A helium coolant purification system was designed for the proposed 800 MWT Pebble Bed Reactor. The purification system will operate on a coolant side stream with a flow rate 1% of the total coolant flow and there are provisions for radioactive and non-radioactive contamination removal.

The preparation of essentially pure dimethyl selenide for the Isotype Separations Group, Chemistry Division, is described. The compound was prepared by warming a mixture of selenium, sodium hydroxide, sodium formaldehyde sulfoxylate in aqueous solution for several hours at 50°C and then adding methyl iodide. Dimethyl selenide was removed by distillation.

In certain phases uranium processing it is poss!ble for operating personnel to acquire internal deposits of uranium. This body burden can be acquired by injection, as in contamination of a wound, by ingestion, of by inhalation. In order to estimate internal exposures, some means of determining the location and extent of these internal deposits is essential.

An improved heat sink for space vehicles was proposed in ORNL-CF-59-1-21. Subsequent work on the problem has indicated that there is a substantial probability of a puncture of such a radiator by meteors. To meet this problem a modified geometry has been evolved for which the probability of a meteor puncture should be reduced by a factor of at least 100 relative to the radiator of the original proposal at the expense of an increase in radiator weight of possibly 30%. This memorandum outlines the approach to the problem and a series of similar promising solutions.

A preliminary study of dispersing agents for thorium oxide has been completed and several of the dispersants have possible uses. Also many of the industrial dispersing agents tested are not usable with thorium oxide due to induced behavior causing balling and caking. The effects of nitric acid concentration have been observed to also effect each dispersing agent.

Portions of the NaOH-NaH phase diagram were studied by means of differential thermal analysis. Under certain conditions NaH will either react with NaOH according to the equation NaH + NaOH in equilibrium Na/sub 2/O + H/ sub 2/ or thermally dissociate according to the equation NaH in equilibrium Na + 1/2 H/sub 2/. Both of these reactions are suppressed by a high H/sub 2/ pressure; and NaH neither reacts nor dissociates to an extent sufficient to affect the results reported. This was evidenced by the fact that changing the H/ sub 2/ pressure above the system in the range indicated did not change, within the limit of error of the experiments ( approximately plus or minus 5 deg ), the temperature at which a phase change was started or completed. Therefore the concentrations of Na and Na/sub 2/O present in the samples must have been small in all cases. Under a high H/sub 2/ pressure, therefore, the system may be considered as essentially binary, consisting of NaOH and NaH. (auth)

Sodium hydride dissolves in and reacts with molten NaOH to give an equilibrium mixture of NaH, NaOH, Na/sub 2/O, Na, and H. In the case where there is a gaseous phase (hydrogen) and only one condensed phase, the system is defined by the temperature, pressure, and one composition variable. The equilibrium, H/ sub 2/ pressure, which is a measure of the H/sub 2/ activity within the melt, was determined as a function of the composition of the condensed phase(s) at 600, 700, and 500 deg for equilibrium mixtures with original compositions of 2.5 to 97.5, 5.0 to 95.0, l0.0 to 90.0, and 20.0 to 50.0 mole% NaH-NaOH. The equilibrium H/sub 2/ pressure-composition isotherms obtained by removing measured increments of H/sub 2/ were reproduced by reabsorbing H/sub 2/. Results for the 5.0 mole % NaH mixture were duplicated by starting with an equivalent quantity of either Na in NaOH or Na/sub 2/O in NaOH, and reacting with measured increments of H/sub 2/. The system is discussed in relation to the interdependent reactions involved, the phase rule, the thermodynamics of certain reactions, and experimental techniques employed. (auth)

A 4-ft-7-in. adsorption column packed with 1/8 in. sodium fluoride pellets was designed to reduce the hydrogen fluoride content of the fluorine being used by the Volatility Pilot Plant from 5% to less than 0.01%. It will be a non-isothermal packed bed with the bas inlet heated to 100 C to avoid plugging and the exit cooed to 25 C for more complete HF removal.

