The PRTR is entering a phase of operation in which there is an increased danger of high level fission product release from failed fuel. The higher fuel operating temperatures result in a much larger percentage fission gas release from the fuel crystal; also it is believed that the central voids and inter-crystal cracks greatly improve the transport of fission gases from the far end of a fuel rod to a defect location. 6 tabs.

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This research and development report details activities of the Irradiation Processing Department during the month of August 1965.

An evaluation of Hanford reactor process water treatment practices was made in an effort to ascertain the reasons for variations in the effluent activity between reactors. Recommendations are made for improvements in unit processes as well as for the over-all treatment process based on field inspection of the water treatment plants. In addition, a research program is proposed to supplement the recommendations. The proposed research is designed to uncover methods of more efficient filtration as well as other procedures which might eventually lead to significant effluent activity reductions. The recommendations and research results will be applied toward process optimization.

This Hanford Report to the Working Committee details activities and research programs in current reactor production fuels, N-Reactor fuels, N-Reactor fuel development, and current reactor fuel development.

Plant status, reactor physics, technical activities to include fuel element comparisons, ceramic targets, fuel element testing, alignment studies, tube removal, and nuclear safety specifications are all considered in this progress report.

Five monitor columns, containing some pre-irradiation weighed inner and outer fuel elements, were discharged under authorization of PT-NR-4, Sup A at tube average exposures in the range from 1554 to 1631 MWD/T. Uranium swelling values have been calculated from post-irradiation weight measurements and these swelling values have been compared with those swelling values obtained in the KER loop tests. There is acceptable agreement between KER and NPR results for swelling of inner uranium; there is acceptable agreement for outer fuel elements at higher exposures but ends-of-the-column elements (i.e. low-exposure values) are not in agreement. The current swelling values have been extrapolated to determine the tube-average exposure that will result in outer clad strains of one percent. At this conservative value of clad strain, one can predict entirely adequate performance of outer fuel elements to tube average exposures of 2200 MWD/T and inner fuel elements to 2700 MWD/T tube average exposure. Elements that will soon be discharged at tube average exposures ranging from 1900 to 2100 MWD/T, can be evaluated for swelling during the next outage and a much better prediction can be made on the relation of exposure vs swelling and hence exposure vs fuel element integrity. The continuing objective …

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This report covers the activities with regard to on-site customer irradiations as of September 10, 1965. The report covers the status of material undergoing irradiation, awaiting disposition, material shipped during the month, and current status of all reactor test holes.

This document consists of tabulated calculations of the energy budget for the Hanford production reactors. Examples of the heat transport calculations are illustrated along with the complete final data. Travel times are estimated.

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The radionuclides in reactor effluent are produced almost exclusively from the irradiation of patent elements which are adsorbed on the film which forms on the in-reactor surfaces. The nuclides are subsequently desorbed into the effluent water, one of the areas of activity in the research and development program directed toward reducing reactor effluent radioactivity has been to find some means of minimizing the adsorption of parent materials and/or desorption of radioisotopes. In the course of these investigations, it was shown in laboratory coupon tests that sodium silicate added to process water reduced the adsorption of arsenic by a factor of five. A single-tube in-reactor test was run using a silicate addition of 20 ppm SiO{sub 2} to give some indication of the effect of flux and other reactor variables on the behavior of the silicate. The results from the test showed that the concentration of P{sup 32}, As{sup 76}, and Np{sup 239} in the effluent were reduced by a factor of two to three. Lesser reductions were obtained for other isotopes, and the Na{sup 24} concentration increased by a factor of two. This report summarizes the data obtained and observations made during the six months the test has been in …

Brief summaries are given for the status of activities in the following areas: Reactor fuels; Reactor engineering; Reactor physics; Radiological engineering; Process standards; Process change authorization; Fuel failure experience; Pile physics assistance; Process physics studies; Testing of fuel elements; and Testing and irradiation services.

This report provides the status of production test control tube loadings in reactor process tubes containing significant amounts of SS materials.

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It is shown experimentally that transistor circuits may have their fundamental sensitivities to ionizing radiation reduced by factors of ten by addition of a simple reverse biased diode. Bipolar transistors, field effect transistors and microcircuits have been hardened. (auth)

An investigation of the ultimate compressive strength of fuel elements as a function of temperature was conducted. A series of tests were performed on a group of two inch long fuel elements with maximum fuel loading. The tests were performed over the temperature range of 3600°R to 2500°R. The maximum increase in ultimate compressive stress was approximately 33 percent of the ambient temperature (530°R) ultimate stress of 12,000 psi and occurred in the temperature range from 3800°R to 4000°R. At temperatures higher than 4600°R, the ultimate stress decreased rapidly and noticeable plastic deformation was observed.

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An analysis is made to predict the behavior of a nuclear rocket engine in the event of an in-flight failure subsequent to stantup of the nuclear stage. The accident model assumes loss-of-coolant to the reactor at time of system failure. It is also assumed that the only direct action taken subsequent to failure is separation of a re-entry, vehicle from the remainder of the stage. This vehicle consists of the reactor, pressure vessel, and nozzle. The behavior of the vehicle during its residence time in space and during its re-entry into the Earth's atmosphere is examined and described for three different missions. Two of these are lunar missions employing a NERVA-I class engine, and the third is a Mars fly-by using a NERVA-II class engine. The potential exposure of the Eart's population to nuclear radiation is presented for each of the three missions. (auth)

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