The possibilities of producing graphite impermeable to gases and liquids, and suitable for nuclear applications, are discussed. The physical properties of such graphites depend on the different sealing processes and starting materials developed especially for the purpose. In the development of reactor grade graphite, described in a previous article, Germany was well behind other countries. As far as the making of impermeable graphite is concerned, however, advanced processes for and experience in the manufacture of impermeable materials for the chemical industry were available here before the publication of other research in this field. Liquid -- and gas -- impregnation methods are discussed mainly. The problems of sealing and the degree of impermeability that can be obtained are shown using different measuring techniques. Finally, a description is given of "pyrographite", a newly rediscovered old graphite material.

Pressure drop and heat transfer in annular tubes have been measured. The surface of the inner tube was artificially roughened. The investigation covered a range of Reynolds numbers from 200 to 100,000. The maximum heat flux, directed from the inner tube to the fluid, was about 316 W/cm2 or 2.7 x 10(6) kcal/m(2) h. The heat transfer medium was water. The hydraulic diameter of the annular tube served as characteristic length in the dimensionless Reynolds and Nusselt number. The results of all experiments are presented in graphic form. Heat transfer and pressure drop increase with the degree of roughness and depend on the type and arrangement of the roughness elements chosen.

It has been shown by studying the possibilities for determining the constants in stepwise complex formation by the interaction of hydrolyzed metal ions and ligands, that in a vast number of cases the direct measurements of metal ion and ligand activities ae necessary. Complete formulae for calculation are given and the conditions which allow some simplifications are established.

A report covering the month of January. During this time the results of neutralizing the silicon tetrafluoride scavenging system were analyzed and further progress was made on suppression of aerosol formation from fuel elements.

The power of nuclear track research emulsion as a fast neutron dosimeter is examined in the exposure of a human phantom to PuBe neutrons. Semiautomatic track scanning and high-speed data analysis obviate the major disadvantages of this dosimeter, and allow the following basic information to be obtained without a serious cost in time: the rulative proton recoil energy spectrum, the absolute differential proton track den sity spectrum, and the average proton recoil energy at various locations in the phantom. From this are calculated the total absorbed local tissue doze due is proton recoils, the local thermal neutron intensity, and that portion of the tissue doze due to thermal [formula] tracks.

One of the techniques by which highly ionized plasmas can be generated in the laboratory makes use of strong, electromagnetically driven shock waves propagating into a cold gas. In this paper the phenomenon is analyzed as a one-dimensional single-fluid hydromagnetic problem, neglecting dissipation behind the wave.

In previous letters we have reported a Av resonance, called Y*1, observed through the study of the interaction of 1.15-Bev/c K* mesons in hydrogen in the Lawrence Radiation Laboratory 15-in. bubble chamber. We now wish to report the results of the study of the three reactions [formula](1), [formula](2) and [formula](3).

"The understanding of interactions between the elementary particles, which no number thirty, and range from the massless neutrino to the Е with a mass of approximately 1320 Mev, in a prerequisite to the knowledge of the structure of matter. Although the complex aggregate of elementary particles which constitute matter displays properties different from those of the basic building blocks, nevertheless it is impossible to gain a firm knowledge of the behavior of matter without first ascertaining in the intrinsic properties of elementary particles, as well as the dynamics of their interactions. The present work is concerned with inelastic proton-proton scattering at 1.35 Bev, where single pion production predominates."

"In spite of the present scarcity of experimental data it appears likely that the lepton modes of hyperon decay will eventually become an important source of information concerning strangeness-changing weak interactions. In this paper we use dispersion relations to estimate the effect of intermediate meson states upon the lepton decay modes of hyperons."

This article presents data on the effects of gamma radiation on the physical properties of a number of different classes of plastics materials of construction which were screened to determine their capabilities and limitations for service in applications where ionizing radiation exists. The irradiations were carried out in air at ambient temperature at normal atmospheric pressure to various gamma radiation doses up to 1*10^9 r depending upon the stability of the materials to radiation exposure. The different classes of materials and materials within a class are compared.

Startup and testing of the PRTR has been divided into three parts, Design Tests, Critical Tests and Power Tests. This document is the third of the series and describes the performance tests which will be made under nuclear power and the general schedule of operation. Responsibility for the startup of the reactor has been assigned to the PRTR Startup Council. The Power Test Sub-Council, which has done the planning of the initial power operation, was established by the Startup Council. Operation of the reactor will be the responsibility of the PRTR Operation (PRTRO).

This technical report describes two atmospheric pressure rigs with associated equipment designed for investigating heat transfer and friction characteristics of different type of "roughened" surfaces to develop suitable surfaces for an H.T.G.C.R. fuel element. Operating details are given together with techniques used to measure heat transfer and friction on a surface assembly incorporating selected "roughened" surfaces. Methods for calculating Stanton numbers, friction factors, and Reynolds numbers for such an assembly are included.

Abstract; The fission product data obtained during SM-1 Core I operation (June 1957 - May 1960) is reviewed briefly and interpreted. Evidence is presented to indicate that a fuel element defect was responsible for the high fission product activity level observed in the primary coolant. Relative escape coefficients are calculated and the defect size estimated. Anticipated fission product levels during SM-1 Core II and SM-1A Core I operation are estimated from alpha surface contamination data on completed fuel elements. The importance of in-line sampling for monitoring fission product activity is stressed as well as the need for failed fuel element detection methods.

