The hazards study was made to determine the radiation level from the secondary sodium lines of the l0-Mw ESCR, and to evaluate the corresponding radioactive concentration in the secondary loop and the maximum permissible effective activation flux in the intermediate heat exchanger. The results are presented graphicallyn the radiation level from the coolant lines during normal operation as a function of line diameter and Na/sup 4/ concentration; the effective thermal neutron activation flux level in the intermediate heat exchanger as a function of the Na/sup 4/ concentration in the secondary loop and the ratio of the time spent in the flux field to the time spent in making one cycle; average Na concentration in the reactor room atmosphere for operative and inoperative ventilation systems, assuming that all the sodium in the secondary system is released to the room and burns; and the downwind concentration of Na/ sup 24/ resulting from the release of all the secondary sodium on an open pad area exterior to the reactor building. An analysis of the results shows that in the event of an accident the toxicological hazards are more severe than the radiological hazards. Recommendations are given for minimizing the toxological hazards. (B.O.G.)

None

The characteristics of Rankine space power plants in the zero gravity aspect of the environment of space were lnvestigated. The expected effects of Rankine space power plants are described. Discussions of experimental techniques for studying these phenomena show that this information can be obtained rapidly and economically. Recommendations for a program to supplement SNAP II and slmllar Ranklne space power development efforts in this vital area are made, and consist of: the development and testing of a small system that adequately simulates a complete Ranklne system, first in zero grayity and finally, in the complete orbltal environment; followed by, the development and similar testing of a complete Rankine system using SNAP ll hardware. (auth)

A device is described for the filtration of charged energetic heavy particles resulting in the production of adjacent or separated Bragg peaks within the range of the particles in tissue. Two or more layers of intense ionization at different depths separated by layers of less ionlzation in tissue can be produced. A cylinder of uniform ionization which cuts off sharply in depth in tissue can also be produced. (auth)

For presentation at ASCE Convention in Reno, Nevada on Thursday, June 23, 1860. A description of the Hallam Nuclear Power Facslity is presented. The history of the project, program participants, site description, component development program, reaetor building, reactor structure, reactor core, sodium systems, instrumentation and control, fuel and component handling, auxsilary sustems, special design features, and advantages of sodium graphite reactor systems are discussed. (M.C.G.)

A possible phase transition in liquid He{sup 3} has been investigated theoretically by generalizing the Bardeen, Cooper, and Schrieffer equations for the transition temperature in the manner suggested by Cooper, Mills, and Sessler. The equations are transformed into a form suitable for numerical solution and an expression is given for the transition temperature at which liquid He{sup 3} will change to highly correlated phase. Following a suggestion of Hottelson, it is shown that the phase transition is a consequence of the interaction of particles in relative D-states. The predicted value of the transition temperature depends on the assumed form of the effective single-particle potential and the interaction between He{sup 3} atoms. The most important aspects of the single-particle potential are related to the thermodynamic properties of the liquid just above the transition temperature. Two choices of the two-particle interaction, oonstituent with experiments, yield a second-order transition at a temperature between approximately 0.01 K and 0.1 K. The highly correlated phase should exhibit enhanced fluidity.

The magnitude of the energy gap in nuclear matter associated with a highly correlated ground state of the type believed to be important in the theory of superconductivity has been evaluated theoretically. The integral equation of Cooper, Mills, and Sessler is linearized and transformed into a form suitable for numerical solution. The energy gap, calculated by using an appropriate single-particle potential and the Gammel-Thaler two-body potential, is found to be a very strong function of the density of nuclear matter, and of the effective mass at the Fermi surface. It is concluded that the magnitude of the energy gap for nuclear matter should not be compared directly with experimental values for finite nuclei, although the results suggest that if the theory is extended to apply to finite nuclei it probably would be in agreement with experiment.

From abstract: "This report summarizes the past year of study of fine, molded, multicrystalline graphites with particular reference to the interrelation among certain properties with the composition, fabricating techniques, and thermal treatments found to influence them."

Quarterly report describing progress on a project to develop and evaluate high-temperature tungsten alloys. This report discusses initial experiments to compare different alloy combinations.

This report is a compilation of available data on ruptures that have occurred since KER start-up. Some of the data presented in this report are inconclusive, but are reported to insure a complete summary of all potential and confirmed ruptures.

This report discusses the use of liquid ammonia as a tracer for the detection of cooling water leaks into the piles. It is both safe and cheap and can be detected by methods adaptable to direct-reading instrumentation on a continuous-flow sample. Moderate capital costs and materials costs of less than $50 per pile test are anticipated.

To run a second BPA outage test at the DR-Reactor using the same instrumentation and procedure but increasing the trip-out level fro 800 MW up to a maximum of 1200 MW.

On the 0000-0800 shift, October 15, 1959, the delayed neutron monitor on KER Loop 2 gave a high coolant activity signal indicating a possible fuel element failure in this loop. KE reactor was shutdown immediately thereafter. This report is being written to summarize the events pertinent to this KE reactor scram and to discuss the results and significance of data from analyses on coolant and coupon samples taken from KER Loop 2.

This report discusses the isothermal hydraulic demand characteristics were determined by laboratory experiment for full charges of self-supported I&E fuel elements in a zirconium process tube. Pressure drop, flow rate data, and the calculations of annulus-to-hole flow ratio are presented. For self-supported fuel elements, pressure drop does not vary with temperature as much as it dies for non-self-supported furl elements.

