Preface: The intermediate heat exchanger is designed for operation in a nuclear power plant using liquid sodium as the primary and secondary coolant. Since the primary fluid coming from the reactor is radioactive, the purpose of the IHX is to transfer heat to a nonradioactive fluid which then goes to a steam generator. Because of this activity the until will be enclosed in a concrete pit and will not be accessible during periods of operation. Immediately after shut down it will be necessary to allow time for radioactive decay before the unit will be accessible to personnel. Because of inaccessibility and possible long periods allowed for decay time, it is imperative that the unit give trouble free operation. During periods of shut down, the internals should have easy access for inspection and repair if necessary so that down time is held to a minimum. The general arrangement of the heat exchanger described in this report presents a conventional design utilizing known materials and existing methods of fabrication. In further consideration of all concepts, designs and analyses developed during this period of the program, it is felt that this preliminary design will provide an intermediate sodium heat exchanger of lower cost …

From foreword: This report describes the goal of the overall SDR program to demonstrate the commercial generation of electrical power.

A conceptual study on design of sodium-cooled reactor steam generators was conducted. Included is a detailed description of the preliminary design and analysis, based on the use of known materials and existing methods of fabrication. (See also APAE-41 Vols. I and III.) (J.R.D.)

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The energy distribution of ions and electrons in DCX are being studied by means of the Fokker-Planck approximation to the Boltzmann equation. An IBM- 704 code, called FOPP, was constructed to solve simultaneously the coupled Fokker-Planck equations for each of the two species of particles. This report discusses the difference scheme employed and derives the boundary conditions necessary in order that this difference scheme conserve energy and particles in the absence of sources and sinks. In particular, detailed discussion is given of problems arising from the use of two grid sizes, which proved advantageous on account of the great difference in the mass of ions and electrons. (auth)

A fuel-cycle economic study was made for a 315-Mw(e) graphite-moderated U/sup 235/-Th-fueled fused-salt reactor. Fuel cycle costs of approximately 1.3 mills/kwh may be possible for such reactors when reprocessed for U/sup 233/ and U/ sup 235/ recover y at the end of a 9-year cycle. Continuous removal of fission products during the reactor cycle does not appear to offer any great economic advantage for the converter reactor considered. (auth)

To determine the safety of the array in the storage vault for the SM-2 experimental fuel plates, two criticality criteria were applied. A maximum of 18 fuel plates was stored in sthainless steel tubes and the tubes belted to a frame on the wall to prevent movement. No tube could go critical by itseIf. The vauit was then assumed completely flooded by water. In the first calculation, the fuel array was assumed to be distributed uniformly over the wall forming a large slab. This method indicated the array might be critical if the steel tube and cadmium lining were neglected. In the second method, a conservative calculation, wnich included the steel tube and cadmium lining was made. This method indicataed the array was subcritical. Calculations were then made of the criticalty of the SM-2 vault without the steel--cadmium tubes and wcoden blocks. The multiplication factor of the vault was also calculated. In order to determine the accuracy of these calculations, an ORNL critical experimental array was calculated applying the same analytical techniques. (M.C.G.)

The rate of oxidation of chromium(III) to chromium(VI), catalyzed by ruthenium, was determined at various temperatures and oxygen concentrations. The rate at 300 deg C was too rapid for measurement by aliquot sampling. In the temperature range of 225 to 275 deg C, oxidation was rapid and the rate increased with oxygen concentration. A linear dependence of initial oxidation rate on the reciprocal of chromium(VI) concentration suggested that a rate-controlling step in the reaction mechanism may be desorption of chromium(VI) from the ruthenium catalyst. The activation energy calculated for the reaction is 19 kcal/mole. (auth)

Document issued by the Office of the Attorney General of Texas in Austin, Texas, providing an interpretation of Texas law. It provides the opinion of the Texas Attorney General, Will Wilson, regarding a legal question submitted for clarification: Whether the Civil Air Patrol is exempt from the payment of the license fee for motor vehicles under the provisions of Article 6675 A-3, V.A.C.S.

Document issued by the Office of the Attorney General of Texas in Austin, Texas, providing an interpretation of Texas law. It provides the opinion of the Texas Attorney General, Will Wilson, regarding a legal question submitted for clarification: Applicability of the Occupation Tax levied by Section 31, Article 7047, Vernon's Annotated Civil Statutes, to rodeos promoted or produced by specific sponsors.

