Results are reported from a series of experimental highexplosive shots under inversion conditions at the Nevada Test Site which were made in an attempt to refine blast prediction techniques. Applications of the data in determinations of the amount of energy which remains in the blast wave as it reaches acoustic level and in determinations of the magnitude of the reflection factor when the blast wave strikes the ground are discussed. Data on shock wave propagation are presented graphically. It is concluded that the blast phenomenology of high-altitude shots can be predicted by using modified Sach's scaling. With some extrapolation to the height-of-burst versus blast-yield curve, it should be possible to make order-of-magnitude predictions of blast effects from high-altitude shots up to heights of burst of 1,000,000 ft. (C.H.)

BS>Some of the fundamental relationships in a nuclear rocket engine are discussed. The equations required to calculate the performance of the rocket are presented. The problems associated with these calculations are also pointed out. (auth)

Report of the annual conference on weights and measures, hosted by the U.S. National Bureau of Standards in Washington D.C. It includes conference proceedings, a list of attendees, information about committees and officers, and other reports or commentaries discussed at the meetings.

None

Introduction: "This is the tenth letter report on ARF Project C 127, entitled "Preliminary Studies of Scavenging Systems Related to Radioactive Fallout." This report covers the period from October 1 to December 1, 1959."

Quarterly progress report describing work related to a study of radiation physics problems involved in the design of high-level cesium-137 gamma sources. It outlines work completed and ongoing goals for the project.

"Abstract: Data on analog magnetic tape characteristics are presented along with results from conventional and modulator playback system operational tests. Theoretical considerations imply that the modulator head should be loss sensitive to short term variations in record and playback tape speed. It is also pointed out that although a tape recorder operating at 7.5 in./sec has only 12 kcs bandwidth for sinusoidal wave forms. the record and play-back amplifiers for pulse recording can operate with other bandwidths. These amplifiers must be capable of passing a square wave with a frequency of 1.5 kcs. A square wave of such frequency represents the condition of 350 microsecond pulses with a 700 microsecond period. (J.R.D)."

The Fuel Development operation at Hanford uses a variety of in-reactor facilities to test experimental and prototypical fuel elements. High pressure-high temperature water loops in constant use are the 6 in. {times} 9 in. ETR core facility, the 3 in. {times} 3 in. ETR reflector facility and four front-to-rear loops in the 100-K East reactor at Hanford. Low pressure water cooled test facilities in use are located in the MTR the various Hanford reactors. Stainless steel sheathed the thermocouples 1/16-inch diameter insulated with MgO or Al{sub 2}O{sub 3} when properly fabricated and installed, will reliably measure in-reactor temperature up to 1000 C and at least 4000 MWD/T exposure. This report provides a brief description of some of the temperature monitored in-reactor experiments.

This Program detailed in this report is related to the Product Development, Palmolive, and Non-Production Fuels Programs under New and Special Products and Processes Category where the prime separations plants supply raw materials or processing support of these other programs. Likewise, Increased Plant Return Program items affecting the Purex, Redox, and UO{sub 3} Plants must be integrated with the Continuity of Operations Program. Since the primary separations plants feed the 234-5 Facility, the 234-5 Program is also related to Continutiy of Operation.

Defect testing of the irradiated specimens will provide kinetic autoclave defect test data on the behavior and performance of coextruded Zr-2 clad fuel failures; this should reveal the effects of exposure on the coextruded U-Zr-2 bond.

The C Reactor was proposed for producing tritium. To evaluate the performance of enriched U-Al J elements and natural Li-Al alloy target (N) elements, 60 charges containing both J and N pieces were irradiated under a variety of conditions in C Reactor. No ruptures were sustained; however Tube 3276-C was discharged because of a suspect. Corrosion rates of J elements were not worse than for natural U irradiated under same conditions. Differences between corrosion of J elements prepared by three different methods were not significant.

The objective of this production test authorization is to permit the irradiation of sixteen, 0.394 inch diameter, metallic uranium spheres at temperatures between 300C and 800C in the 3674 ank 3865 KW front-to-rear test tubes. Eight spheres will be of natural uranium and eight of enriched (3% U-235) uranium. Temperature controlled and temperature monitored test capsules containing either four natural or three percent enriched uranium spheres will be irradiated. Pre- and post -irradiation examinations will be performed to determine the extent of swelling. The temperature of the spheres will be controlled at the same constant level during reactor operation and during shutdowns. Capsule exposures will vary from one to 24 months. This production test authorization will permit the irradiation of two capsules each containing four natural uranium spheres and two capsules each containing four three percent enriched uranium spheres.

