None

None

None

None

None

None

None

This is a report of the radioactive liquid wastes discharged to the ground for July, 1964, at Hanford. The report lists plutonium and uranium by mass, {beta}-emitters by curies, and quantity of waste in liters.

None

None

None

An N-Reactor inlet connector piping break was investigated in a laboratory study to determine the effect of this break on the affected process tube`s thermal-hydraulic performance. A full-scale mock-up of the piping between the inlet and outlet risers was used which included actual reactor valves and fittings. An electrically heated half-length section which was equivalent to the downstream half of the reactor fuel column was used to model the heat-generating portion. The simulation of the connector break at the inlet riser was made possible by actuation of high-speed valves. Six runs were made for these tests. Two runs were at 350 psi with initial flow rates of 47,000 lb/hr and test section powers of zero and 1,250 kw. Four runs were made at 1,400 psi with initial flow rates of 85,000 lb/hr and test section powers of zero and 2,500 kw. During two of these runs at 1,400 psi, the butterfly valve was closed to its stop during the flow reversal with scram delays of 1.5 and 5.0 seconds between the time of the ``connector break`` and the start of power reduction. Experimental results are given.

None

This report provides the responses to questions concerning chemical processing capability under various reactor loadings and production programs.

A Third Party inspection of the reactor work platforms was conducted by representatives of the Travelers Insurance Company in 1958. An inspection report submitted by these representatives described hazardous conditions noted and presented a series of recommendations to improve the operational safety of the systems. Project CGI-960, ``C`` & ``D`` Work Platform Safety Improvements -- All Reactors, vas initiated to modify the platforms in compliance with the Third Party recommendations. The American Standard Safety Code for Elevators Dumbwaiters and Escalators (A-17.1) is used as a guide by the Third Party in formulating their recommendations. This code is used because there is no other applicable code for this type of equipment. While the work platforms do not and in some cases can not comply with this code because of operational use, every effort is made to comply with the intent of the code.

None

This report is a summary of Hanford releases to the atmosphere of {sup 131}I, {sup 133}Xe, {sup 135}Xe, and tritium, during the period Sept. 1961 to Dec. 1963. The {sup 131}I released from the two separations plants stacks is measured continuously. Other potential sources of radioiodine to the atmosphere are also monitored, but the release rates from these other facilities are insignificant. The release rates of tritium, {sup 133}Xe, and {sup 135}Xe are not measured. The data presented for them was calculated from production data and reactor operating conditions.

The concentration dependence of sorption rates of selected gases on chabazite has been determined and interpreted in terms of the various rate controlling mechanisms present. Argon and several other gases exhibit concentration independent activated diffusion into the bulk of the zeolite. Freon 21 occupies only the pores or crevices of the mineral, and the rates are controlled by Knudsen or gaseous bulk diffusion, depending on the concentration range. Finally, carbon dioxide sorption rates strongly increase with sorbate concentrations. This probably means that a low concentrations the slow adsorption on the internal sites is the rate controlling process, while at higher concentrations the diffusion rate of the gas into the bulk of the zeolite becomes important.

This report addresses the composite diffusion mechanisms and diffusional anisotropy of helium in Beryl, Aqumarine, Quartz, and Related Glasses.

None

None

This interim report details the irradiation testing of co-product fuel-target assemblies in KER loops 3 and 4 which was authorized by PT-IP-645-D. The purpose was to demonstrate and characterize the simultaneous generation of plutonium and tritium during the irradiation of simulated N-Reactor co-product assemblies. Thirteen assemblies, each consisting of a 1.25% enriched outer fuel or driver tube and an inner, coaxial, lithium bearing target rod, were loaded in KER-3 and irradiated to approximately to approximately 1300 MWD/T tube average at N-Reactor conditions. One of the important aspects of the test was the fuel swelling performance in the large, tubular, enriched driver elements. This report describes the fuel swelling behavior of the driver elements.

This report discusses the irradiation testing of thick-walled single-tube, 1.6% enriched test elements (KSE-5`s) in KER loops 1 and 2 to evaluate swelling behavior of uranium at high temperatures. Test charge K-2-21 consisted of four KSE-5`s containing uranium with 400 ppm iron and 800 ppm aluminum. The object of the test irradiation was to establish the swelling performance of the fuel containing the iron-aluminum additions using the N-fuel composition as a reference fuel. This latter fuel has shown good high temperature performance in the British high-temperature gas-cooled reactors, but has not previously been evaluated at ``high`` specific powers in a high-temperature pressurized-water reactor environment.

A modification program is now in progress at the C and K Reactors consisting of an extensive renovation of the graphite channels in the vertical safety rod ststems. The present VSR channels are being enlarged by a graphite coring operation and channel sleeves will be installed in the larger channels. One problem associated with the coring operation is the danger of damaging top thermal shield cooling tubes located close to the VSR channels to such an extent that these tubes will have to be removed from service. If such a condition should exist at one or a number of locations in the top shield of the reactors after reactor startup, the question remains -- what would the resulting temperatures be of the various components of the top shields? This study was initiated to determine temperature distributions in the top shield complex at the C and K Reactors for various top thermal shield coolant system conditions. Since the top thermal shield cooling system at C Reactor is different than those at the K Reactors, the study was conducted separately for the two different systems.