The technical bases for Phase I of the Fission Product Program have previously been presented for Phase I design. This report represents its sequel for Phase II of the program. Phase I provides the means of segregating, concentrating, and aging crude fission product fractions with market value. Phase II is to provide the means for single-line purification and packaging of megacurie quantities of fission products (principally strontium-90). The technical bases for Phase II project scoping studies are presented herein in the form of process flowsheets and tabulated data. Equipment needs are also described. Conceptual processes and flow diagrams for Phase III are presented for the Waste Management Program and the Fission Product Program.

Abstract: This technical report presents design criteria to prevent the brittle fracture of ferritic reactor vessels that cold occur as a result of the rise in NDT caused by fast neutron irradiation. The criteria require that maximum principal stress in the vessel does not exceed 18 percent of yield stress at temperatures below NDT + 60 degree F. Under certain conditions the allowable stress may be based on the irradiated yield stress. A discussion of brittle fracture and an explanation of the criteria are included.

Low flow calibration tests have now been completed at all old reactors as a part of the program to determine the adequacy of the old reactor last ditch water backup system more accurately. In addition, the high tank discharge line strainer gates have been removed at B, D, and F reactors in order to increase the emergency high tank flow rates to the reactors and thereby increase high tank water backup adequacy. The purpose of this document is to list the results of the last two low flow calibration tests at DR and C reactors (calibration test results at B, D, F, and H reactors are listed in reference 7) and to list the increased high tank flow rates that were obtained at B, D, and F reactors when the strainer gates were removed.

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The purpose of the production test was to {open_quotes}investigate the reactivity and temperature effects and the associated operating problems resulting from the use of nitrogen instead of carbon dioxide as a constituent of the K Reactor atmosphere.{close_quotes} This report summarizes the reactivity and reactivity-associated graphite temperature effects observed during the first four months of the test which was initiated at KE Reactor on December 5, 1960.

Information on small power reactor projects of the USAEC is summarized. General information concerning the projects as a whole is given. Specific projects discussed include: the Elk River Power Reactor, the Piqua Nuclear Power Facility, the BONUS Power Reactor, the Pathfinder Power Reactor, the small-size pressurized water power reactor, and the experimental low-power process heat reactor. (M.C.G.)

This report addresses the design properties for graphite.

At the request of IPD Personnel, calculations on neutron and gamma attenuation were made for the NPR shield. The calculations were made using a new shielding computer code developed for the IBM 7090. The calculations show the thermal neutron flux, total neutron dose rate, and gamma dose rate distribution through the entire shield assembly. The calculations show that the side and top primary shield design is adequate to reduce the radiation level below design tolerances. The radiation leakage through the front shield was higher than the design tolerances. Two alternate biological shield materials were studied for use on the front face. These two materials were iron serpentine concrete mixtures with densities of 245 lb/ft{sup 3} and 265 lb/ft{sup 3} (designated by I-S-245-P and I-S-265-P, respectively). Both of these concretes reduced the radiation below design tolerances. It is recommended that the present front face biological shield be changed from I-S-220-P to I-S-245-P. With this change the NPR shield is adequate according to these calculations. The calculations reported here do not include leakage through penetration in the shield.

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Laboratory results are presented for a flowsheet study of a process for decontaminating ORNL low level waste water. The water is adjusted to approximately pH 12 with NaOH, clarified, and passed through a bed of phenolic cation exchange resin. This study, using a phenoliccarboxylic resin, showed essentially the same results as those previously reported using a phenolic- sulfonic resin, i.e., radioactivities in ORNL waste were reduced to the order of 10% of MPC. The phenolic-carboxylic resin has the advantage that it can be regenerated with 0.5 M HNO/sub 3/ instead of the 5 M HCl required for the phenolic-sulfonic resin. Volume reduction factors of 2000 to 3000 were achieved. (auth)

Laboratory results are presented for a flowsheet study of a process for decontaminating ORNL low level waste water. The water is adjusted to approximately pH 12 with NaOH, clarified, and passed through a bed of phenolic cation exchange resin. This study, using a phenolic-carboxylic resin, showed essentially the same results as those previously reported using a phenolic-sulfonic resin, i.e., radioactivities in ORNL waste were reduced to the order of 10% of MPC. The phenolic-carboxylic resin has the advantage that it can be regenerated with 0.5 M HNO/sub 3/ instead of the 5 M HCl required for the phenolic-sulfonic resin. Volume reduction factors of 2000 to 3000 were achieved.