The November 1959 monthly report for the Chemical Processing Department of the Hanford Atomic Products Operation includes information regarding research and engineering efforts with respect to the Purex and Redox process technology. Also discussed is the production operation, finished product operation, power and general maintenance, financial operation, engineering and research operations, and employee operation. (MB)

This document presents a nomograph from which the relationship between reactor operating parameters, tube power, and outlet temperature can be correlated with rupture rate. The index indicates the severity of the reactor climate during irradiation and does not include the metal quality parameters defined in the rupture rate equation. The general form of the Operational Severity Index Equation is OSI=P{sup 3.3}/1000{times}t{sub 0}{sup 8.7}/100, where OSI, is the unitless Operational Severity Index, P is the tube power in kW, and t{sub 0} is the tube outlet temperature, in degrees C.

This document provides details of activities of the Hanford Operations Office during the month of September 1959. (FI)

This monthly document details activities of the Hanford Operations Office during the month of November 1959. (FI)

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

The possibility of overboring in existing Hanford reactors to increase slug size has been suggested as a means of increasing production. The effect on control system strength of such a modification is, of course, one of the factors which must be considered. This report presents the results of a study requested by the Reactor Physics Unit of the local buckling of control rods as a function of the migration area. The migration area varies for different lattice make-ups, but due to the complexity, calculation of T and L{sup 2} for changes in slug and hole size was not done here; instead, the range of values given by the ``Giant`` Program was assumed.

The objective of this test is to demonstrate the feasibility of projection fuel elements for use In existing process tubing and to determine the reduction in rupture rates or hot-spot incidence so achieved. This test is to authorize, (a) charging 20 columns of bumper type fuel elements and 20 columns of control elements per reactor into B, D, DR, F, and R Reactors for irradiation up to 1200 MWD/T exposure, and (b) irradiation of four columns each of enriched (0.947%) bumper and enriched (0.947%) normal type fuel elements until two ruptures are sustained in each (or until one group shows a significant improvement).

In view of the more-or-less eminent conversion of at least some of the older Hanford reactors to minimum process tubes it has been requested the feasibility and economics of increasing the process channel size by overboring be studied. This report is concerned with the feasibility from an operational physics standpoint of either raising reactor power levels with present aluminum process tubes and redesigned fuel elements raising reactor power levels with zirconium replacement tubes of current outside diameter, or raising power levels with either zirconium or aluminum tubes of 200 mils greater outside diameter. This report is of a survey nature only, and the data contained herein should be considered in that light.

The Expansion Study Task Force has evaluated several cases of Hanford reactor operation at power levels considerably higher than is presently obtained in the six older reactors. These higher power levels result in more rigorous operating conditions of temperature, heat flux, neutron flux, hydraulics, reactor control, etc. The purpose of this document, the various components of which were prepared by Process and Reactor Development Sub-Section personnel, is to assess the technical feasibility of operation under the proposed conditions, and to delineate those specific areas of development effort which may be necessary to provide adequate support for an expansion program.

Abstract: Testing of the Natural Circulation Steam Generator has been conducted with the Sodium Reactor Experiment. A necessary modification of the original feed-water control system was utilized.

Abstract: In the present study, the variation of moderator composition was postulated to be effected by the inclusion of varying amounts of beryllium oxide in the graphite of an SGR.

An X-8001 clad, I & E natural uranium fuel element irradiated in tube 2762D to 886 MWD/T was discharged in February 1959 and sent to Radiometallurgy Laboratory for examination at the request of Irradiation Processing Department. After discharging tube 2762D which was loaded with X-8001 clad fuel elements, a fuel element (KL018D) was observed to be very badly pitted. Visual and photographic inspection revealed the pits had penetrated into the AlSi layer. The pitting appeared as the result of erosion-corrosion as there was no evidence for intergranular or transgranular corrosion and no mechanical deformation. Chemical analysis for %Ni and Fe indicated normal concentrations.