Results of an initial examination of non-fuel-bearing materials for the Marine Gas Cooled Reactor are presented. Areas given specific attention include structural and cladding materials and graphite moderator. Other areas not included because of classification include zirconium hydride moderator and control materials. In each report section, available pertinent information is summarized and a plan for future work is given.

This is the fifty-fourth progress report of the Laboratory for Nuclear Science at the Massachusetts Institute of Technology. Progress reported covers the period for Physics and Chemistry through April 30,1959. Progress is reported on: (1) Chemistry of the fission elements group, (2) Nuclear chemistry (inorganic) group, (3) Nuclear chemistry (organic) group, (4) Cosmic ray group, (5) High energy accelerator physics group, (6) Bubble chamber group, (7) Liner acceleratory group, (8) Rockefeller generator group, (9) ONR generator group, (8) Radioactivity group, (9) Cyclotron group, (10) Theoretical group, (11) Personnel listing.

"The laboratory experiments and field testing which were carried out in the development of a homogeneous gas labeling device are described. Detailed drawings of the injector and its most critical components are included. Recommendations are made for a production model of the device along with estimated costs of construction."

The methods of successive displacements or relaxation methods are investigated for a class of nonlinear problems. In particular it is shown that these methods are applicable to a large class of nonlinear problems arising from variational problems which yield elliptic equations. Constructive existence and uniqueness theorems are presented for the discrete problem and criteria are given for a practical method of obtaining solutions. The example of a discrete Plateau problem is used to illustrate the feasibility of the results. The processes are also shown to apply to uniformly elliptic problems.

"The present work describes some computer-oriented techniques for the application of the laws of quantum mechanics to the problems of determining basic structure of diatomic molecules. At times, calculations of the type described here yield results which can be compared directly with experiment to check the validity of the hypotheses and technique used they may predict the results of unobserved experimental phenomena; and, in many instances, they provide quantities which are necessary for the interpretation of experimental data but which cannot be directly measured."

"This is the fourth Quarterly Report on Task II dealing with Methods Development for Radio-Tracing of Coal Product Hydrocarbons With Tritium. The problem of radio-chemical purity of traces is resolved by developing a new method of labeling with tritium that generates essentially no highly tagged side products or impurities."

"The purpose of this report is to examine the safety aspects of PNPF restart on continued operation, after completion of the core cleanup and system modifications."

Introduction: In conjunction with the final design and development of a 70 MW sodium intermediate heat exchanger and a sodium steam generator, an analysis is required which can be used as a basis for a determination to scale up or scale down the designs. Included in this analysis are those considerations leading to the recommendation of the best prototype test unit and to some of the limits imposed on scaling up or down when considering future applications of designs other than those actually tested. In addition, these considerations include aspects required to accurately predict the performance, operation, mechanical reliability, and feasibility of fabrication of the 70 MW design.

Abstract. The design, selection, and calculation of the Army Package Power Reactor shielding are described. The APPR-1, a prototype of a package reactor for remote locations, has a primary shield of iron and water. this shield has been adopted to permit fast erection and to provide low transported weight. economically, including transportation cost, the iron water shield is better than a lead water shield and is competitive with a concrete shield for a remote site. Because of the location at Fort Belvoir,Va., the shielding requirements for the APR-1 are considerably more stringent than those for a reactor at a remote base. Since the secondary shielding which surrounds the entire primary system must provide protection for personnel at any location outside the vapor container, concrete is provided for this need.

Technical report describing the Brookhaven National Laboratory nuclear reactor and the physical studies of heat transfer and efficiency conducted to improve nuclear reactor operations.

Abstract: "Steady states decomposition product partial pressures of water exposed to an epithermal neutron flux of about 2 x 10-11, at 235F, were 0.4 to 0.65 psi. Recombination studies were made by adding hydrogen and oxygen gases to water under irradiation at different temperature and flux levels. The data show a significant effect of temperature on the rate of pressure drop of the added hydrogen and oxygen. In operating at two flux levels, 1.2 x 10-11 and 2.0 x 10-11, no difference in the rate of pressure drop with flux could be detected with the instrumentation on the test apparatus."

Report discussing progress made in metallurgy by the Ames Laboratory during the the term from October 1, 1950 to March 31, 1951.

Report discussing the progress made on various chemistry research projects at Ames Laboratory during the period from October 1, 1950 to March 31, 1951.

Abstract: An analysis of the improved Model Number 350 Process Monitor was made so that it could be determined if this instrument would fulfill the need for gamma radiation detection instruments in the production buildings of Y-12. The results of these tests indicated the possibility of marked improvement in the monitor characteristics with minor redesigns. After these improvements have been made, this instrument will be a satisfactory monitor for this area.

A report about the start of preliminary measurements of the thermal conductivity of clad uranium and Zircaloy 2 prior to irradiation.

A report about elevated-temperature properties of dilute uranium-aluminum and uranium-zirconium alloys. Hot-hardness measurements show that small amounts of wither aluminum or zirconium increase the hardness of uranium at elevated temperatures.

Originally issued as S and P 1963A, Parts I and II. This report covers a design and feasibility study of a pebble bed reactor-steam power plant of 125 megawatt electrical output. The reactor design which evolved from this study is a two-region thermal breeder, operating on the uranium-thorium cycle, in which all core structural materials are graphite. Fuel is in the form of unclad spherical elements of graphite, containing fissile and fertile material. The primary loop consists of the reactor plus three steam generators and blowers in parallel. Plant design and system analysis including cost analysis and capital cost summary are given.

Abstract: The structural design criteria and the system basic loads for the SNAP 2 compact power unit are presented.

From introduction: A thorough investigation, both experimental and theoretical was carried out on plaque type radiators. Five phases of the work are reported: preliminary design analysis, experimental measurements, computational program and analytical studies, followed by a general discussion of source design principles. The results are summarized in graphs and formulae which may be applied to other design requirements.

From abstract: A gamma-radiation telemetering system was utilized by Civil Effects Test Group personnel to measure fallout levels at the Nevada Test Site from the vicinity of Ground Zero out to, and including, an area bounded by Reno, Nevada; Salt Lake City, Utah; Kingman, Arizona; and Barstow, California.

Introduction: Recently a new type of charged particle detector, the so-called spark chamber, has been developed. The detector has certain advantages in high energy physics over bubble chambers aside from its much lower cost.

Abstract: A loop has been constructed to study the removal of water and highly volatile materials from Organic Moderated Reactor coolant by vacuum degasification. An analysis of the process was made to determine the most important parameters for study during the experimental program.

Abstract: Accurate measurement of OMRE fuel-element surface temperatures is important for evaluation of the performance of organic compound as moderator and coolant materials in nuclear reactors; a mixture of polyphenyls has been used in reactor operation to-date.

Abstract: The influence of plutonium on the nuclear characteristics of organic moderated cores is studied.