Further studies are reported on the analytical behavior and experimental testing of Cs/sup 137/ plaque irradiator designs. Low-level sources used for the initial experiments consisted of about 7 mc of Cs/sup 137/ aqueous solution in brass trays 20 by 20 by 1/2 in. high having wall thicknesses of 1/16 in. Calibration tests were made preliminary to radiation field mapping. (T.R. H.)

The advantages of the use of organic liquids in fission reactors to minmize corrosion and pressure problems were studied relative to the solution of thorium in such fluids. Thorio-organic compounds were prepared from organic acids, diketones, and other chelating compounds. Salts of carboxylic and phospho- organic acids were insoluble. The chelate with dibenzoylmethane was soluble in molten biphenyl but was decomposed at 300 deg C. The general low solubility of thorio-organic compounds in nonpolar solvents can be explained by steric effects. The large thorium atom has the ability to form strong coordination complexes with adjacent molecules, leading to coordination polymers. The effect can be minimized by shielding the thorium nucleus with large organic groups such as dibenzoylmethane. The large, branched organic groups needed to impart solubility limit the maximum solubility. (auth)

Cobalt samples frequently are irradiated in nuclear reactors to produce gamma sources and can be irradiated as integral flux monitors because of the long half-life of the isotope produced. At the present time a small cobalt sample is being irradiated within the KW Reactor Snout facility for future use as a radiographic source for inspection of finished product in the Chemical Processing Department. Analysis was made to estimate the buildup of activity in this sample; the general equation may be of interest and value for other cobalt sample irradiations.

The hydrodyraruios code BOCH and the physics code PUREE' were used in the anaiysis of specific VMR cores. Eight core designs of different geometries were selected for initial evaluation. Extensive analysis of the first of these cores was completed. aid more general analysis of the other cores is in progress. The analog representation of the VMR kinetics was completed and applied to the aaalysis of the VMR over a large range of parameters for curves of positive void versus reactivity. Funiner checks of the first three energy groups of PUREE' against experinnental data were completed. The P/sub 3/-POP method of computing the thermal group constants was used in conjunction with the first three groups to analyze the reference core designs and the critical experiment. The more complex blackness method of computing the PUREE' thermal group constants appeared to be functioning properly and is being checked on very tight lattices to determine the magnitude of the change in fuel-pin blackness due to incident neutron distortion by neighboring fuel pins. Mechanical and system design of selected cores was started. Corrosion rate of the moderator systems was computed. and preliminary analysis of a clean-up system was started. (For preceding period see ATL-A-103.) (W.D.M.)

In developing a method of numerical analysis for the solution of thermal- stress problems special emphasis was given to fuel elements with internal coolant channels. Numerical techniques ior reducing the partial differential equation system te a form suitable for numerical solution and a new iteratlve method of solving large systems of linear algebraic equations were employed. Computer codes were devised to obtain the numerical solution of the thermal-stress problems and were used to obtain numerical results for single-hole and seven-hole hexagonal elements and plate-type elements. Comparisons were made between analytical results and numerical results for the case of t:.ie simple annulus shape. (auth)

A study was made to determine the optimum arrangement of the fringe flow zones at KE Reactor. Also considered, was the possibility of converting part or all of the existing fringe zone (actually three flow zones) from solid metal to I & E metal to decrease rupture potential in these low flow zones. The necessity for this study was indicated by high tube outlet temperatures in the fringe flow zones and the recent occurrence of two solid metal ruptures in process tubes located in the fringe zone of the reactor. Two additional solid ruptures occurred during the period that conversion to I & E in zone 2, as recommended below, was being completed.

Temperatures were calculated relative to the storage of radioactive solid waste as a function of time and radial distance for radioactive solid cylinders in infinite solid media of "average soil," "average rock," and salt. A resistance at the cylinder--infinite medium boundary was included in the form of an air space. For the range of parameters used and withia the practical limits of accuracy, the maximum temperature rise increased linearly with the heat generation rate. The fission product spectrum was not significant in the determination of the maximum temperature rise. Under the pessimistic storage conditions assumed, the storage of cylinders of a practical size appears feasible without excessive temperature rise. A maximum temperature rise of 1000 deg F would be produced with an initial heat generation rate of 1300 to 1600 Btu/hr-ft/ sup 3/ for cylinders with a 5-in. radius, with 350 to 450 Btu/hr-ft/sup 3/ for a 10-in. radius, and with 175 to 210 Btu/hr-ft/sup 3/ for a 15-in. radius, assuming a thermal conductivity of the radioactive cylinder of 0. 1 Btu/hr-ft- deg F. (auth)

Some of the controllable variables have been investigated in casting of lead alloys having 0.06 to 0.09 wt.% Ca. The alloy has been found to respond to solution heat-treatment, room-temperature precipitation hardening, and accelerated hardening at 100 ts C. Tensile strengths to 7000 psi have been obtained. (auth)

Steady state hydraulic demand curves were obtained for tube powers of 500, 1000, 1500 and 2000 KW with an inlet water temperature of 20C and a rear header pressure of 25 psig. These curves are shown in figures. The point of initial unstable flow for various tube powers is shown for a front header pressure of 325 psig. The flow rate that would lead to the initial point of unstable flow as a result of a sudden plug upstream of the Panellit tap is shown in a figure.

