Engineering Work Requests ESE-48 & 49 established a program for performance of Qualification Approval Tests on RPC`s 0.2 and 40 meghom Carbon Composition Resistors. This report is an accounting of these tests wherein (50) samples of each component were subjected to QA tests per drawings 8O2C250 and 802C251. Tests are explained and results tabulated.

The saturation voltage and current characteristics of a compensated ionization chamber (Q-1045) were measured with special regard to high current and voltage ranges. The chamber can be operated at currents up to I ma with a 2000 volt power supply. (auth)

Corrosion specimens of the five major aluminum alloys used in the construction of the Oak Ridge Research Reactor have been exposed to the high- purity primary cooling water in the ORR core and in the extennal portion of the primary cooling loop to determine their corrosion rates under actual operating conditions. These alloys, 1100, 3003, 5052, 5154. and 8061, exhibited average corrosion rates of less than 2.6 mpy during the first 500-hr test period and less than 0.5 mpy for a 4032-hr test. Very superficial pitting was observed. and no case of intergranular corrosion was found. (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)

A test was made to determine if there is sufficient excess reactivity in the core to override a peak xenon transient and to obtain dath for rod-worth calculations. After 149.8 EFPH's of plant operation on seed 2, there was sufficient excess reactivity to override peak xenon. Criticality during peak xenon was maintained with group III and group IV rods inserted, group I rods withdrawn to approximately -69 in., and group II rods withdrawn to approximately - 29 in. The override of peak xenon occurred 8 hr and 45 min after shutdown. (W.L.H.)

The Stationary Low Power Reactor No. 1 is a three Mw boiling water reactor designed to demonstrate the feasibility of a nuclear reactor to supply electrical power and space heat for remote sites. In addition to performance evaluation the facility provides training for military personnel. The reactor was operated for 1159 hr during the quarter for a total core burnup of 20.3%. Power generation was 103.7 Mwd for a total power accumulation of 466.9 Mwd. Eight malfunctions occurred during the quarter for a total unscheduled downtime of 40 hr 44 min, As a result of malfunctions, aluminum keys will be installed on the control rod drives, instrument well covers were removed, and an order has been placed for a station auxiliaries breaker with a higher temperature rating. Data were taken on seven tests during the quarter. Four of these are expected to be completed during the next quarter. All equipment items on order for the SL-1 power extrapolation expansion program are scheduled to be delivered before June 15, 1960. Although condenser dampers and damper controls, process instrumentation, equipment cabinets, and the radiation monitoring equipment have not yet been ordered, construction will not be delayed. The two tie-ins to the existing …

"Nuclear Science Series" on the National Research Council. Committee on Nuclear Science.

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)

S> Thicknesses of ordinary concrete required to shield the demineralizers and filters in the HFIR primary water system were computed for normal operating conditions and for abnormal conditions such as a meltdown of the fuel within the reactor. About 4 1/2 ft, 3 1/4 ft, and 4 1/4 ft of concrete are required to shield the cation exchange unit, the anion exchange unit, and the filter unit, respectively, to the most stringent of the following radiation levels: (a) 0.75 mr/hr for normal reactor operation or reactor operation with one defective fuel plate; (b) 1 r/hr immediately following the meltdown of 1% of the fuel; and (c) 1 r/hr 24 hours following a total fuel meltdown. Shielding thicknesses may be estimated for other tolerances from

The chemical vapor deposition of niobium, molybdenum, tungsten, chnomium, carbcn, and niobium--vanadium alloys in a fluidized bed of UO/sub 2/ powder particles wss used to provide uniform, dense, nonporous coatings on the individual particles. in the case of niobium, which received major attention, hydrogen reduction of niobium pentachloride vapor was used as the vapor- deposition reaction. The most serious problem was that of maintaining bed fluidity aad avoiding agglomeration. This problem was overcome to permit routine operation of the coating equipment. In the entire program of 68 experimental runs, only 1.1 per cent of the product was lost by agglomeration. In routine operation, this loss should be even lower. (auth)

