Prior to the startup of the Purex Plant it was considered desirable to study the ``self-concentration`` of wastes likely to be encountered in the Purex waste storage system. Consequently, a 39-inch diameter 35-foot high tank was built at the Hot Semiworks. This tank was designed to represent a core of the 75-foot diameter Purex waste tanks. During the period of Purex testing at the Hot Semiworks the tank was filled with ``typical`` Purex waste and self-concentration phenomena were studied. As pertinent data were developed from the studies it was informally transmitted to interested parties for use in establishing safe waste storage criterion for the Purex system. This document summarizes the Hot Semiworks waste self-concentrator basic data. No attempt has been made to cross-correlate this data with actual Purex experience although data are sufficiently detailed to permit calculation of such correlations.

Areas covered in this report are as follows: nuclear safety in the Fuel Preparation Dept.; studies related to the present production reactors; studies related to future production reactors; studies related to separation plants; studies form the neutron cross section program; reactor development (gas cooled reactor program and test reactor operations); and biology and medicine.

This report discusses two of the direct cast core I & E fuel elements irradiated in C Reactor under PT-IP-93 which were received at the Radiometallurgy Laboratory for examination during July 1959. One of the elements was badly warped and distorted while the other of the same series had remained relatively unchanged. Both pieces had received an average exposure of 689 MWD/T. Warp and diameter measurements show that the one element had warped approximately .040 in. and had become distorted in two longitudinal planes at the male end. The distortion planes are at right angles which gives the transverse section a square a appearance. A comparison of the grain size from three positions along the warped element with a section cut from the unwarped element revealed a greater number of large grains in the zones of highest distortion. The maximum size grains, were in the order of 2.5 mm -- 3.8 mm vs normal .2 mm. Also prevalent in all sections examined vas a columnar type grain structure around the outside periphery. It is evident that the pre-irradiation heat treatment failed to refine the grain size sufficiently to prevent nonuniform growth of this element during irradiation. A corrosion pit was …

This monthly document details activities of the Hanford Operations Office during the month of March 1959. (FI)

This monthly document details activities of the Hanford Operations Office during the month of May 1959. (FI)

This report records the basis underlying the decision already reached to recommend a change of process tube fittings at B, D, F, DR and H reactors, and also proposes for consideration a change in the devices for monitoring effluent water temperature.

Experiments were conducted to determine the relative flow characteristics of J-pigtail adapters supplied by Lampson Products Co., Seattle.

Water studies were performed on the conical bottom, bottom-polar inlet, core vessel to provide design information for the high-pressure model being installed in the 300SM loop. Maximum heat transfer coefficients are obtained in the conical portion of the core vessel when the inlet nozzle is inserted into the core up to the point of extension of the imaginary continuation of the top hemisphere. The flow rate in the by-pass of the 300SM loop was calculated to be 9.5to 11 gpm (7to 8 ft/ sec). The assumptions on which this calculation was based were checked experimentally by instapling a simiiar by-pass in a low pressure mockup using an almost identical core vessel, The flow pattern was observed by using dye probes. Boundary layer flow exists at the wall of the upper hemisphere, while the flow in the conical portion of the core is characterized by large eddies extending to the vessel wall. (auth)

The advantages of storing nuclear pulse height information on tape for later use with an analyzer are pointed out. Experiments with a magnetic modulator playback head which would offset inaccuracies due to tape speed variation are described. (T.R.H.)

The corrosion behavior of types 304L and 309SCb stainless steel was examined during the production and first half of the development periods of Thorex pilot plant operation between December 15, 1954 and November 10, 1955. The process vessels in which corrosion data were obtained included the batch dissolver tank the feed adjustment tank the AP catch tank and the BT vapor separator. There was no significant difference between the corrosion resistance of the two alloys in any of the environments. Vapor-phase corrosion attack in some vessels was more severe than solution-phase attack in the same vessels, Dependent upon the particular process environment, corrosion rates in the vapor phase ranged from 0.1 to 46 mpy; solution-phase corrosion rates ranged from 0.1 to 28 mpy. The most severe attack in both solution and vapor phases occurred in the feed adjustment tank There was serious corrosion in the BT vapor separator also. Corrosion was moderate (8 mpy or less) in the batch dissolver tank and was almost negligible (0.2 mpy or less) in the AP catch tank. The attack on rolled surfaces of both types of stainless steel was uniform, with the exception that shallow, intergranular penetration was experienced in the more corrosive …

A preliminary study was made of the nuclear stability of reactors in which particles containing U/sup 233/ and Th/sup 232/ are fluidized by liquid water. D/sub 2/O moderated reactors were found to be least sensitive to changes in bed height at thorium concentrations in the range 80 to 120 g/liter, H/sub 2/O reactors in the rarnge 1500 io 2400 g Th/ liter, and mixed H/sub 2/O-D/sub 2/O reactors at thorium concentrations between those for pure moderators. A reactor operated at the thorium concentration at which it is least sensitive to change in bed height was found to be stable in response to a cyclical variation in height. (auth)