One of the considerations in designing a flat plate fuel element is the resistance of the fuel plates, especially the outer plates in the fuel plate array, to deflection and permanent deformation as a result of pressure differentials. An investigation was recently initiated wit the objective of obtaining preliminary information on the APPR-type fuel element to determine the effect of pressure differentials on the outer plates in the fuel assembly. The APPR-1 fuel element consists of 18 flat composite stainless steel fuel plates, joined to grooved 50 mil thick type 304L stainless steel side plates by brazing with Coast Metals N. P. alloy.

Small Heat Exchanger ORNL #1, type SHE #2, was removed form test stand B after 2071 hours of operation. Thirty-five samples were removed form the entire heat exchanger. The corrosion found on the outside of the tubes exposed to the fluoride mixture ranged to a maximum depth of .004 inches; however, the frequency of occurrence along the tube wall was heavier at the NaK outlet header, which was the hottest area in the heat exchanger. The depth of attack observed on the fluoride side of this heat exchanger was uniform from header to header and did not exceed .004 inches.

Loop L-4-12 was the seventh corrosion test loop operated in HB-4 beam hole at the LITR. The loop was inserted on January 24, 1956, and removed on April 17, 1956. This loop had a titanium core which was attached to the 347 stainless steel loop with special titanium to stainless steel couplings.

The fabrication of heat exchangers and radiators to be used in conjunction with high-temperature nuclear reactors may present exceedingly complex problems. Rigid heat transfer requirements may necessitate the use of compact assemblies of thin-walled small-diameter tubes as integral parts of the heat transfer units. Intricate designs may also be required in which cooling fins must be securely joined to the tubes at closely spaced intervals. In addition to the difficulties in fabrication imposed by the designs themselves, the high operating temperatures involved require the careful selection of materials and joining techniques. The choice of fabrication procedure for a given component must not only be based upon the stresses and temperatures to be encountered, but also upon special factors peculiar to nuclear service. Since many reactor applications employ highly corrosive environments, compatibility of the structural ma terials with the corrosive media is of paramount importance. The low nuclear cross-section require ment for brazing alloys to be used inside the re actor also places stringent limitations on the possible choices of in-pile applications. The use of boron in alloys for certain service may not be considered feasible, for example, because of its high nuclear absorption cross section. Although welding is used extensively …

Several plates of 98.7% Th - 1.2% U 235 (clad in aluminum) were irradiated in the MTR for an integrated flux of 2.6 x 10 21 neutrons/cm2. Although these samples represent an early development in bonding of aluminum to thorium and there are better methods at present, the bond proved to be quite strong and both clad and core were dimensionally stable under irradiation. The production of uranium 233 was as much as theory would indicate and the total amount of fissionable material material after irradiation and after decay of the protactinium 233 was greater than before irradiation. A fuel element of this nature appears to offer excellent potentialities from the standpoint of radiation stability.

This report presents data obtained from processing 33 Clementine Reactor fuel elements in the ORNL Metal Recovery Plant to recover approximately 15 kg of plutonium and 0.16 g of americium.

The proposed ORNL Research Reactor is designed to serve as a general purpose research tool delivering a maximum thermal flux of 8x10^13 n/cm2-sec at the initial power level of five megawatts. Operation at power levels up to ten megawatts is proposed for such items as sufficient cooling capacity is available to handle the increased heat load. The reactor will use MTR-type fuel elements and beryllium reflector pieces in a 7 x 9 grid with moderation and cooling provided by forced circulation of demineralized water. The reactor tanks are submerged in a barytes concrete pool, filled with water, which serves as a biological shield. Experimental facilities include two 18" diameter "Engineering Test Facilities" and six 6" diameter beam holes. In addition, access to the core is available through the water of the pool. The result on the surrounding population of release to the atmosphere of a large fraction of the radioactive material in the core has been computed by two methods. It is shown that under certain conditions off-area personnel could be subjected to greater than the maximum permissible exposure. An analysis of the maximum hazard caused by the release of the entire contents of the core to the local watershed …

This memorandum sets forth a recommended uniform basis for designing the ORN shield.This includes design values for power level and emergent radiation, standards values for various material properties, and basic radiation intensities.