Document issued by the Office of the Attorney General of Texas in Austin, Texas, providing an interpretation of Texas law. It provides the opinion of the Texas Attorney General, Will Wilson, regarding a legal question submitted for clarification: Sale of personal stock of officers and directors of an insurance company and the duty owed by such officers and directors to the insurance company.

The monocarbide and the mononitride of uranium are potentially useful ceramic nuclear fuel materials. This paper reports the results of exploratory investigations of the reactions of uranium monocarbide and uranium mononitride with boiling water. Uranium dioxide, chemically stable in deoxygenated boiling water, was used as a control.

An estimate of the critical current for the case of breakup by an arc has been obtained. It is shown that the same technique allows a quick estimate of the critical current in the case of no arc, using neutral and plasma breakup instead. (auth)

Foreign research and power reactors are tabulated. Nuclear power buildup goals are given for each nation on which information is available. (J.H.D.)

The Early Program is designed to compute the peak power and total energy release vhich will occur in a reactor due to a step increase of reactivity. These excursions may be terminated by one of several automatic shutoff mechanisms caused by the power excursion itself. The calculation assumes no thermodynanaic delays and assumes void formation is due to heat conduction from the cladding to the moderator. (W.D.M.)

The early program is designed to compute the peak power and total energy release which will occur in a reactor due to a step increase of reactivity. The calculation assumes no thermodynamic delays and assumes void formation is due to heat conduction from the cladding to the moderator. For periods in the 1 to 10 msec. range, the calculation yields results which are in excellent agreement with SPERT experimental results. (W.D.M.)

Data obtained in test of Bureau of Mines electrocerium are presented. The cerium was subjected to spectrochemical analysis, x-ray diffraction analysis, thermal analysis, metallographic tests, and high-pressure experiments.

Judging scratch depth or surface roughness by unaided visual inspection under controlled conditions, while rapid and popular, is not quantitative. Comparison methods improve reproducibility but are generally not applicable to evaluation of depths of single widely spaced scratches. Stylus-type contour recorders yield valuable scratch contour data but may themselves plow through soft materials and fine details. Depth measuring microscopes are particularly applicable to measurement of pinhole depth but do not graphically reveal profiles and provide only a small field of view. The comparatively large field of view and graphic display of contour provided by profile microscopes make them particularly suitable for evaluation scratch depth as well as surface roughness. A HAPO-constructed instrument has demonstrated an accuracy of +/- 50 micro inches in the range of 50 to 15,000 micro-inches scratch depth. It is a pocket-sized, portable, and can be used on horizontal and vertical surfaces by untrained persons with only brief instruction.

Some calculations have been made to estimate the characteristics of a neutron-burst type fast reactor similar to Godiva but made up of relatively small component parts--the so-called "layered assembly." One spherical and three cylindrical assemblies have been considered. Critical masses, assuming 5% voids, range from 58 to 65 kg of 93.4% enriched U/sup 235/. For a reactivity addition of 0.33 dollars above prompt criticals bursts between 2 x 10/sup 17/ and 6.7 x 10/ sup 17/ fissions were computed with accompanying temperature rises varying from 514 to 1600 deg C. The burst width at half-maximum was about 12 microseconds. To obtain an idea of the possibilities of stress reduction which might be achieved by layerings an assembly made of small rings was considered. While the critical masses obtained here are believed to be fairly accurates the predictions concerning mechanical energy generated, total fissions, and burst width may be subject to sizeable error due to the many simplitications required to allow hand computations. Neverthelesss considerable improvement in safety and burst-size is indicated by the use of a "layered assembly" instead of an assembly composed of relatively thick parts. (auth)

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.

Document issued by the Office of the Attorney General of Texas in Austin, Texas, providing an interpretation of Texas law. It provides the opinion of the Texas Attorney General, Will Wilson, regarding a legal question submitted for clarification: Whether the Secretary of State can accept and file an application for a Certificate of Authority under the provisions of Article 1513a of Vernon's Civil Statutes for a purpose which is not identical in language and does not include the entire purpose clause authorized by said statute.

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Report documenting the ongoing research and developments of the Oak Ridge National Laboratory's Homogeneous Reactor Project.