The power and estimated reprocessing load are tabulated for existing and proposed United States and United States-built reactors of 10 Kw or greater thermal power.

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.

In the use of a reactor as a device to measure neutron cross sections, the pile oscillator is often employed because of its much higher sensitivity compared to that available in static reactivity measurements. The factor limiting the available in static reactivity measurements. The factor limiting the available sensitivity is the generation of random noise, which originates in the statistical nature of the chain reaction. In other words, the nuclear reactions which maintain the chain reaction do not proceed uniformly, but rather at rates which fluctuate in time around some average values. These fluctuations limit the accuracy of measurements in the reactor.

The measured diffusivity of uranyl nitrate in water at 25ºC was 0.7*10^e cm^2/sec with about 40% average deviation. A program was started to develop nonnuclear uses for depleted uranium. Two continuous DRUHM reaction tuns were terminate due to erratic operation of the sodium metering system. In the second Fluorox run with crude UF, which lasted for 29 hr, a total material balance of 94.8% was obtained and 17.9& of the theoretical amount of UF was collected in cold traps and chemical traps.

Annual report submitted by the Boy Scouts of America to Congress describing highlights from , activities, organizational leadership, and other information about scouting programs.

This is a preliminary report on the investigation of the N.S. Savannah replacement fuel rod vibration amplitude when excited by water flowing parallel to the axis of the fuel rod. Methods of calculating the fuel rod frequency in air were verified by experiment. Work is continuing to establish the validity of methods of calculating natural frequencies in water and to confirm the correlation of flow induced vibration amplitude found in reference (1).

This report summarizes the radiation and shielding analysis for a 30,000 SHP natural circulation boiling water reactor for ship propulsion. The reactor is proposed for installation in a 60,000 DWT, 18 knot tanker of the T-7 class.

The first series of tests in the KER-REP-1 loop proved that a fission product contaminated loop could be decontaminated to a safe level for contact maintenance. Since a good decontamination process was available, there was much that could be improved about this process. Further testing of this process and several variations of other processes have been scheduled. The evaluation of these processes includes specific decontamination factors, process corrosion, and loop activity reduction factors (loop decontamination factors).

Information was desired on applicability of seismic methods of geophysical exploration in helping to better define and refine hydrological and geological factors affecting ground disposal of liquid radioactive wastes. To date most of information on the area geology has been obtained from surface mapping and well drilling, although it has long been recognized that geophysical techniques may greatly assist in defining the geology of the region. Although direct measurements are generally preferred to indirect measurements have proven extremely valuable in many applications. Careful analysis of field data from geophysical measurements, and correlation with known geologic conditions, should provide significant knowledge of earth formations. Geophysical methods do not directly determine depth, type and composition of underground earth materials, but rely rather on detecting differences in physical properties (elastic wave velocity, magnetic susceptibility, density or electrical resistivity) of such earth materials.

This is a report of what may be an invention in the preparation of uranium dioxide of a quality particularly useful for the fabrications of fuel elements for nutronic reactors.

In considering the problems associated with the measurement of SSNM content of Non-Production Fuels, two distinct problems are apparent. The first is the technical problem of obtaining highly accurate measurements in a complicated chemical and physical system. The second is the administrative problem of choosing a measurement system which provides a basic of financial settlement acceptable to both the seller (reactor operator) and the buyer (Commission).

In early July, 1958, it appeared that Columbia River temperature at HAPO would be near 24.5ºC by the end of August. River temperature were averaging 4º to 5º above 1957 figures and were 3º to 4º above the ten year highs. It seemed desirable to examine the problem to determine if any corrective measure could be taken, since it was apparent that production losses were imminent.

Hastelloy F is a prime contender as a material of construction for plant dissolvers in the power fuel reprocessing program. Consequently, the fabrication and installation of dissolver was undertaken to delineate any unknown problems associated with the use of Hastelloy F; and, at the same time, to provide a vessel for development studies on the Niflex or the Sulfex processes. The purpose of this report is to describe the actual basis for design as well as to present the problems encountered during the fabrication of the vessels.

The purpose of this procedure is to present a detailed, chronological presentation of the preliminary decontamination and post decontamination steps necessary to fulfill the requirements of the Production Test Authorization IP-239-N. The procedure attempts to present the required operation in sufficient detail to successfully accomplish the intent of the test. Certain procedures involve operations of a standard nature and have not been elaborated upon to any great extent, as it is expected that the reactor operations and radiation monitoring personnel will implement these instructions according to standard operating procedures.

