From Introduction: "This Monograph has as its purpose the compilation and analysis of the extensive radio refractive index data available within the Central Radio Propagation Laboratory of the National Bureau of Standards. This will be accomplished by presenting both tabulations of basic data for specific locations and charts for interpolation to any location."

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.)

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)

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.

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)

The neutrons in an operating nuclear reactor are generated primarily by the fission events which are taking place. The great bulk of these relatively high energy or fast neutrons are slowed down or thermalized by a series of elastic collisions with the moderator nuclei which comprise the bulk of the volume of the reactor core. Once slowed down, the neutrons diffuse through the reactor core until they are absorbed or eliminated by some other process. Each of these three groups of neutrons, i.e., the fast or source neutrons, the intermediate or slowing down neutrons, and the slow or thermal neutrons, has a characteristic energy distribution. At a constant power level or rate of fissioning, an equilibrium is soon established among the groups at any point in the reactor. If it is assumed that a smooth transition exists between the different energy groups, it is possible to evaluate the entire neutron spectrum at a point in the reactor by determining the parameters which characterize each of the three groups. This has been done in the F Reactor Quickie Facility using radioactivants.

The objective of this irradiation is to further verify the corrosion rate of tubular-type fuel elements under conditions of high specific power and central core temperatures. This fuel will be the inner tube only of an NPR fuel assembly. As in previous tests, this inner tube rupture will be used to further substantiate the rupture detection instrumentation that is being used in the development of the NPR. Previously unirradiated fuel will be used in this test. The reactor is to operate at full power during the test. Permission is requested for charging two tubular elements The top element will have attached to it a hydraulic mechanism for opening a defect in the outer surface of the tube. The second or bottom element, will be used as a heater element to maintain loop temperature.

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 …

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. …

A correlation between the optical densities if the Pb-Sn alloy system and UO2 with respect to Co^60 gamma radiation has been determined. This enables one to fabricate density standards of whatever geometry may be desired for one in the gamma absorptiometer by simply casting a Pb-Sn alloy of the proper composition to correspond to the density required.

The following report describes the injection and trapping of high current electron beams in order to construct an electron gun and the first 2 Mev section of the accelerator.

The probabilities for nine electronic transitions among the low lying excited states in the C2 molecule are calculated by the dipole moment operator method and are given in the form of oscillator strength (or f values).

During normal operation the NPR will function as a closed system and the coolant will be maintained at a high degree of purity. Periodically, however, the system will have to be opened for fuel element discharge and/or decontamination. During these periods the reactor will be operated on single pass cooling. The use of deionized and deoxygenated water as the single pass coolant is not economically feasible due to the large quantities of coolant that will be required. At present it appears that filtered water will be used for this purpose. Since filtered water contains substantial amounts of dissolved solids (approximately 100 parts per million), and since it is saturated with air, this type of operation will have to be investigated with regard to corrosion problems. A particularly serious problem will be present after the system has been decontaminated, since the high temperature oxide films will have been removed and the bars metal surfaces will be exposed to the coolant.

The primary purpose of a cleanup system in a recirculating water loop is to maintain the best possible water quality conditions. This is normally accomplished by continuously purifying all or a portion of the coolant. A secondary objective of the cleanup system is to help maintain the system pH at a constant value. A system that will satisfactorily accomplish both of these objectives is at times difficult to obtain. Generally the pH control characteristics are sacrificed in favor of the more important cleanup requirements. A somewhat new approach to the problem pf cleanup system design appears to offer a solution to this problem for high pH systems.

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.

In some engineering applications it has become necessary to operate equipment containing small diameter rods or tubes oriented parallel to flow stream. In the case of several nuclear reactors such as the Plutonium Recycle Test Reactor and the S. S. Savannah Maritime Reactor, bundles of small diameter rods are used as the fuel elements. The situation also has application to heat exchangers. A method for accurately predicting the pressure drop characteristics of various bundle configurations would be helpful in the design and selection of equipment. Some progress has been made toward gaining a greater knowledge of the bundle characteristics under low temperature conditions. This report includes the development of relationship for the effect of temperature on pressure drop.

Steady state hydraulic demand curves were obtained for tube powers of 500, 1000, 1500 and 2000 KW with an inlet water temperature of 20C and a rear header pressure of 25 psig. These curves are shown in figures. The point of initial unstable flow for various tube powers is shown for a front header pressure of 325 psig. The flow rate that would lead to the initial point of unstable flow as a result of a sudden plug upstream of the Panellit tap is shown in a figure.

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

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)

Reactivity measurements made by the pulsed neutron technique were compared with results obtained by conventional techniques. The pulsed neutron results were in good agreement with those obtained by stable period measurement and rod drop. Differential effectiveness of partially inserted rods was shown to be well represented by elementary perturbation theory. Finally, the pulsed neutron technique was found to be the only good method for measurement of large reactivity changes.

Vanadium group beryllides were prepared and studied by x - ray powder diffraction and crystallographic methods. Properties included phase studies, sintering studies, vapor pressure measurements, and thermal diffusivity measurements,

The concept of thermonuclear explosives as a potentially cheap and almost inexhaustible energy source for mankind's non military needs has for several years been under active consideration at the Lawrence Radiation Laboratory. Many of the proposed peaceful applications involve underground nuclear explosions, and several experiments at the AEC Nevada Test Site have provided valuable insight into the phenomenology of such explosions. Among the possible uses currently under consideration are excavation, heat production, isotope production, mining, recovery of oil from shales and tar sands, improvements of ground water supplies, and the construction of earth fill dams. In addition a program of experimental research in the laboratory and in the field is under way. Sometime in 1961 Project Gnome if approved will be conducted in New Mexico. The purpose of Gnome, a contained nuclear explosion in a salt deposit, is to study the feasibility of heat recovery and isotope production, neutron scattering experiments will also be included. Other proposed nuclear projects will involve the creation of a small harbor near Cape Thompson, Alaska as the result of an experiment designed to investigate the cratering effects of nuclear explosives; a proposal to investigate the recovery of oil from Canadian tar sands using thermonuclear …

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 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)