In determining radionuclide sorption by clay minerals, shortcomings in the filtration technique for solid separation and in the contact times selected for testing were noted. Filters were found to have a high affinity for cesium and strontium when these elements were present in tracerlevel concentrations. Sorption equilibrium was not established in 24 hr, and the contact time was extended to 7 days. The affinity of the clay minerals illite, kaolinite, montmorillonite, and vermiculite for selected radionuclides was established after these shortcomings in the testing procedure were corrected. Illite exhibited a very high affinity for Cs/sup 137/ (K/sub d/ =200,000 after 7-day contact); none of the clay minerals demonstrated exceptionally high affinity for strontium (K/ sub d/ = 4000 for kaolinite, which was the highest measured value). The behavior of cobalt and zirconium-niobium was complicated by the colloidal nature of the elements. The uptake of the above radionuclides by a composited sample of Clinch River sediment showed that cesium and strontium behaved in accordance with the mineralogic character of the sediment. The sediment sorbed more cobalt than was accountable by the mineral composition; organic matter interactions with the cobalt are thereby suspected. (auth)

A description is given of the Advanced Sodium Graphite Reactor Power Plant, including the reactor, heat transfer systems, generatirg plant, control systems, and the economics of producing 256 Mw(e). The safety of this design is due to its unusually low operating pressure, absence of chemically incompatible materials in the core, and excellent stability under atatic and dynamic conditions. The reactor is being constructed at Hallam, Nebraska, at a probable cost of 0 to 0/kw, exclusive of the first core costs. The 151 fuel elements of uranium carbide are enriched to 2.75 at.% U/sup 235/ and clad in stainless steel. The average thermal neutron flux in the fuel is 8 x 10/sup 13/ n/cm/sup 2/sec. (B.O.G.)

Uniform, dense coatings of alumina about 5 to 150 mu thick were applied to uranium dioxide particles 44 to 350 mu in diameter by hydrolysis of aluminum chloride vapor in a fluidized bed of the particles at 1830 deg F. The coated particles were resistant to nitric acid leaching, to oxidation in 1830 deg F air, and to thermal cycling from 6OO to 2500 deg F. After low neutron exposures, the coated particles showed excellent fission-gas retention at temperatures up to 2400 deg F in inert gas. Although not optimized in the study, the coating process appears to have commercial feasibility. (auth)

Tests were conducted to assure proper operation of the control-rod mechanisms of the Shippingport Pressurized Water Reactor under normal operating conditions. (C.J.G.)

The design, fabrication, and installation of a fuel-element guide assembly for the Sodium Reactcr Experiment are described. Improvements in this assembly over other models are outlined, and recommendations for the testing of SRE core elements are presented. (T.F.H.)

The concentrations of and activities due to N/sup 16/ and N/sup 17/ in the HFIR primary coolant water were calculated. At the pressure vessel exit, the N/sup 16/ activity is 3.9 x 10/sup 6/ dis/secml and the N/sup 17/ activity is 6.9 x 10/sup 2/ dis/sec-ml. Comparison of the N/sup 16/ activity with the data obtained from the ORR water system indicates that the calculated results are slightly conservative. (auth)

A helium coolant side-stream purification system consisting of parallel sections for radioactive and nonradioactive decontamination was designed for the proposed Pebble Bed Reactor Experiment. Primary equipment components are two gas coolers, gas heater, charcoal delay trap, CuO oxidizer, molecular sieve adsorber, and full flow filter. The charcoal delay trap is sized to provide a hold-up of 30 min for Kr isotopes. 6 hr hold-up for Xe isotopes, and 99.9% retention of iodine isotopes resulting in "decontamination factors" varying from 1 for Kr/sup 85/ to 556 for I/sup 131/. Nonradioactive decontamination will result in a steady state concentration of CO/sub 2/ in the coolant of 20.8 ppm or less. Total cost of the system excluding auxiliary equipment and containment is estimated to be .690. (auth)

A wave equation for the pressure in a mixture of water and gas bubbles (such as found in the KEWB Reactor) is derived. The driving force may be either the production of additional bubbles or the growth of existent bubbles. Several illustrative solutions to the wave equation are found. (W.L.H.)

