This r e p o r t was prepared by Battelle-Northwest under Contract No. AT(45-1)-1830 f o r the Atomic Energy Commission, Division of Reactor Development and Technology, to summarize technical progress made in the Fast Flux Test Facility Program during November 1969.

In phase I tungsten clad cermet fuel specimens were thermal cycled, to study the effects of fuel loading, fuel particle size, stablized fuel, duplex coatings, and fabrication techniques on dimensional stability during thermal cycling. In phase II the best combination of the factors studies in phase I were combined in one specimen for evaluation.

This Systems Study is submitted in accordance with the requirements of AEC Manual Chapter 1801, and proposes the procurement and implementation of a Plant Process Surveillance System for N Reactor. The envisioned system consists of instrumentation, data acquisition equipment, a digital computer, and input/output equipment. It is intended that the system will be a tool for reliable administration of safety limits and standards, simplification of plant operation, and reduction of operating costs by replacing the tasks of shift operations personnel, an instrument technician and a process engineer.

This report was prepared by Battelle-Northwest under Contract No. AT(45-1)-1830 for the Atomic·Energy Commission, Division of Reactor Development and Technology, to summarize technical progress made in the Fast Flux Test Facility Program during August 1969 .

This report was prepared by Battelle-Northwest under Contract No. AT(45-1)-1830 for the Atomic Energy Commission, Division of Reactor Development and Technology, to summarize technical progress made in the Fast Flux Test Facility Program during December 1969.

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This report gives the status of production tests control tube loadings in reactor process tubes containing significant amounts of SS material. Data are given in table form. For further description of column headings and the current discharge goal exposure plan refer to Document DUN-3443.

While the safety status of the Richland facilities has in the past been deemed adequate all aspects of nuclear technology have progressed and evolved including standards of nuclear safety. During FY 1968 the nuclear safety R&D efforts continued along the lines and at about the schedule predicted at the beginning of the year. Hence, the safety problems to be studied have been identified and the program seems well-defined. While some program modifications and changes in emphasis are to be expected, the five-year outline shown here is considered to be a reasonable representation of the safety work of highest priority to be studied.

This report presents the details of the status of special reactor process tube loadings for the month of February 1968.

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

As commercial fuels reprocessing develops, large quantities of radioactive and stable fission rare gases will become available. These gases pose both a problem and an opportunity. When efficient methods of control and recovery are developed, the cost of these materials will be greatly reduced. With increased availability and reduced costs, commercial and government applications would be accelerated. Thus the fission gases which present potential environmental problems, would become useful by-products.

Short-time tensile tests were conducted on irradiated (2 x 10/sup 20/ nvt) and unirradiated Hastelloy-N bar from heats 5911 and 6252. No significant difference in mechanical properties was noted between the two heats. The 1200/sup 0/F ultimate tensile strength was decreased by irradiation from 65 to 80 ksi and 48 to 52 ksi. The 1400/sup 0/F ultimate tensile strength was decreased from 45 to 51 ksi to 32 to 34 ksi. The yield strength was not greatly affected by irradiation at either 1200 or 1400/sup 0/F. The 1200/sup 0/F elongation at fracture was decreased from 15 to 45% to 5 to 11%. The 1400/sup 0/F ductility was decreased from 8 to 26% to 1 to 1-1/2%. Unirradiated uniaxial stress-rupture tests were conducted at both AI and ORNL. No definite difference in stress-rupture properties between heats was noted at either 1200 or 1400/sup 0/F in ORNL tests, although some of the 1400/sup 0/F stress-rupture tests indicated that heat 6252 exhibited greater ductility. The AI verification tests showed a definite ductility difference, with heat 6252 exhibiting greater ductility. This phenomenon was attributed to an inhibition in crack propagation due to microsegregation.

This report gives the status of production test control tube loadings in reactor process tubes containing significant amounts of SS materials. The data are given in table form. For further description of column headings and the current discharge goal exposure plan refer to Document DUN-1048.

This report gives the status of production test control tube loadings in reactor process tubes containing significant amounts of SS materials. These data are given in table form. For further description of column headings and the current discharge goal exposure plan refer to Document DUN-1048.

This report presents details of the status of special reactor process tube loadings for the month of February 1967.

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This report gives the status of production test control tube loadings in reactor process tubes containing significant amounts of SS materials. The data are given in table form. For further description of column headings and the current discharge goal exposure plan refer to Document DUN-1048.

This report gives the status of production test control tube loadings in reactor process tubes containing significant amounts of SS materials. Data are given in table form. For further description of column headings and the current discharge goal exposure plan refer to Document DUN-1048.

Heat discharged to the Columbia River in reactor cooling water is one of the more important pollutants from Hanford operations. The effect of this heat on river temperatures has been studied for several years, primarily because of the potential effect on the Columbia River fisheries. This document presents a summary of recent river temperature experience, with comparison data for a better perspective.

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This report gives the status of production test control tube loadings in reactor process tubes containing significant amounts of SS materials. Data are given in table form. For further description of column headings and the current discharge goal exposure plan refer to Document RL-REA-837.

The development of the concepts for general-purpose scanning systems is discussed, based on the consideration of ultrasonics operating in liquid sodium. Technical information and data, such as ultrasonic attenuation in liquid sodium, piezoelectric materials operating in 600 F sodium, ect., are provided from theoretical and experimental studies. The report also includes explanations, analyses, and applications of the single-narrow-beam scanning systems and the electronic acoustic image conversion methods.

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