A summary of the activities in the 4th operational year is presented for the ORR. On-stream time at 30 Mw was relatively steady with 75.4 and 83.7% representing the lowest and highest quarters. Modification of the ball-latch mechanism of the shim-rod-drives is in progress. The primary cooling system bypass control valve was modified and a d-c pony motor was added at the No. 3 primary pump. This addition was made to increase the reliability of adequate water flow for afterheat cooling. A study of coreboiling detection is being conducted. (J.R.D.)

The practical application of heavy water purity determination by infrared absorption spectroscopy using standard laboratory equipment was demonstrated in connection with the operation of heavy water-moderated Special Power Excursion Reactor Test (SPERT II) Facillty at the NRTS. (auth)

The four experimental through-tubes provided in the Experimental Gas Cooled Reactor will extend directly through the core of the reactor and penetrate both the upper and lower pressure vessel heads. Each tube is anchored in an upper head nozzle and the bottom end is allowed to slide in a lower head nozzle. This lower nozzle is basically a T'' section that provides bottom access to the through-tube and a side access for the piping which connects the throughtube to the experimenter's cell. Due to differential thermal expansion of the through- tubes relative to the reactor pressure vessel, vertical movement of the through- tube within the T'' section will be experienced. At the same time a horizontal thrust applied to each tube by thermal expansion of the piping to the experimental cell will result in metalto-metal contact between each tube and the lower T'' section. Tests were conducted on three types of bearing material proposed for use on the through-tubes and T'' sections to minimize galling which can be expected to occur. Stellite No. 12 has been demonstrated to be an adequate bearing material for the intended application. (auth)

Progress on the gas-cooled reactor program is reported. Separate abstracts were prepared for each of the 14 sections. (M.C.G.)

A system for generating detailed tables of gamma ray cross section data has been devised for use on the IBM7090 computer. This sy;tem obviates the preparation of large amounts of cross section data. It also provides a scheme for rapid access to these tabulated values. (auth)

Openation of a 6-in.-dia foam separation column with Sr/sup 89/ tracer and dodecylbenzenesulfonate as a surfactant and foaming agent was continued. The catalytic oxidation of H/sub 2/, CO, and CH/sub 4/ was studied using a nickel- chromepalladium ribbon catalyst. A Mark I prototype fuel assembly was sheared to within 1.5-in. of the end by modifying the gas hydraulic system of the shear. The force required to shear a highly carburized Mark I fuel assembly ductile tubing. Demonstration of the mechanical dejacketing of the SRE Core I fuel burned to approximates 675 Mwd/ton and cooled about 2 years is complete, and decontamination and equipment removal from the segmenting cell is approximately 90% complete. Ten SRE Core I fuel jackets were dissolved in aqua regia and analysis showed negligible U and Pu. A semicontinuous leach run, in which -2 mesh graphite fuel containing 2.6% U was leached in 90% HNO/sub 3/ at 60 deg C, gave only 0.37% U loss. Graphically estimated spectral factors for radiation between tubes within fuel bundles and improved wall radiation factors were rised to calculate the temperature distribution expected within spent fuel elements. Further studies of the dissolution of zirconium oxide by HF in fused salt resulted …

A reference pulse generator, transistorized timedifference-to-pulse- height converter, and beam pulse scaler are described and discussed as to their function in the fast neutron time-of-flight system for use with cyclotrons, The system is used for angular distribution and cross section measurements. Results are presented showing time resolution, absolute counting efficiency, and beam pulse scaling of the system. (N.W.R.)

Sections are included on reactor project activities, materials properties, metal and ceramic fabrication, applied research, and fundamental research. Separate abstracts were prepared for each section. (J.R.D.)

The report is divided into the following sections: physical metallurgy, metallic fuels development, fabrication development, fuel evaluation, and plutonium recycle program.

This is the monthly report for the Hanford Laboratories Operation June 1962. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, physics and instrumentation research, operations research and synthesis, programming, and radiation protection operation are discussed.

The critical-mass incident at Recuplex has necessitated resumption of reworking Task 1 supernatants through the Purex process. The supernatants contain relatively small quantities of plutonium and will be shipped as treated nitric acid solutions in PR cans. Current plans are to introduce the solutions into the Purex process at the normal rework position, the HAF Make-up Tank E6. This report summarizes the anticipated character of supernatant recycle material, outlines process control requirements and recommends procedures for assuring critical mass control within the limits of the current process control specifications.

