Equilibrium krypton and xenon adsorption isotherms were obtained for four varieties of charcoal and for Linde Molecular Sieves Types 4A and 5A, generally at 0, 25, and 60°c. Such data are of interest in connection with design and evaluation of adsorbers for radioactive noble gas fission products. The isotherms were fitted, by linear regression analysis, to straight-line forms of the Freundlich and Langmuir equations. The Freundlich linear equation gave the better fit and the parameters of this equation are presented for each of the isotherms. Also presented are the constants for an equation representing the temperature dependence of arbitrary adsorption coefficients, the coefficients having been calculated from the Freundlich isotherm parameters. Some aspects of the applicability and accuracy of these results are discussed.

A proposed modification to the experiment off-gas system for the EGCR-LITR Capsule Experiment is evaluated by comparing the atmospheric radiation doses to laboratory personnel that could result in the event of an experiment failure while utilizing the present off-gas system to those doses that would result if various modifications to the existing system were effected. The modifications considered include the addition of an iodine vapor adsorber and a krypton-xenon adsorber. The addition of an iodine vapor adsorber to the off-gas system, which presently includes an absolute filter, reduces the atmospheric dose to the thyroid from 510 mr to 1.4mr. The total body dose is also reduced from 2.2 to 1.3 mr. The total body dose can be further reduced to 0.7 mr if a large krypton-xenon adsorber is included. Basd upon these calculated atmospheric radiation doses it is concluded that the addition of an activated charcoal iodine vapor adsorber will reduce the atmospheric hazard to an acceptable level.

A short-term study was made to investigate the removal of radioiodine vapor from air-steam mixtures as could occur immediately following a nuclear incident in a pressurized water reactor. Activated charcoal traps, designed to simulate a small section of a commercial charcoal canister, were tested at gas velocities of 23.9 to 74.9 ft/min over the temperature range of 75 to 118°C. The iodine removal efficiency was found to range from 99.80 to 99.4% with an average of 99.9%. One test performed at a temperature of 105°C and with gas velocity of 290 ft/min indicated that the iodine removal efficiency was reduced to 99.54%. Activated charcoal exhibits a high efficiency for iodine vapor removal under these test conditions and appears suitable for application in the decontamination of air-steam mixtures.

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).

After several failures had occurred in the HRT leak detector system, several lengths of this tubing were removed for metallurgical examination. The tubing was of type 304 stainless steel and was 1/4" in diameter with a 0.065 wall. The tubing had been purchased as three different lots, the first in 45 ft. lengths and the other two as standards lengths. Tubing from the first lot was used primarily for the shield penetration and, therefore, sections of it are present in all lines of the system. It appears that chloride contamination entered the system in a portion of the first lot of tubing used for the shield penetration. The exact source of the chloride cannot be determined, but after considering the results and visiting the manufacturer's plant, it appears most likely the contamination was during the manufacturing process.

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 design of internally cooled electrical coils for the production of high frequency intensity magnetic fields presents many new aspects and combinations of the familiar modes of heat transfer. However, the customary methodology appears to be sufficient for preliminary analysis and understanding of those problems. This methodology comprises the derivation of a qualitative, approximate equation expressing the relative performance of the various parts of a system, followed by an examination of this equation in order to locate the limiting features of the system. These features are then investigated by more powerful methods, which in turn provide guidance for development research in the laboratory.

The Eyewash program, written by James H. Alexander and Nancy D. Given, provides solutions of reactor criticality problems in spherical geometry by means of the group diffusion method. It employs thirty lethargy groups (plus one thermal group) in nine regions. The input consists principally of specifying the geometrical scaling factor, boundaries and compositions of the various regions, and temperature level. The output includes the value of vc that would render the system critical, the relative fission density distribution, fissions, absorptions, and leakages in each lethargy group in each region, and, if desired, an edit of the flux at each space point, each lethargy, and an edit of the macroscopic cross sections for each lethargy, each region. OCUSOL-A is a minor modification and extension of Eyewash. It provides for the computation and editing, on the supervisory control typewriter, of the total absorptions in selected nuclides in the various regions. This information is useful in the computation of breeding ratios and the preparation of detailed neutron balances, and in the estimation of flux-averaged cross sections for use in estimating the rate of change of concentration of the various nuclides with burn-up. The program also provides for saving and transferring the final fission …