None

During the report period, the nature of the ORNL supporting activities gradually changed, reflecting the progress in the conetruction of the N.S. Savannah. Design reviews continued to require considerable attention. The Inspection Engineering Dept. of the Jab. continued to act as an inspection agency for the ALC, witnessing inspections and tests during fabrication of components of the nuclear reactor system. An enviromnental analysis was made of safeguard aspects of operation of the N.S. Savannah at the NYSC site in Cannden. A study of the safety response of the reactor on the ORNL Analog Computer further defined the important role of the Doppler coefficient in controlling reactivity excursions. Health physics aspects of the operation were studied. Limited waste disposal studies indicated that sea disposal of exhausted demineralizer resins may be facilitated by casting the radioactive resins into concrete. Installation of a pressurized-water in-pile test loop in the ORR neared completion. The neutron flux distribution in the loop was determined by experimental measurements in a nuclear mockup of the in-pile section. Metallurgical activities included nil-ductility testing of steel from the reactor vessel, chemical analyses of primary system components, and investigations of the properties of electroless - nickel brazed joints. Limited fabrication studies …

The formation of heavy negative ions by the attachment of low-energy electrons to oxygen molecules was studied for small amounts of oxygen mixed with methane or carbon dioxide. The rate of attachment in both cases was found to depend on the electron energy, the pressure of the oxygen and the non-attaching gas, and on the kind of non-attaching gas. In general, the attachment increases as electron enprgy decreases or as either oxygen or total pressure increases. The value of the attachment coefficient in oxygencarbon dioxide mixtures is about 100 times its value in oxygen-methane mixtures. This large difference is probably due in part to differences in electron energy and partly to differences in the stabilizing qualities of the two molecules. Dissociative attachment, which should be pressure independent, does not occur at the low energies that were used in this work. Both methane and carbon dioxide are to differences in the stabilizing qualities of the two molecules. Dissociative attachment, which should be pressure independent, does not occur at the low energies that were used in this work. Both methane and carbon dioxide are sometimes used as filling gases for Geiger and proportional counters. The high sensitivity of carbon dioxide to oxygen …

Lengths of pistillate inflorescences of sedges (Carex spp.) growing in the contaminated soils of White Oak Lake bed were measured in relation to radiation fields. Carex Frankii Kunth and Carex vulpinoidea Michx. populations were sampled from areas with air dose rates of 0, 10, 20, and 40 mr/hr. Analyses of variance showed that mean lengths of inflorescences of C. vulpinoidea were simllar to each other in these areas but that those of C. Frankii were significantly different. However, the sites differ in such factors as soil moisture, fertility, and alkalinity, so that these differences in the length of inflorescence may not be accounted for primarily by the exposure dose rate of the radiation field. (auth)

A study was made to determine the coolant flow required to prevent the temperature of the graphite moderator and reflector blocks from exceeding 1100 deg F at full-power operating conditions. The temperature distribution in the graphite blocks was also determined. Tha moderator was primarily cooled by coolant flowing in the 1/8-inch annulus formed by the fuel-assembly sleeves and the fuel-channel holes in the moderator blocks. Coolant flow in the controlrod channels and in the experimental-tube coolant annuli also cooled the moderator, and this effect was considered in setting moderator-annulus coolant flow. The coolant flows required for each of four zones in the reactor core were determined. The total moderator-annulus coolant flow (excluding 16 unfueled channels) was 15,200 lb/hr. The moderator-block temperature distributions for fullpower reactor operation after 20 years of operation are given. The maximum temperature (1100 deg F average over the cross section) occurred near the top of the graphite block. The temperature in the majority of the graphite columns varied from 600 to 1000 deg F over the lower half of the column and from 1000 to 1100 deg F over the top half of the column. The maximum graphite surface temperature was less than 1100 deg F. …

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.