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This is the second summary report of Hanford waste disposal experience. The data was tabulated from reports submitted by the contractors to the Richland Office of the Atomic Energy Commission. For a detailed description of Hanford waste management practices and plans the reader is directed to Hanford Waste Management Plans. There were no instances where rates of release exceeded applicable release guides and generally release rates were less than 10% of RL Appendix 0510 guides. There was a general long-term downward trend in releases to the atmosphere and to the river reflecting retirement of production facilities including the 100-B reactor in February 1968, and also reflecting the results of programs intended to reduce releases even further. 14 refs., 6 figs., 5 tabs.

The following activities were conducted during the month of April 1969: continued effort toward completion of the detailed design of the PMC-2 converter, conducted technical liaison with Westinghouse with respect to the power cable specification, conducted potting test for evaluation of material suitable for SNAP-23A converter PMC-2 leak specification, and completed design of parts for the Weld Development Program. Test data evaluation and analysis were performed on the following long-term test devices during February 1969: experimental mockup converter No. 2 (EMC-2), experimental mockup converter No. 3 (EMC-3), prototype mockup converter No. 1B (PMC-1B), 10-couple modules 1 and 3, 10-couple modules 5 through 12 on design-of-experiments tests, 6-couple module 23A3, and converter 23 P-2.

This report itemizes the irradiations performed by Testing and Irradiation Services for Battelle-Northwest. It lists the material being irradiated, awaiting disposition and material shipped during the report period.

Tests were conducted on an intermetallic compound and a loaded carbon to determine hugoniot equation of state up to 500 kbar at 72 degrees F, using flat-plate impact techniques-- a multiple-wave structure characteristic with an elastic precursor.

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The major technical bases for the FFTF first core fuel assembly conceptual design are presented in four general categories: (1) thermal, (2) hydraulic, (3) structural and (4) performance-manufacturing design characteristics. Recommended design values and conceptual engineering drawings of the fuel pin and subassembly are included .

This document comprises a compilation of preliminary data generated during the reporting months of March, April, May, 1969. Data relate to fuel element development, cladding development, fuel quality assurance, fuels evaluation, and offsite fuel programs.

The choice of materials and other vessel design decisions necessary to provide the desired life expectancy for the FTR vessel are partially dependent upon estimates of the number and type of reactor shutdowns and startups which may be anticipated. Current estimates of these so-called "cycles" are given, including scram frequency, experimental outage frequency, standard shutdowns and startups, and rapid controlled shutdowns. Also discussed are abnormal heatup or cooldown, and tentative goals for temperature controls. MTR, ETR, and typical PRTR operating histories are tabulated.

A permanent magnet probe flowmeter and a Pitot-static bubbler flowmeter are considered for coolant flow measurement in the Fast Test Reactor. The permanent magnet probe offers a sensitivity on the order of tenths of a millivolt per foot/second flow velocity, appropriate linearity of response, simplicity of operation, and no contact between liquid metal and electrical insulation. Its applicability is contingent on magnetic material having sufficient magnetization retention in high temperature and high neutron flux environments. The Pitot-static bubbler offers low sensitivity to nuclear radiation, use of off-the-shelf materials, and high, though non-linear, sensitivity. Positive displacement flow in the bubbler may eliminate plugging of the probe tubing by sodium oxide.

The thesis is concerned with the relation between a microscopic approach and a macroscopic approach to the study of the nuclear binding energy as a function of neutron number, proton number and nuclear deformations. First of all we give a general discussion of the potential energy of a system which can be divided into a bulk region and a thin skin layer. We find that this energy can be written down in the usual liquid drop type of expression, i.e., in terms of the volume, the surface area and other macroscopic properties of the system. The discussion is illustrated by a study of noninteracting particles in an orthorhombic potential well with zero potential inside and infinite potential outside. The total energy is calculated both exactly (a microscopic approach) and also from a liquid drop type of expression (a macroscopic approach). It turns out that the latter approach reproduces the smooth average of the exact results very well. We next make a digression to study the saddle point shapes of a charged conducting drop on a pure liquid drop model. We compare the properties of a conducting drop with those of a drop whose charges are distributed uniformly throughout its volume. …

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A series of three tests will be performed to evaluate: temperature gradients between wall and central coolant channels in an FFTF fuel assembly to assess thermal stress level in the pins. The potential hot spot caused by the presence of a wire wrap spacer. A single electrically-heated 7-pin fuel assembly model will be used to perform these tests. The configuration is summarized in Table 1 and the instrumentation in Figure 1. Pins 1, 2, and 7, along with the local coolant thermocouples, characterize the central and outer channels. Pins 4 and 6 provide data on the hot spot under the spacer, while pin 3 (along with pin 2) provides data on the gradient across an outer row pin.

The FTR vessel wall will be cooled with inlet primary coolant which bypasses the core and flows through an annulus formed by the vessel wall and the thermal barrier. heat transfer to the by-pass coolant stream is by raial conduction from the surrounding sodium pool through the thermal barrier wall. In the analysis, the temperature rise of the by-pass coolant stream was derived as a function of the wall thickness of the thermal barrier and the flow rate of the by-pass stream.

This document comprises a compilation of preliminary data generated during the reporting month of May, 1969. The data relate to fuel element development, fuels quality assurance, cladding development, fuels evaluation, and offsite fuel programs.