It is recommended that the ``square pile`` concept be adopted for all disaster total control calculations, and that the basic reactor constants listed in HW-62884, except for Ball 3X local strength at the DR Reactor, be used in applying this method. Curves are included for each reactor type, indicating allowable enrichment based on appropriate local control strengths. (The reactors whose operating methods are affected by disaster total control requirements are B, D, F, and DR Reactors; the remaining piles have sufficient geometrical coverage). An example of the analytical method is included.

Alloys containing 55 wt.% iron--22 wt.% nickel-- 17 wt.% chromium--2.5 wt.% molybdenum- 1.0 wt.% niobium-0.03 wt.% carbon-- 0.5 wt.% manganese-- 0.5 wt.% silicon with nominal additions of from 0 to 1.5 wt.% yttrium, and 36 wt.% iron-37 wt.% nickel--18 wt.% chromium--2.5 wt.% molvbdenum- 1.5 wt.% niobium-- 1.0 wt.% aluminum-0.05 wt.% carbon--0.5 wt.% manganese-- 0.5 wt.% silicon with nominal additions of from 0 to 2.0 wt.% yttrium, were prepared by vacuum- induction melting. Alloys containing 55 wt.% iron were successfully forged in air at 1900 deg F, rolled at 1850 deg F to 0.060-in. sheet and cold rolled to 0.015in. sheet. Fabrication of alloys containing 36 wt.% iron with more than 0.5 wt.% yttrium was unsuccessful. Addition of yttrium had relatively no effect on the yield and ultimate strength from room temperature to 1850 deg F. The ductility of fabricable alloys studied was increased at elevated temperatures by increasing yttrium contents. The greatest increase in ductility occurred at 1.5 wt.% yttrium. (auth)

An evaluation of a modulator flux sensitive playback head to be used with an analog magnetic recorder for nuclear pulse information storage was made. The modulator head was found unsultable. A pulse stretching unit was constructed which takes pulses from a standard linear pulse amplifier and provides an output signal suitable for driving a recording head. (For preceding period see ARF-1151- 3.) (C.J.G.)

Core dimensions and nuclear parameters were calculated for a sodium- cooled fast reactor containing UC fuel elements in a skewed hexagonal array. (C.J.G.)

None

As part of the coolant-core evaluation studies in an early phase of the Maritime Gas-Cooled Reactor program, nickel. K-Monel, niobium-1 wt.% zirconium, and graphite were irradiated in a convective flow of helium. A capsule of novel design was used to study the corrosion of the metals at 1500 and 960 ction prod- F in a 243-hr BRR experiment. Four heliumfilled quartz tori, containing the specimens, were encapsulated in a stainless steel capsule. A convective flow of helium was maintained in the closed tori by heating one leg of each torus to 1500 and the other at 960-F. Zirconium foilwas placed in two of the tori to getter gaseous impurities. No attack of the nickel and niobium-1 wt.% zirconium was observed at either temperature. Metallographic examination did show a 1 to 2.5- mil attack of the K-Monel at 1500 ction prod- F in a torus containing no getter. (auth)

The thermal and total epithermal neutron flux distributions in the EGCR from the center of the core through the biological shield were calculated. The maximum values of the flux distribution are presented in graph form. Ordinary concrete was found to permit a thermal flux buildup, similar to that found in the graphite reflector. In heavy concrete there was no such buildup seen, since the thermal flux is attenuated through the complete shield. A Val Prod code using multigroup diffusion theory was used to calculate the neutron flux distributions. Calculations are shown. (M.C.G.)

The proposed increased water capacity for 100-K plants will increase the electrical load to be supplied. The load study showed that the capacity of the existing 13.8 kV system is adequate to carry the increased loads proposed for Project CG-775, while for the 5 kV system, an expanded power system is proposed. Likewise, the voltage regulation on the kV system bus will be excessive, and voltage regulators should be added.

Production of Pu nitrate, UO{sub 3}, and unfabricated Pu metal met schedules. Decontamination performance of Purex process continued below standard. The cerium-144 cask is being redesigned. A ``powered ferret``, for driving a scintillation counter through a conduit to monitor ground activity beneath waste storage tanks, is being designed.