The use of a scalloped cross-sectionsl coolant channel has been suggested as a possible solution of the fuel-element bowing problem inherent in the septafoil type of geometry. Using simplified assumptions, a method has been developed for calculating the red spacing and scallop size necessary to produce equal average fuel-element surface temperatures in the central and peripheral regions of the coolant flow channel at the mid-section of each fuel-red cluster under a given set of reactor flow conditions. Since the extent of rod-bowing is related to the surface temperature distribution, this requirement should minimize fuel-element deflection. ( auth) l6629 In heavy-water-cooled and -moderated power reactors such as NPD-2 and CANDU the coolant will be insulated from the moderator by a gas layer external to the pressure tubes. Therefore, these tubes must be designed to operate at the maximum coolant temperature. However, by placing the insulation inside the pressure tubes they can be kept cold; a thiner wall or a higher reactor operating temperature can be used. and a wider range of pressure tube materials be considered. Two kinds of internal insulstion are considered: the cooled pressure tube in which stagnant heavy water is the insulation, and the cooled stress tube in …

Studies were made to determine the actual surface temperature of OMRE fuel elements if the thermocouple were not present. Chromel-alumel thermocouples are being attached to the fuel plate cladding of Type 304 stainless steel. These wires are in contact with the coolant stream. Heat transfer from the thermocouple junction, by conduction along the lead-wires and by forced convection to the coolant, produces a lowering of the surface temperature in the region of the junction which results in an error in surface temperature measurement. (W.L.H.)

Standard water treatment practices have proven adequate for scale and corrosion control on the secondary side of the ORR pool-cooling heat exchanger. Corrosion rates measured on specimens exposed to secondary water showed maximum corrosion rates of approximately 1 mpy during the first exposure period and these rates decreased during subsequent exposure. Examination of the heat exchanger tube bundle after 6 months and 13 months of operation showed only minor film formation and very little evidence of corrosion. (auth)

A study was made of the recrystallization behavior of beta-quenched uranium by incorporating the effects of transfer time from the beta heat-treating bath to the quench media. A critical transfer lime which relates to the time required to immerse samples at a temperature above the beta-alpha transformation in a quench media directly affects the beta-cooled structure and its ease of recrystallization upon alpha phase annealing. Within the critical transfer time, the ease of recrystallization and degree of grain refinement of beta-cooled samples as measured by alpha-phase annealing temperatures and times are superior to samples quenched beyond the critical transfer time. (auth)

Formulas are derived for the equilibrium orbit, isochronous condition, vertical and horizontal betatron frequencies, and for the effects of the 3/3 radial resonance in a three-fold geometry. The magnetic field is represented by a Fourier series in azimuth with amplitudes expanded in a Taylor series about the reference radius. The form is such that the various parameters may be deduced from an arbitrary set of field measurements in the median plane and the results obtained by direct substitution in algebraic formulas.

A set of equations was written which allowed the calculation of equilibrium concentrations in the solvent phase of nitric acid and uranyl nitrate, given the initial TBP concentration in the solvent and the aqueous phase concentrations of acid and uranium. An ORACLE subroutine was written based on these equations which will calculate a set of equilibrium values in about 35 milliseconds. The subroutine was incorporated into a short program to calculate equilibrium points. A set of calculated points are presented in graphical form for 5, up to 7 M and uranyl nitrate concentrations up to 0.6 M in the aqueous phase. Since these calculations were based on limited data, it is the purpose of this memo to invite comment on their accuracy. The equations were reworked from a Russian paper by A. M. Rozen and L. P. Khorkhorina, using emperical polynomial fits for some poorly defined parameters, and were based principally on data from J. W. Codding. This project was undertaken to provide a subroutine to calculate equilibria in a general purpose ORACLE program to calculate the performance of multi-stage compound extraction cascades with more than one significant solute. (auth)

The burns to the skin of anesthetized rats were determined for the thermal radiation pulses of a carbon arc on a hot-wet uniform when spaced 5 mm from the skin. The radiant exposures to cause burns resulting in eschar were tion pulses corresponding to 250, 1000, 2900, and 10,000 kiloton detonations, respectively. The threshold lesions were caused by volatile products not associated with ignition. The associated temperatures were recorded. (auth)

Tests were conducted to determine the amount of error introduced in reactor power data by radiation-induced voltages in cables and electrometer preamplifier chassis. The results, obtained near the central exposure facility of the KEWB (Reacter Safety Experiments), showed no observable radiation effects under the conditions of present use. Cable insulation resistance was measured during the radiation bursts. (C.J.G.)

In 1958, planned large scale use of the new I & E slug geometry at more severe operating conditions than had been generally experienced suggested a possible compromise in reactor life and safety if a reasonable degree of rupture control with the new type of element was not maintained. The formalized slug corrosion limit (Top-of-Annulus limit) was issued as a Process Standard at the time of the full-scale loading of I & E geometry fuel elements to provide this limit for reactor operation. The loading of I & E slugs at all reactors has been accomplished and initial power level increases have been made. To date, 67 I & E ruptures have been sustained including both `hole` and `annulus` failures. The type and behavior of ruptures to be expected with I & E geometry are now characterized. Recent studies have indicated that the I & E failure experience is consistent with the general mathematical rupture model formulated from analysis of solid slug experience. Increased confidence in the use of this model in combination with Optimization Studies permits greater emphasis to be placed on the rupture model as a guide for reactor operation. It is the purpose of this report to …