Chloride-induced stress-corrosion cracking has been encountered in the Homogeneous Reactor Test during the preliminary testing. The rector is constructed of austenitic stainless steels. It is unique in that it will operate at 250 to 300 C with an aqueous uranyl sulfate solution fuel containing 200 to 500 ppm of dissolved oxygen. The cracking has occurred in a secondary system used for detecting leaks in the flanged joints of the primary systems and in the grooves of flanges in the primary systems. Tubing used in the leak-detection system was found to be contaminated with chloride introduced during manufacture.

A second test of the mockup of the Zircaloy - stainless steel transition joint as used in the HRT reactor vessel has been completed. The joint and bellows have now received 104 thermal cycles and 148 mechanical deflections. The joint and bellows have functioned properly; corrosion damage has been negligible, except for a small area on the bellows which has undergone pitting attack. Long-term runs with uranyl sulfate solutions of the concentration proposed or use in the HRT have shown the solution to be stable at 300 C. Substituting heavy water for normal water caused no difference in either corrosion or solution stability. Experiments in which chromic acid was used to pretreat stainless steel have shown that, under certain conditions, the pretreated film can exist in uranyl sulfate solutions at flow rates in excess of the critical velocity for relatively long periods of time. The practicability of using titanium inserts in high turbulent areas of stainless steel loops to minimize corrosion has been demonstrated. The corrosiveness of beryllium sulfate solutions containing dissolved uranium trioxide has been determined at 250 and 280 C. Laboratory studies with regard to stress-corrosion cracking have shown that high stressed type 347 stainless steel will crack …

A system for concentrating uranyl nitrate solutions was designed and installed in the Thorex Pilot Plant. A total of 16,060 g of uranium was concentrated in the system in 68 batch runs. A total of 14,400 g total uranium (14.180 g U/sup 233/) was recovered as product suitable for shipment. Uranium loss to the evaporator condensate was 0.03% of the total uranium processed. The material balance across the system was 98.4%. The average concentration of uranium in the evaporator feed solution was 29 g/liter; the average concentration in the evaporated solution was 298 g U/liter and in the product solution was 199 g/liter. Radiation readings of bottles containing product solutions were taken with a hard-shell cutie pie immediately after each run, and these readings ranged from 35 to 1100 mr/hr. The radiation levels of the bottles of product solution shipped averaged 78 mr/hr. Bottles of product solution reading in excess of 300 mr/hr, maximum allowable for shipment. were reprocessed in the second-cycle solvent extraction system (Thorex) and reconcentrated. The products from seven runs had radiation levels in excess of 300 mr/hr at the time of concentration, or the activities had grown to that level by the time of shipment. The …

The following temperature measurement tabulation consists of two parts. Part I lists all HRT thermocouples, their location, the junction box thru which the leads pass, and their termination, if on an instrument. Part II lists all temperature read out instruments and their location. A total of 577 thermocouples are listed in this tabulation. The roughly 77,000 ft of wire used in connecting them up cost $6,799. Temperatures are read on 24 instruments. Cost of these was approximately $15,688. Accessories such as patch panels, conduit, disconnects, etc., used in installing the thermocouples cost about $8,069. Total cost for material and instruments for temperature measure comes to approximately $30,556.

An estimate has been made of the gamma radiation levels at access holes in the HRT Shielding when the plugs have been removed to service or maintain the reactor. In every case the radiation level at the holes was greater than the maximum permissible exposure rate of 0.3 roentgens per week. The radiation through the holes can be attenuated to some extent by flooding the reactor cell up to the flange to be disconnected. However, shielding would still be required and it is more practical to provide a small additional shield thickness to compensate for the moderate attenuation that could be gained from flooding.