None

None

The objective of this test are: (1) to operate two single pass tubes in the 1706-KE in-reactor facility using water treated in the water treatment pilot plant, (2) to obtain samples of effluent water, and (3) to obtain corrosion measurements from weighed fuel, dummies, and coupons.

This is the monthly report for the Hanford Laboratories Operation March 1962. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, physics and instrumentation research, operations research and synthesis, programming, and radiation protection operation are discussed.

This report provides technical specifications applicable to the eight operating production reactor facilities, namely B, C, D, DR, F, H, KE and KW (referred to as the Hanford Production Reactors), located at the Hanford Site in the State of Washington. Basic descriptions of the reactor facilities are also contained in the report.

Hood 7-A will be used to handle filtrate and miscellaneous aqueous waste streams from the RMC button line. This document defines the critical mass control specifications for hood 7-A.

The method of neutron activation analysis of U by delayed-neutron counting was investigated in order to ascertain if the method would be suitable for routine application to such analyses. It was shown that the method can be used extensively and routinely for the determination of U. Emphasis was placed on the determination of U in the types of sample materials encountered in nuclear technology. Determinations of U were made on such materials as ores, granite, sea sediments, biological tissue, graphite, and metal alloys. The method is based upon the fact that delayed neutrons are emitted from fission products from the interaction of neutrons with U/sup 235/. Since the U/sup 235/ component of U undergoes most of the fissions when a sample is in a neutron flux, the method is predominately one for the determination of U/sup 235/. The total U in a sample or the isotopic composition of the U in a sample can be determined provided there is a prior knowledge of one of these quantities. The U/sup 235/ content of a test sample is obtained by comparing its delayed-neutron count to that obtained with a comparator sample containing a known quantity of U/sup 235/. (auth)

The losses of charged particles from a detector aperture, due to multiple scattering of the particles in thin layers of material preceding the detector, are calculated with the aid of a high-speed digitai computer. The rootmean-square multiple-scattering angle is computed after the method of Moliere, and graphs are presented for protons and alpha particles scattering in nitrogen, aluminum, argon, copper, silver, and gold, and for deuterons and tritons scattering in aluminum and copper. The range of particle energy covered is from 5 to 100 Mev. A second program is used to construct a map from which losses, as a function of the root-mean-square multiple-scattering angle and the scattering configuration, may be directly read. The FORTRAN listings for both programs are given. (auth)

The M-101 (an all steel) shipping container and safety cage was in use for about 15 years for transporting and storage of enriched U. With slight modifications and the addition of an inner container, the M-101 can be used to ship Pu metal. Various physical tests performed on the M-101 and inner container are described, and critical array sizes under various conditions of spacing and flooding are discussed. (auth)

None

Sintering is broadly defined as all aspects of the adhesion, densification, shrinkage, and consolidation which accompanies the reduction in surface energy of a powder compact which occurs upon application of heat and/or pressure. The process of sintering is then further divided into the initiation, material transport, and grain growthrecrystallization stages. A general review is made of each of these stages, noting typical work which has contributed to progress in each area involved. (auth)

The results of a study of the installation and operational characteristics of a SNAP reactor system integrated with a manned space station are presented. The reference system selected was an 11 kw(e) version of a SNAP 2 system employing multiple power conversion units coupled to a single reactor source. Of prime importance is the reactor radiation shield required for the manned system and the use of design features which minimize the shield weight. The weight of the radiation shield is highly dependent upon the geometrical configuration of the space station and the reactor since shadow'' shielding of the manned compartments is required for minimum weight systems. The installation and shielding requirements of the 11 kw(e) system were considered for two types of space station configurations; one was a 10 ft dia cylindrical station with a reactor separation distance of 50 ft and the other a 150 ft dia toroidal station with the reactor located in the hub. The weight of the power system installed in the cylindrical space station was about 9000 lb of which 6000 lb was required for shielding. The weight of the system for a toroidal station was approximates 25,000 lbs of which 20,000 lbs was required …

This report addresses the determination of coefficient of friction.