The performances of aqueous-homogeneous (AHBR), molten-salt (MSBR), liquid-bismuth (LBBR), gas cooled graphite-moderated (GGBR), and deuterium-moderated gas-cooled (DGBR) breeder reactors were evaluated in respect to fuel yield, fuel cycle costs, and development status. A net electrical plant capability of 1000 Mwe was selected, and the fuel and fertile streams were processed continuously on-site.

The performances of aqueous-homogeneous (AHBR), molten-salt (MSBR), liquid-bismuth (LBBR), gas-cooled graphite-moderated (GCBR), and deuterium-moderated gas-cooled (DGBR) breeder reactors were evaluated in respect to fuel yield, fuel cycle costs, and development status. A net electrical plant capability of 1000 Mwe was selected, and the fuel and fertile streams were processed continuously on-site.

Oscillations have been observed in a magnetically supported cylindrical rod of plasma. This rod of plasma can be the discharge occurring in the defining aperture of a Mode II, pressure gradient arc. Similar oscillations can also occur in the column of a Mode I arc. These oscillations appear to be the mechanism that drives the Mode II blowup phenomena.

The in-plie slurry loop work is now being considered as a joint program between the PAR project and ORNL. It is proposed that PAR design, fabricate and test the in-pile loops and that ORNL operate the loops in-pile, dismantle the loops after irradiation and made the appropriate measurements to determine the radiation effects. This report gives the objects of the slurry in-pile program and outlines the facilities and operations required to execute ORNL's part of this program.

Calculated data and graphs describing the effects of batch thermal-neutron irradiations on the buildup of fission products in natural uranium, enriched uranium, and thorium are presented together with empirical equations and plots correlating total fission product activities and/or decontamination factors. Fluxes of 1012-1015 are considered.

The radiation hazards associated with recycled nuclear reactor fuels will greatly complicate the handling and refabrication of these fuels. This problem is most serious with U-233 and plutonium fuels where the presence of U-232 and the heavier isotopes of plutonium contribute energetic alpha, gamma, and neutron radiations at levels many times that from isotopically pure U-233 and Pu-239. This report summarizes present knowledge of the radiation hazards associated with recycled fuel and the additional data needed to make a thorough evaluation of these hazards.

Available knowledge of precision limits in S.S. accountability measurements and/or calculations by reactor and chemical processing groups is surveyed and summarizes. Experience in comparisons of reactor (production and research) calculations versus chemical plant accountability measurements is also reported. A general tentative conclusion is that available precisions (+/- 0.54 to +/- 0.78% ) in chemical plant measurements is also reported. A general tentative conclusion is that available precisions (+/- 1.0 to +/- 11.0%) possible by calculations (nuclear and/or engineering) of power reactor systems; however, with operation and empirical experience (e.g. less than +/-1.0%

An evaluation of the relative health hazards of radioisotopes produced in nuclear reactors is reported. The most important hazards were indicated to be I131, the Sr90 - Y90 chain, the Ce144 -Pr144 chain, Sr 89, the Ba140-La40 chain, Y91, the Zr95-Nb95 chain, Pr143, La140 , and Pa233. The most critical body organs affected by air-borne contamination are the thyroid gland, the bone marrow, the lungs, and the gastrointestinal tract. Where possible, continuous daily removal of gaseous and solid fission products from the reactor environment can be shown to permit very significant reductions in the total hazards. Homogeneous reactors, such as the Thermal Breeder Reactor and the Homogeneous Plutonium Producer Reactor, specifically studied in this report, are designed with daily removal cycles and may be considered potentially safer than heterogeneous reactors.

Activity and ionic impurity specifications are presented for Thorex thorium nitrate products. Two sets of specifications are given, one set for direct handling during refabrication of production reactor thorium metal slugs and the second for refabrication of future power reactor thorium metal elements by semi-remote technics. Consideration was given to the health hazard problems associated with each process step between the Thorex process and final refabricated source material in order to arrive at these specifications.