When it was discovered that two HRT flange bolts of a lot of 16 spares contained serious cracks, a program was launched to (1) determine the cause for the cracking, and (2) find methods for non-destructive testing the remainder of the 672 bolts shipment, a large portion of which had been installed in the HRT. Concurrently, inspection of 8 ferrules removed from an HRT flange revealed hairline cracking in 4 of them. Magnaglo, a magnetic particle inspection method using a fluorescent dye, proved to be the only definitive method for inspecting the bolts. The evidence gathered on the bolts pointed to quench cracking as the cause for the defects. Nothing abnormal was disclosed in regard to the bolt material. The alloy and heat treatment at present prescribed for the HRT bolts and ferrules are considered suitable. However, recommendations are made for plating with zinc, instead of formerly prescribed cadmium, to a thickness of 0.0002 inch, followed by a hydrogen relief treatment and a final bichromate chemical dip.

The horizontal and vertical power distribution for a 5 x 7 fuel element loading of the TSR is presented. (auth)

Short reports to the members of the Nuclear Cross Section Advisory Group from three groups: (1) High voltage group; (2) Fast chopper time-of flight spectrometer; and (3) Electronuclear research division.

An annual report is given on the operation and costs of waste-disposal facilities at ORNL laboratories and operating buildings in the Bethel Valley area. The operations of the hot-chemical and metal-waste systems, the process-waste system, and the radioactive-gas-disposal system which utilized the 250-ft stack located in the Radioisotope area are discussed. The miscellaneous operations which include the SS (source and special nuclear) material control, SS material recovery, off-shift service for research divisions, water demineralization plant operations, and hydrogen liquefaction are included. However, the disposal of cooling water from LITR, off-gases from the Hot Pilot Plant, and the ORNL Graphite Reactor building are not covered by the report.

This semi-annual report of the Applied Health Physics Section of the Health Physics Division provides both narrative and quantitative descriptions of area air monitoring activities, fallout activity, rain water analyses, Clinch River and White Oak Creek analyses, background radiation monitoring, and monitoring of the work and protective garments laundry. Measurements for gamma and beta activity are reported in millicuries/cubic centimeter. The report does not provide information on specific radionuclides.

This report is presented in order to provide a satisfactory system of symbols and identifications for process-instrumentation equipment and to promote a uniformity of practice that will simplify and expedite instrumentation work. It is intended that the systems presented here should be capable of designating and identifying the multitude of instrumentation items which are used for control and operation of conventional processes, as well as for specialized work peculiar to ORNL. Instrument Society of America standards have been adhered to whenever practical.

This report supersedes ORNL CF-57-2-1, which was an extension and revision of ORNL CF-54-6-72. Details concerning a recommended system of flow-plan symbols and drawing are given. The system is designed to identify the function of all major instrument components and to show schematically the operation of the instrument relative to the particular process. The system is used for identification and designation. The system is a modification of the Instrument Society of American Recommended Practice (RP 5.1).

To determine the maximum permissible concentration (MPC) of a radionuclide which may be ingested or inhaled with exceeding accepted levels of exposure tot he human body, it is necessary to know the average normal concentration of the natural element in the various tissues of the body. Any element may be ingested or inhaled as the element or as a compound. Although nearly all elements have been observed in living tissue, few quantitative observations had been made before 1950. Preliminary studies showed that individual variations in concentration of all elements were very wide and that geographical variations for certain elements occurred. Collection of tissues from autopsies in 9 cities has been done as of February 1957.

All tissues included in this report were received from Dallas, Texas and were collected, ashed and analyzed according to the methods described in ORNL-CF-57-2-2. The average value of the concentration for each element in a tissue is the average of the concentrations in only those samples of the tissue which contained the element. For example, if, of 25 lives analyzed, only 10 contained lead, the average for lead in liver is the average of those 10 values. For those tissues in which an element does not appear in every sample, the lowest value is reported as less than the limit of sensitivity of the method for that element in that tissue. For the elements, lithium and rubidium, the ash was a composited ash sample for each particular tissue. As an example, in lieu of analyzing 20 individual aortas, only on composited aorta as was analyzed.