Nuclear safety in the dissolution of irradiated 0.95 U235 enriched fuel has been investigated. In particular, critical conditions of fuel of this enrichment in a 52-inch diameter dissolver crib were studied. Since a crib this size is not safe by geometry, dissolution procedures as well as maximum safe batch sizes were analyzed. Uranium-water lattices are normally studied in systems in which rods are uniformly dispersed in the moderator. The results of such a study for 1.34-inch diameter solid rods as well as I. and E. fuel having a 1.37-inch O.D. by a 0.48-inch I.D. have already reported.

The spatial distribution of thermal neutrons in a reactor lattices cell is of fundamental importance for many reasons. First of all, this information allows the determination of the relative absorption rates in each component of the cell which in turn enables a determination of the thermal utilization, and the diffusion length of the lattice. In addition, the observed spatial distributions of thermal neutrons in the lattice cell is of major interest in testing various approximations to the solution of the transport equation such as the P1 and P3 solutions.

This report describes the design and testing of a 1.17-Bev/c separated K{sup -} beam designed in the fall of 1958 in connection with a 15-in. hydrogen bubble chamber experiment.

Introduction: Seldom does the actual arrangement of control elements in a nuclear reactor confers to the ideal and convenient mathematical array. In order to achieve shim control. safety and regulation, it is desirable to design with rods of different sizes and materials. With given fuel element arrangement, typically in square or hexagonal lattice spacing, there will be rods located at different distances form the center of the core and from each other. As the reactor operates, absorbers will be withdrawn, leaving further asymmetries in the location of those remaining. It is the purpose of this report to develop in detail a two-group diffusion theory with as complete generality as possible. The method is as yet restricted to the unreflected core, or to the reflected core by use of reflector savings and bare equivalent geometries.

A major problem in the development of a pressurized water reactor coolant system for the NPS is the rupture performance of the fuel elements. As water temperatures are increased to 300 C, uranium corrosion rates increase rapidly. Swelling of the uranium fuel by corrosion could cause the process tube to burst or reduce the tube cooling water flow below acceptable limits. The desirability of slow cooling of the water to avoid thermal shocks to the reactor piping after a rupture is detected further complicates discharge and decontamination problems as fuel will continue to corrode with attendant fuel element damage during the cooling period. Coextruded uranium-zircaloy-2 clad fuel elements are scheduled for use in the NPR. The rupture behavior of this type fuel material after heat treatment was studied in ELMO-4, an out-of-reactor recirculating water loop. Several types of initial defects were studied. Fuel materials with five different heat treatment histories and with different types of defects were tested to determine their rupture behaviors. The five conditions were (1) as-extruded material as received from Nuclear Metals, Inc., (2) beta heat-treated and water quenched, (3) beta heat-treated and air-cooled, (4) beta heat-treated, isothermally treated at 600 C and air-cooled, and (5) beta …

It is the objective of this study to investigate preliminary expansion program requirements for process water, as supplied by 190 Building equipment; from the point of view of practical pumping, flywheel and pump suction head requirements. These requirements are to be determined at this time in such a form and accuracy as to be useful in refined estimating for budget study purposes. In order to obtain the objectives of this study at this time it has been decided to consider five different conditions of process water flow to a reactor. These conditions are flow to the reactor under summer conditions of operation in gallons per minute with a corresponding top of riser pressure in pounds per square inch: 85,000 gal/min, 580 psi; 100,000 gal/min, 480 psi; 130,000 gal/min, 280 psi; 150,000 gal/min, 280 psi; and 150,000 gal/min, 150 psi.

This document details activities of the Fuels Preparation Department during the month of April 1959. (FI)

Hot press canning of internally and externally cooled fuel elements has been developed to a point where the process is feasible. Complete specifications have been written for the process covering component, dies and punches, furnace construction, nickel plating, component cleaning, component assembly, sizing, hot pressing and inspection. Drawings covering each major item are included.

In early July 1958, it appeared that Columbia River temperatures at HAPO would be near 24--50{degree}C by the end of August. River temperatures were averaging 40 to 50{degree}C above 1957 figures and were 3{degree} to 4{degree} above the ten year highs. It seemed desirable to examine the problem to determine if any corrective measure could be taken, since it was apparent that production losses were imminent. The large storage of cold water behind Grand Coulee Dam, normally untapped, was a source of possible relief. A plan for use was proposed for the peak high temperature period and agreed to by the Bureau of Reclamation.

The temperatures of the Columbia River was reduced 1 to 30 Centigrade with beneficial effects at HAPO. It is reasonable to expect that future benefits may be possible. It is desirable that the temperature of the river be controlled each year to the maximum extent possible. Instrumentation improvements requested to effect optimum savings. Records of river temperatures and flows should continue to be maintained by IPD as a necessary part of temperature optimization. Where possible, the coincident use of the river cooling technique should be made for the benefit of anadromous fish.