A program has been coded for the ORACLE which will solve, using Monte Carlo technique, the transport problem for monodirectional, monoenergetic gamma radiation incident at an angle Θ, upon an infinite laminated slab of finite thickness. Each of the laminations (or regions) is itself an infinite, homogeneous slab of finite thickness. The code is designed to give estimates of energy deposition, energy flux, tissue dose rate, reflected and transmitted energy current, and the angular and energy distribution of the reflected and transmitted energy current. All the answers except for energy deposition and reflected and transmitted energy current are optional.

The thermal-neutron flux prevailing near the upper hole in the HRE-2 core wall was determined from measurements of the induced activity in a specimen cut from the edge of the hole. The specimen was removed following run 21 and was thus exposed to reactor neutrons during runs 13, 14, 16,17, 18. 19, 20, and 21. The neutron flux, calculated for an average core power level of 2 Mw, was 2.8 x10^13 neutrons/cm^2 sec. the core wall area from which the specimen was removed was located approximately 16-in. from the core center.

A fifth in-pile loop experiment, L-4-8, was completed. The loop operated in-pile for a total of 1637 hr, during which time the LITR energy output was 4377 Mwhr. The average fission power in the loop based o cesium analyses was 622 w when the LITR was at full power (3 Mw). Based on oxygen data, the generalized corrosion rate for the first 300 hr was 4.0 mpy; the rate for the remaining 1357 hr was 0.7 mpy. The nickel data gave parallel results. The corrosion of the type 347 stainless steel, Zircaloy-2, and Ti-55AX [unintelligible] exposed in the core and in in-line holders was generally consistent with that observed in previous in-pile loop experiments. Some differences with steel were attributed to the fact that this was the first loop containing steel specimens operated with 0.04 m H2SO4 present in the uranyl sulfate charge solution (0.17 m UO2SO4, 0.03 m CuSO4). Stress specimens, made from the alloys Zircaloy-2, type 17-4 PH stainless steel, and Ti-C-130-AM, were exposed in care, in-line, and pressurizer locations. Microscopic examination and average weight loss gave no indication of effects attributable to the stressed condition of the specimens.

Results of the examinations of corrosion specimens exposed in the HRT are presented. Specimens examined included (1) core screen samples, (2) core specimen array No. 1, (3) blanket specimen array No. 2, (4) core solution line specimen arrays No. 103 and 103A, (5) blanket solution line specimen array No. 203. Complete information is still not available on all the specimens removed from the reactor, however, those data which have been accumulated are presented. These include corrosion rates, computed from specimen weight-changes, and results of chemical analyses of scales removed from the specimens or specimen holders. Also included in the report is a summary of the HRT operating schedule during exposure of the specimen and of the status of examinations for those specimens removed from the reactor prior to run 20.

The radiolytic decomposition of tributyl phosphate has been show by burr to product gases, acids, and polymers. the acids that are formed can limit the usefulness of tributyl phosphate for the processing of radioactive materials. The study of acid formation on radiolysis therefor becomes important to us for the understanding of the mechanism of acid formation, the influence of added materials and the effect of the structure of organo- phosphorous compounds. Samples of purified esters have been irradiated in the ORNL cobalt source. Conclusions: (1) The formation of acid from tributyl phosphate during irradiation with cobalt gamma rays involves a mechanism that is not a primary step. Evidence has been obtained for a secondary reaction that is minimized by the presence of certain compounds. (2) Toluene acting as an inhibitor for radiolytic formation of acid may indicate the usefulness of an aromatic hydrocarbon diluent in processing. (3) The aromatic hydrocarbon, when phenyl, was not effective when it was part of the molecule. (4) First tests revealed that dibutyl butyl-phosphonate produced less acid the TBP and may warrant further consideration for process application.

A stimulation study of the deign as of June 1960 of the gas-cooled ORR Loop No. 1 was made using the ORNL analog computer. The proposed method of temperature control is evaluated, and the dynamic behavior of the loop for accidents and component failures is presented in graphical form.