The large aqueous pilot plant facilities at ORNL were cleaned and are being put in standby condition. Experience was gained during the year as anion exchange was used to recover more than one kg of plutonium left in the exploded evaporator system. This experience is being applied to a new recovery program just beginning in cell 1 of Building 4507. (auth)

The results of a limited analysis of the extended SM-2 critical experiments are given. A review of the analytical models is presented to ascertain the accuracy of the reactivity calculations for the reference SMi-2 Core I. A new reference B-l0 loading was derived based on analysis and experimental measurements. Calculations of power distributions and burnup behavior of the reference core are presented and comparisons to available measured data made. (auth)

Available pressure data on the B and D downcomers and approach piping is tabulated herein. All readings were taken with an air bubbler system; readings given in psi were taken on a Bourdon-type pressure-vacuum gauge, readings given in inches of water were read on an oil-filled manometer. In each case, fluctuations occurred and the mean value was estimated by eye. Data are presented as a function of flow rate and temperature.

The Idaho Chemical Processing Plant for several years has been studying various pulsing systems, used in extraction columns for processing irradiated nucIear fuel elements, in an effort to eliminate the mechanical difficuittes in present pulsing systems. An air puising system has been demonstrated, and a discussion is given of experimental work, pulser design, and plant operational experience with this system. It is concluded, from work completed thus far that air puising of extraction columns is a considerable improvement over other systems. (auth)

The development of nuclear reactors and other sources of nuclear particles has provided the analyst with a new method which has been successfully applied to the determination of microgram and even submicrogram quantities of many elements. In this method, known as "radioactivation analysis," the element to be determined is "activated" through some type of nuclear reaction which produces a radioactive isotope of the element. Since the radioisotope produced decays with its own characteristic radiation and half-life, it is possible, through radiochemical measurements following radioactivation, to devise a method of analysis which is very specific for particular elements. Whenever necessary, the radioisotope is separated by chemical means and its radioactivity is then measured by some type of radiation counter.

This report, for August 1961 from the Chemical Processing Department at HAPO, discusses the following; Production operation; Purex operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; employee relations; and special separation processing and auxiliaries operation.

Activities in research programs are summarized in the areas of power reactor fuel processing, fluoride volatility processing, molten salt reactor fuel processing, homogeneous reactor fuel processing, waste treatment and disposal pilot plant decontamination, GCR coolant purification studies, equipment decontamination, HRP thoria blanket development, fuel cycle development, transuranium element studies, production of U/sup 232/, uranium processing, fission product recovery, thorium recovery from granite, solvent extraction technology, mechanisms of separation processes, radiation effects on catalysts, ion exchange technology, chemical engineering research, chemical applications of nuclear explosions, reactor evaluation studies, and assistance programs. (J.R.D.)

Report documenting the ongoing research and developments of the Chemical Technology Division of the Oak Ridge National Laboratory.

A development program was initiated under Project CG-775 to test the proposed. process water pump modifications for Project CGI-883, Increased Process Water flow 100-K Area. Both the high lift and low lift pumps of the 190-KE pumping Unit were modified during May, 1959 for in-service prototype testing of the proposed equipment revisions. The testing proved the feasibility of the modification program but a problem, in the form of capitation damage to the high lift pump impeller, was noted. An incremental production gain at the NE reactor was realized as a result of the test program. Interim funds were authorized in February 1960 to initiate detail design on total KE and KW water plant expansion and to modify one KW process paying set for an early incremental production gain at the KW reactor. The modified KW process pumping unit was placed in service in mid-1960 and a proportionate reactor production gain was realized. Shortly after the modified KW pumping unit was placed in service it was noted that the total process water flow in KW was somewhat greater than in KE. An investigation was made to evaluate the low lift pumping Units in each area to find out if they contributed …

The history of gas-cooled reactor developmert is reviewed. The 40-Mw HTGR r the Peach Bottom Power Plart is described. Other projects to follow this prototype are outlined. Fuel development and progress in graphite core technology are discussed. The program sponsored by the Empire State Atomic Development Associates for a high-temperature gas-cooled reactor with an output of 300 to 500 Mw(e) is described. (M.C.G.)

Spatial patterns of radioactive contuamination on forest foliage were measured by gamma spectrometry and are discussed with respect to local vs. world- wide origin of the fallout and implications for ecology, health physics, and management of nuclear facilities. In September 1959, I/sup 131/ on dogwood leaf samples varied from over 500 mu mu c/g dry wt near Oak Ridge National Laboratory stacks to 1 to 7 mu mu c/g near the margins of the Oak Ridge Reservation. Stack fallout tended to occur closer to the source than was calculated from hourly wind data by an IBM 610 computer program based on Culkowski' s adaptation of the SuttonChamberlain theory of atmospheric diffusion and deposition. Over most of the Reservation levels of Ru/sup 106/ Cs/sup 137/ Zr/sup 9/ >s/sup 5/Nb/sup 95/ and Ce/sup 144/ were similar to levels found elsewhere (2 to 9, 1 to 3, 2 to 9, and 10 to 20 mu mu c/g respectively) and were presumably controlled by weapons fallout. Higher levels were found in small areas and indicate the need for attention to localized contamination, even though indirect estilevels considered hazardous from the standpoint of health physics. (auth)

The purpose of this report is to present some experimental data concerning boiling burnout and to discuss briefly the significance of these data in relation to the Hanford production reactors.