninth letter report on ARF Project C 127, entitled "Preliminary Studies of Scavenging Systems Related to Radioactive Fallout." This report covers the period from August 1 to October 1, 1959."

In compliance with a request from Coolant Testing Operation, the Reactor Engineering Operation has made a study to determine the maximum operating pressure limits for the pertinent Zircaloy-2 process tubes. Since these tubes shall be used for testing NPR fuel elements, it is considered desirable that KER Loops 2 and 3 permit operation at temperatures of around 300{degrees}C while the NPR prototype facility permit operation at about 316{degrees}C in a manner such that there is minimum hazard to the KE-Reactor and to personnel.

The objectives of the test described in this report are to obtain current knowledge of the flow characteristics of the B and C Reactor flow systems under various transient conditions that have a reasonable probability of occurrence. These tests will be used to revaluate the B and C Reactor bulk temperature limits, the Ball 3X low pressure trip settings, and quantitative adequacy of emergency flow from the combined high tanks - export system.

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 …

Information related to the performance and operation of the Idaho Chemical Processing Plant is presented. The information is presented as answers to questions posed for Eurochemic assistance program. (J.R.D.)

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)

Foreign research and power reactors are tabulated. Nuclear power buildup goals are given for each nation on which information is available. (J.H.D.)

Comments by ICPP personnel on questions posed by Eurochemic Company are presented. Included is information on a fumeless dissolution system, recycle and rework system, liquid waste storage, criticality. in-line instrumentation, and a report listing Idaho Operations Area costs. (J.R.D.)

Data on critical assemblies of fissionable uranium metals are summarized. Empirical studies are made for spccific reflectors and gcometrics to determine the fcasibility of extrapolating these data for conditions concerning nuclcear safcty problems. Also included are the influencs on critical systems by various reficctors, U-235 isotopic enrichment, density, and small metal pieces homogcncously distributed in water. (auth)

As a result of a request for targets of uniform cross section for determination of the fission cross section of neptunium, an evaporation technique was developed for preparing films of the desired uniformity. Despite the fact that neptunium is an alpha emitter, it was possibIe to contain the activity and to subsequently recover the neptunium. (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)

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 integrated processes of nuclear industry are considered to define the nature of wastes. Processes for recovery and preparation of U and Th fuels produce wastes containing concentrated radioactive materials which present problems of confinement and dispersal. Fundamentals of waste treatment are considered from the standpoint of processes in which radioactive materials become a factor such as naturally occurring feed materials, fission products, and elements produced by parasitic neutron capture. In addition, the origin of concentrated fission product wastes is examined, as well as characteristics of present wastes and the level of fission products in wastes. Also, comments are included on high-level wastes from processes other than solvent extraction, active gaseous wastes, and low- to intermediate-level liquid wastes. (J.R.D.)

BS>Part I of this report describes the analysis of a cylindrical shell subjected to a temperature variation in the axial direction. This part is divided into two sections. Section 1 gives the formulas and the procedure for their application. ln Section 2, the derivations of the cylindrical shell formulas are given. A cylindrical shell subjected to a linear temperature variation through the thickness is considered in Part II. Part III cortains some sample problems which require the metheds of Parts I and II for their solutions. Although this pert demonstrates the use of the formulas derived, a fanniliarity with (1) is a prerequisite to an understanding of the material presented. (auth)

Information relative to the performance and operation of silver reactors is presented. The information is given as answers to questions submitted in the USAECEurochemic Company cooperative program. The answers to 8 questions are presented along with the basis for comments. (J.R.D.)