The objectives of this production test detailed in this report is to evaluate the behavior during irradiation of tubular fuel elements with hot-headed end closures. With natural uranium twenty-inch tube- and-tube elements will be irradiated to a goal of 2500 MWD/T in the KER loops. The inner tubes will be closed by the hot-heading technique and the outer tubes will have normal welded closures.

A single swelling capsule was irradiated in DIR reactor as a test of the operation and design of a series of similar capsules to be charged later in a large scale uranium swelling experiment. The ratio of measured to calculated uranium temperature increased from 0.7 during the first two-thirds of the irradiation to 1.08 during the last one-third. Most of this change can be accounted for by a flux shift toward the rear of the reactor during the last one-third of the irradiation. Because of this flux shift it was impossible to determine an accurate temperature correction factor to be applied to the full scale test to follow. Radiometallurgy revealed an asymmetric radial temperature distribution in the fuel. Capsule tests showed that this asymmetry was caused by in complete NaK coverage of the rod; Capsule design change will prevent this.

The objective of this document is to authorize large-scale irradiations of alloyed, low hydrogen, dingot uranium fuel elements in order to evaluate their suitability as a reactor fuel. Although purer than standard ingot uranium, the dingot uranium produced in the past by Mallinckrodt Chemical Plant has demonstrated inferior dimensional stability during irradiation. This instability has been manifest as warp and as swelling at the center of the fuel elements. To reduce this instability 150 ppM iron and 100 ppM silicon haze been added to refine the grain structure of the uranium. Destructive examination of grain structure and nondestructive examination of apparent core quality have indicated that this material is equivalent or superior to ingot material. In light of forecasted metal requirements, the Mallinckrodt Chemical Plant must be brought on stream as rapidly as possible. This PITA schedule will meet the programmed metal delivery requirements with a minimum of risk potential while accelerating the evaluation of dingot uranium as to its suitability as a reactor fuel. Completed irradiation performance data from small scale testing will not be available before dingot material is charged under this PITA. Therefore, the possibility exists that unforeseen dingot uranium properties will result in poor performance. Low …

This report details the proposed E-N, tritium-plutonium producing reactor loading is made up of striped columns of lithium-aluminum (mint) target slugs and enriched uranium (.947 w/o U{sup 235}) slugs. Both target and uranium slugs are of the I & E geometry. The ratio of mint (N) to uranium (E) is determined by the requirement of sufficient k-excess to sustain an operable reactor. The designer of the lattice loading has a choice between natural lithium (7-5 w/o Li{sup 6}) or lithium enriched to {approximately} 38.5 w/o Li{sup 6} for the mint slugs, assuming enriched or ``black`` mint is available. It is possible to show at least 0.8% increase in total conversion ratio for an E-N load for enriched mint (38.5 w/o Li{sup 6}) over natural mint. The basis of the calculations rests on measured E-N length ratios for equivalent neutron multiplicative properties for both types of mint (in a dry pile) plus analytical calculations. It is shown that both increased Pu and H{sup 3} production are obtained by using blacker mint. The basic phenomena are (1) increased resonance capture in U{sup 238} due to more uranium volume in the black mint E-N lattice and (2) more efficient utilization of neutrons in …

Introduction: In connection with the development of a comprehensive plan for a ten year civilian power reactor program, the Evaluation and Planning Branch, Division of Reactor Development, United States Atomic Energy Commission has had a series of status reports prepared to present a comprehensive review of the technical and economic status of various nuclear power reactor concepts.

This document details activities of the irradiation processing department during the month of December, 1959. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering Operation; Employee Relations Operation; and Financial Operation.

Abstract: The measurement and calculation of the reactivity absorbed by Xe135 as functions of time after shutdown for the SRE are described.

The objective of this irradiation is to verify the corrosion rate of a cluster-type fuel element under conditions of high specific power and central core temperatures. The test will also be used in the development of rupture detection instrumentation and decontamination procedures as a necessary part in the development of the NPR. In this test a fuel element irradiated to 3200 MWD/T will be ruptured, the rate of rupture product released will be determined and the gamma spectrum from fission products released into the coolant will be observed. Permission is requested for charging three 7-rod cluster fuel elements (one previously irradiated to 3200 MWD/T at the Hanford projects, the other two unirradiated) into the GEH-R33P7 loop of the ETR. The irradiated element will have attached to it a hydraulic mechanism for opening a defect in one of its fuel rods. The other two elements are to serve as heaters to enable the loop to operate at desired temperatures.

This report details planning to run a second outage test at the DR-Reactor using the same instrumentation and procedure as an earlier test but increasing the trip-out level from 800 MW up to a maximum of 1200 MW.

In order to predict graphite damage rates for the NPE, it is necessary to have some knowledge of the neutron fluxes at various energy levels. In particular it is important to know flux levels relative to those existing where graphite samples have been irradiated experimentally. Three-group flux levels have been calculated for regions far from boundaries in the core of the NPR and of a K-reactor. The results reported here have been obtained from more detailed calculations using better cross section data than used in obtaining a previous set of values (for NPR only). The calculations of effective group cross sections for the reactor lattice still are not entirely satisfactory however, and it is hoped that some improvement can be made. Since any revision of the cross sections will affect the calculated flux levels, the values reported here should be regarded as preliminary. It is felt that the relationship between NPR and K fluxes should be fairly satisfactory, since identical calculational methods were used for both.