All tissues included in this report were received from Miami, Florida and the majority are from Jackson memorial Hospital. They were collected, ashed and analyzed according to the methods described in ORNL-CF-57-2-2. The average value of the concentration for each element in a tissue is the average of the concentrations in only those samples of the tissue which contained the element. For example, if, of 25 lives analyzed, only 10 contained lead, the average for lead in liver is the average of those 10 values. For those tissues in which an element does not appear in every sample, the lowest value is reported as less than the limit of sensitivity of the method for that element in that tissue. For the elements, cesium, lithium and rubidium, the ash was a composited ash sample for each particular tissue. As an example, in lieu of analyzing 20 individual aortas, only on composited aorta as was analyzed.

The circuits described herein and shown functionally in Fig. 1 are to be added to the HRT control circuit to provide control and protection for the revised HRT cooling water system. The circuitry will provide protection against excess pressure in the demineralized cooling water loop and cooling water activity, will initiate action to insure containment of activity in event of an explosion and will provide emergency cooling water from the tower basin when required.

The measured pressure drop across the reactor core, with the sample holder in place, is 6.9 psi, more than twice the estimated value. Better estimates, based on more rigorous mathematical analysis, should be possible for future problems of this type. The 2% density difference which produced the relatively high velocity of approximately 1 fps, in this experiment, will result from a temperature difference of about 8 C. It is concluded that the bulk fluid temperature near the sample holder will be less than 8 C above the average temperature at the same elevation in the core.

The failure of the valve bellows would appear to be due to a combination of stress corrosion and crevice corrosion. Stress corrosion occurred as evidenced by the transgranular branched cracking found in the bellows and in the base which which was joined to the bellows. It seems probable that chlorides were present, which, along with the residual stresses present in the bellows assembly, created the necessary conditions for stress corrosion to occur. Crevice corrosion occurred probably due to heavy deposits of solids at the base of the bellows, which created a condition of oxygen impoverishment. While the crater in the base may have been related to a galvanic effect created by the gold gasket, the contour of the crater would suggest that the cause of the crater was due more to crevice corrosion.

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.

The fuel requirement of a heavy-water moderated, homogenous, power reactor were estimated for a variety of initial loadings, for both bath and continuous methods of fuel removal. This study considered a 12-ft spherical reactor, temperature 250 C, 500 Mw thermal power, 125 Mw electrical power capability, 0.8 load factor, and 4%/year inventory charges for U and D2O. The fuel shipping-and-processing charges were assumed to be $1/gm of fissionable fuel for the "batch" processed reactors, and $0.37/gm for the "continuous" processed reactors, Under these conditions, the minimum fuel costs associated with a 10-year 'batch" operating period were about 1.8 or 3.1 mills/kw-hr, if highly enriched U cost $15/gm or $20/gm, respectively. the analogous costs for the "continuous" processed reactor were about 1.6 and 2.6 mills/kw-hr, respectively.

Blake and his co-workers have shown that uranium and other elements can be extracted from acid solutions by various type of organo-phosphorous compounds. Early investigations in the laboratory have demonstrated the applicability of tri-n-alkyl phosphine oxides to the extraction of metal ions from acidic solutions for analytical purposes. This paper is concerned with a similar qualitative investigation of the extraction of metal ions with a di-alkyl phosphoric acid, di-2-ethylhexyl phosphoric acid (D2EPHA).

Lewis in 1929 determined the thermal conductivity of beryllium metal for the first time. The next measurement reported in the open literature on the thermal conductivity of beryllium was by Powell in 1953. Measurements were also made by Grenell, Linebrink, and Johnson at Battelle Memorial Institute in 1947. Their values are in good agreement with those of Powell.

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.

An order of magnitude estimate was made to determine a minimum cost for 10% pure oxygen-17 when produced by a chemical exchange process. the calculations were based on separations factors of 1.03, 1.01, and 1.003. the cost of product was found to vary from $23 per gram for the large factor to $165 per gram for the smaller.