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)

In order to decide the value and type of information to be obtained from outer space with regard to its pertinence for the evolution of life, a brief review is presented of the current status of our thinking on the origin of life on earth. This points up the particular kinds of chemicals whose presence, or absence, on other astral bodies might be significant. Heretofore, the only data available are the result of telescopic spectroscopy. We report here information indicating the presence in meteorites of complex organic materials, some of them apparently uniquely pertinent to life processes.

None

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.

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.

A 1.6% enriched U rod was irradiated in 105 DR to 450 MWD/T, using a NaK-filled Al capsule. The capsule was discharged in November 1958 and examination begun. Results showed that the NaK in the capsule was trapped in the expansion chambers, leaving part of the fuel rod uncovered, causing gross heating on one side and failure of the rod, preventing the desired data from being obtained.

In accordance with the NPR technical criteria which require that provisions be made for useful isotope production in the control rods, both isotope-producing and non-production rods have been designed. Design Analysis has been requested by Reactor Plant Design to specify the number and location of the isotope rods for the initial installation. This choice, however, cannot be made without knowledge of the fuel element characteristics and a prediction of the reactor operating techniques. Some of the factors affecting the use of isotope-producing rods are discussed in this letter, and the following recommendation is made.

None

The recent increase in the incidence of groove pitting on X-8001 clad fuel elements in the old reactors apparently refutes the earlier hypothesis that surface segregation of the secondary phase of this alloy was the primary cause of the unique, preferential attack sustained during irradiation. Components received within the past fifteen months have exhibited essentially none of the segregation. On the other hand, recent evidence suggests that localized penetration of the autoclave film on X-8001 may influence groove attack. The implications of this hypothesis include the necessity of special handling to preserve the autoclave film integrity or possibly elimination of the film altogether. Either certain conditions or properties of the X-8001 alloy or unusual autoclave conditions intermittently produce non-uniform autoclave films. If some of these film conditions are a result of non-uniform alloy structure in the cans, they may contribute to the groove pitting attack. This report presents the design of a test to compare the scratched and non-uniform autoclave films with uniform unscratched controls under special irradiation conditions to compare the incidence of groove pitting.

The information reported here was requested by M.H. Schack, Facilities Engineering Operation. It pertains to the 100-K water plant expansion Project CG-775. In this project, cooling water pumping capacity will be increased from 178,000 GPM existing to 188,000 GPM and 200,000 GPM for five and six modified pumps respectively. Effects of increased through-put on process tube flow metering elements and on the reactor Panellit gauges have been examined. This information was required to determine cost of modifications if any were required. It was understood the precision of the cost estimate was to be about 40 percent. As a result of the increased process water flow through K reactor, no modifications are necessary to the orifices or venturis. The Panellits will require a mechanical adjustment. The need for such an adjustment already exists and the work has already been planned. If the maintenance proposal is approved, no expenditure of money for Panellits should be required in conjunction with the pump expansion. In some cases it appears that a TAI limit of 130 C may not be met by existing orifices with increased flow rates. However, in all these cases, the 130 C TAI limit could not be met by existing flow …