Approximately 100 ternary and quaternary spongezirconium alloys were screened for structural and cladding applications in a natural-uranium-fueled heavy-watermolerated power reactor. The alloy additions studied included2 to 4 wt.% Sn, 0.5 to 2 wt.% Mo, and 1 to 3 wt.% Nb. The effect of 0.1 wt.% Fe and 0.05 wt.% Ni additions to the experimental alloys was evaluated. All compositions were are melted, rolled at 850 ction prod- C from a helium- atmosphere furnace, vacuum annealed 4 hr at 700 ction prod- C, and furnace cooled. Room- and elevated-temperature hardness measurements were used to estimate the tensile strengths of the alloys, while corrosion resistance was evaluated by 1000-hr exposures to static 300 ction prod- C water. (auth)

A study to improve the elevated-temperature strength of niobium by solloving has resulted not only- in greatly improved strengths at 1200 and 1470 deg F but also in the development of improved fabrication techniques for these alloys. The most important step in the fabric:ition procedure of niobium and niobium-base allows is the initial breakdown of the cast structure. The cast structure of 1.84 wt. 4 chromium, 3.21 wt.% chromium. 4.33 wt. ' zirconium, and 20.5 wt.% titanium-4.28 wt. = chromium allovs and unalloyed niobium was broken known by- forging ingots (protected from oxidation by molybdenum ciins) at 2550 deg F and rolling at 800 deg F. After the initiai breakdown of the cast structure, the alloy-s were coid roiied to a total of 95 per cent reduction with no difficulty .A second fabrication technique was employed for a second set of alloys. Unalloyed niobium and 1.29 wt. % chromium, 2.74 wt. 3 zirconium, 4.5 wt.% molybdenum, and 10 wt. % titanium-3 wt.% chromium alloys were forged and rolled at 1000 deg F to break down the cast structure and then cold rolled to 0.030-in. sheet. the sheet obtained by this technique showed moderate edge cracking. Tensite tests on the coid-worked …

The Criteria for Speed of Control must be met for all cases being considered in the Expansion Studies. These Criteria are dependent upon the ratio Volume of water/volume of uranium in the process channel, the reactor power level, and the rate at which the vertical rods are released and inserted into the reactor. For those cases wherein the graphite channel is overbored by 0.200 inch., the criteria can generally be met for the power levels under consideration since the ratio of Vw/Vu is maintained adequately small by the use of large diameter fuel elements. If the graphite is not overbored, high pressure drop across the fuel element section must be maintained in order to maintain small values of Vw/Vu, which results in a substantial increase in the front header pressure. A method is developed in this document which ties together and defines the hydraulic and physics characteristics of the fuel-process channel geometry necessary to satisfy the Speed of Control Criteria over a range of feasible operating conditions. This study assumes no graphite overboring and employs throughout a smooth bore zircaloy process tube having an inside diameter of 1.650 inches. The fuel elements are self-supported., and the inner hole size was …

This document details activities of the Fuels Preparation Department during the month of January 1960. (FI)

The radial temperature around a sodium reactor heater assembly submerged in water is calculated using a model of the heater cross section found by conformal mapping. Thermocouple readings were also analyzed. When the heat flux is 5 x 10/sup 5/ Btu/hr-ft/sup 2/, a radial temperature drop of about 680 deg C across the center of the thermocouple well is calculated and found to be within 6% of the experimental value. Since most of this drop is across the 0.001-in. helium gap between the heater and its sleeve, it is concluded that the thermocouple will have to be bonded to the sleeve for dependable reading of true sleeve temperature. Drawings of the heater assembly and thermocouple are given. (D. L. C.)

Selected components from a staianless steel nonregenerative heat exchanger removed from the U.S.S. Nautilus (SSN571) have been examined for evidence of stress-corrosion cracking. The examination was conducted on three baffle plates, all primary-face seal welds, and the tubes and the surrounding tube sheets of 7 out of a total of 35 tubes. Maximum operating temperature of the heat exchanger was 260 ction prod- F for the inlet primary water and 164 ction prod- F for the exit secondary water. Chloride content of the secondary water averaged 5 ppm, with a maximum of 15 ppm. Only one suspected stress-corrosion crack was found in the components normally in contact with the secondary water. The crack, 2.5 mils in depth, was located on a fully exposed tube. At the 95 per cent confidence level this represented a probability of finding cracking on fully exposed surfaces in 0 to 60 per cent of the rest of the tubes in the heat exchanger. Severe stress-corrosion cracking was found in tubes in the tube-sheet drain area at the juncture with the inner tube sheet at the inlet end, where secondary water seeped past the expanded tubes. Cracks were detected in five of seven tubes for a …