None

An estimate of the critical current for the case of breakup by an arc has been obtained. It is shown that the same technique allows a quick estimate of the critical current in the case of no arc, using neutral and plasma breakup instead. (auth)

Reprocessing of the ARE fuel was resumed after extensive leak testing in the pilot plant. This was considered necessary to assure no recurrence of gaseous UF/sub 6/ leaks as experienced in Run E-2. In the four additional runs required to complete the program, about 641 kg of fluoride salt containing 40.64 kg of fully enriched uranium was reprocessed. Recovery as UF/sub 6/ product represented 97.97% of the feed, with 0.01% measured losses. An additional 2.14% was reclaimed from NaF beds. The product was of sufficient purity to meet specifications for material designated for reduction to uranium metal. Decontamination from fission products was essentially complete. Calculations based on the entire ARE program indicated 96.38% product recovery, with 0.06% measured losses. An additional 2.50% was reclaimed from NaF beds and equipment washes. (auth)

The Early Program is designed to compute the peak power and total energy release vhich will occur in a reactor due to a step increase of reactivity. These excursions may be terminated by one of several automatic shutoff mechanisms caused by the power excursion itself. The calculation assumes no thermodynanaic delays and assumes void formation is due to heat conduction from the cladding to the moderator. (W.D.M.)

The early program is designed to compute the peak power and total energy release which will occur in a reactor due to a step increase of reactivity. The calculation assumes no thermodynamic delays and assumes void formation is due to heat conduction from the cladding to the moderator. For periods in the 1 to 10 msec. range, the calculation yields results which are in excellent agreement with SPERT experimental results. (W.D.M.)

Air samples were taken with the Andersen sampler and the cyclone separator. The data are reported on the Sr/sup 90/ and total beta activities of the samples. (W. L.H.)

The use of the porposed full reactor operational charge-discharge facilities will to some degree change the method of reactor operation from a ``batch`` to a ``continuous`` process. This change in the operational process will require modifications to coolant flow supply and monitoring equipment. Since a continuous and adequate coolant supply is essential to prevention of the release of fission products from the irradiated fuel elements this modification merits a hazards analysis.

S>These programs are intended to:ssist in establishing reflector cooling requirements, as well as thermal stresses in the reflectors and pressure vessel. The main program (304-Rec I) establishes required velocities between reflectors as a function of geometry, heat generation rates, material and fluid properties, and allowable surface temperature. This is done by iterating on velocities for successive water gaps. When required velocities have been established, corresponding thermal stresses are calculated. Rec II (301) and Prev I (301A) are supplement programs. Rec II is utilized for establishing surface temperatures of a reflector as a function of velocities, geometry, heat generation rates, and matenial and fluid properties. It handles only one reflector per case. This program also calculates thermal stresses in the reflector. Prev I is intended for calculations concerning a pressure vessel, perfectly insulated at the outer surface. It calculates surface temperature and thermal stresses in the vessel as a function of velocity, geometry, heat generation rates, and matenial and fluid properties. All calculations are performed at the midplane, where maximum heat generation rates occur. (auth)

In a preliminary experiment, the integral diffusivity of 1 M FeCl/sub 3/ solution varied uniformly with the fraction difiused: from 0.3 x 10/sup -6/ cm/ sup 2//sec for 10 per cent diffused to 0.8 x 10/sup -6/ for 30 per cent diffused. A short Fluorox run made with crude UF/sub 4/ in the 4-in. fluidized bed showed that UF/sub 6/ could be produced from the impure feed. Denitration of Th(NO/sub 3/)/sub 4/ solutions on fluidized or mechanically agitated beds of ThO/sub 2/ and gave fine ThO/sub 2/ particles for all conditions tested. The rate of sorption of uranium into 40 micron Dowex 21K resin particles from a 0.0042 M uranyl sulfate solution was studied by measuring the uranium loading on individual beads as a function of time. Chloride concentrations of 28 to 51 ppm were produced in the solvent extraction feeds during five feed adjustment runs made with the Darex Reference flowsheet. Decladding of SS-clad /sub 4/ gave essentially identical results as batch decladding. When a Mark I prototype assembly was sheared into 0.75-in. lengths with a"plane of contact" blade in the 126-ton Manco shear, 22.2 g of metal fines (304L stainless steel) 1680 microns or less in size was …