Abstract; This technical report describes the changes in the SM-1 incurred by the experiment, Core Temperature and Flow Instrumentation (Task XIV), and evaluates the possible hazard involved in these changes. Temperature and flow measurements will be taken on a Task XIV instrumented stationary fuel element, instrumented control rod fuel element and other selected points in the SM-1 core to provide data on the core steady state and transient performance. The hazards evaluation consists of a nuclear evaluation, thermal and hydraulic analysis, description of tests to be performed, and discussion of containment integrity and maximum accident considerations.

Abstract: This technical report describes the as-constructed SM-1 penetrated gasket designed for SM-1 Core and Flow Instrumentation (Task XIV). This report supplements APAE No. 79, The Summary Hazards Report for Task XIV, and evaluates the effects of a postulated failure of this gasket. The effects of failure on the Maximum Credible Accident are determined and conclusions and recommendations for the use of this gasket are made.

Introduction: This is a final report on the evaluation of Kanigen electroless nickel plating for surfaces in contact with water and steam i a sodium heated AISI Type 316 stainless steel steam generator. The purpose of the coasting was to afford protection from stress corrosion cracking originating on the water-steam side of the unit. It has been concluded that the kanigen coating does not afford adequate protection for the services condition intended. This work was performed as part of the research and development program for the United States Atomic Energy Commission sodium Components Design Project.

Abstract; The heat exchanger and steam generator for the U.S. Atomic Energy Commission Sodium Components Project will be constructed entirely of type 316 stainless steel. Because of the susceptibility of this alloy to stress corrosion cracking, it is proposed to clad all areas of the steam generator with Inconel where the stainless steel will be exposed to water and steam. This report includes a discussion of the work by numerous investigators that justify the selection of Inconel for this service. A discussion of Inconel type welding alloys is also included.

Abstract: This technical report describes the changes incurred in the SM-1 by the insertion of the SM-1 Core II and special components. The special components consist of impact specimens, a boron gradient rod, SM-2 elements, a PM-1-M element, and high burnup SM-1 Core I elements. The change in hazards, due to operation of SM-1 with Core II and the special components is evaluated. The analysis indicates there is no change in hazards.

Abstract: This technical report describes the changes incurred in the SM-1 by the insertion of the SM-1 Core II without special components. The SM-1 Core II components were made to specifications very nearly identical to those of SM-1 Core I. The differences consist of europium absorber sections, internal europium flux suppressors in the control rod fuel elements, and low impurity cladding. Each of the SM-1 Core II components with the exception of the five absorber sections new in SM-1 Core I were subjected to a Zero Power Experiment at the Alco Critical Facility. The results of this experiment indicate that the SM-1 Core II will have nuclear characteristics very similar to that of the SM-1 Core I. Since SM-1 Core II will be operated with the same mode of rod control, in the same core support structure, and with the same primary coolant flow conditions, the thermal characteristics should be essentially identical to that of SM-1 Core I. Also, all kinetic characteristics of SM-1 Core II should be identical to those of SM-1 Core I. This report demonstrates that there is no increase in potential for a hazardous situation at SM-1 due to the replacement of SM-1 Core I by …

Abstract: This technical report describes the Martin Co. PM-1-M-2 test element and analyzes the potential hazard incurred by its inclusion in the SM-1 Core II. A nuclear analysis develops power distributions and reactivity effects. Hydraulic and thermal analyses develop anticipated burnout heat flux ratios. An evaluation of the risk involved with the inclusion of this element is presented. In view of the narrow margin by which the PM-1-M-2 test element meets the minimum burnout ratios as defined by Alco Products, Inc., it is recommended that if time permits that critical facility design verification be accomplished. The PM-1-M-2 test element meets the minimum requirements for insertion in SM-1 Core II and in view of the importance of this element to the PM-1 and PM-3A program, should be considered for insertion.

Abstract: The removal of both SM-1 Core I high burnup elements from SM-1 Core II and the insertion of the PM-1-M-2 element and the SM-1 Core I spare element in SM-1 Core II is discussed. Nuclear and thermal characteristics of Core II with these changes are presented and conclusions related to the changes in the hazard potential are made. If the core change indicated by this report is made, local peaking factors will be decreased and burnout ratios will be increased. This, of course, in itself leads to a more conservative estimate of core safety. There is no conceivable reason why the perturbation may not be safely made in the SM-1 Core II.

Abstract: This technical report contains a summary of analytical and experimental work performed on SM-1 Core II, with special components is presented. The effects of these special assemblies upon power distribution and core reactivity were calculated and compared to experimental measurements. A thermal analysis was conducted to determine steady state and transient performance of the special test components of the core as well as some of the hotter standard Core II components. Experimental work discussed includes individual reactivity effects of all the special elements and the total effect of all of the elements. Power mappings were also made and are reported.

Abstract: A four decade log count ratemeter based on a three diode pump section is described. The design equations for this system are included. The unit is contained in a chassis 3 1/2" high by 9 1/2" wide by 6" deep. All circuits are constructed on etched circuit boards to facilitate testing, construction, and servicing.

From abstract: "The thorium-hydrogen-carbon system was studied by measuring hydrogen equilibrium pressures at various compositions and temperatures. Two compounds, ThC·ThH2 and ThC·2ThH2, were found and characterized. ThC·ThH2 has a hexagonal closest-packed lattice with ao=3.816Å and co=6.302Å. ThC·2ThH2 has a monoclinic lattice with ao=6.50Å, bo=3.80Å, co=10.91Å and β = 119°. The enthalpies of formation for each compound were determined."