This report addresses the continuous plutonium isolation contactor.

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(1) A comprehensive investigation of methods for the preparation of hydrides of wolfram has been made. A wolfram (IV) hydride-aluminum hydride mixture has been prepared and its N{sub H} and thermal stability determined for its evaluation as a nuclear radiation shield material. Aluminum borohydride has been shown to reduce wolfram (VI) chloride to a subchloride. The alkali borohydrides also reduce hexavalent wolfram, but in no case has a wolfram borohydride been isolated. (2) An investigation of the chemical and physical properties of thorium borohydride, which pertain to its use as alow-temperature nuclear radiatin shield, is presented. Values are taken from the literature when available and are supplemented where necessary by our experimental investigation.

This letter dated March 8, 1957 summarizes the results of preliminary studies on the stepwise expansion of Purex facilities to the ultimate rate of 4.0 capacity factor, using a two-cycle flow-sheet. Briefly, it was suggested that Purex be expanded to the intermediate capacity of 3.5 capacity factor using available FY-1958 funds insofar as possible; decision to expand Purex to 4.0 capacity factor, could have been deferred for subsequent budgetary consideration.

This memorandum discusses the expected life of the piles with regard to additional production capacity.

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This memorandum discusses factors involved in the additional production capacity of Hanford Works.

The objectives of the test described in this report are to evaluate the integrity of I & E MINT fuel elements canned by the ``C`` process to determine the conversion ratio of I & E geometry E-N ``striped loadings,`` and to provide a demonstration loading for a full reactor loading incorporating a ``striped`` center loading and a fringe ``blanket`` loading. This test will involve two portions a 26 tube block loading of I & E enriched uranium and MINT fuel elements arranged in a striped array, and a 10 tube fringe blanket loading of MINT fuel elements supported by the required enriched uranium fuel columns, and accompanied by controls. Either portion may be charged independently.

This Hanford-General Electric Report to the Working Committee details activities in fuel element fabrication, canning, and testing during this reporting period.

This monthly report details the 100 Area technical activities of the Physics Group for the month of May 1950.

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This document discusses a meeting held to inform members of the Biology Division of the status of P-10 problems, to exchange ideas on methods of attacking these problems, and to consider new problems.

This document discusses a production test which was written when the piles were operating under a graphite temperature limit of 410 C. Operation was controlled by as-read temperatures of the originally installed G thermocouples, by thermocouples installed on vertical safety rod thimbles, and by thermocouples in graphite cylinders installed in the C test holes. Since the thimble thermocouples were to be removed under the Ball 3X program, it was planned to install thermocouples in some of the process tube channels to help monitor the graphite temperatures. In order to ensure maximum production levels based on graphite temperatures, the proper temperature relationships between the channel thermocouples and the G and/or C test hole thermocouples were needed. The expected maximum allowable channel thermocouple temperature reading would be higher than the burnout temperature limit of the graphite in the-pile.

The design basis is composed of requirements and conditions for the design of the reactor plant (composed of the reactor and heat dissipation system). Its intent is to insure that the final product meets the economic, safety, and technical objectives of the project. The design basis is dependent on the ground rules, objectives, technical criteria, and practical design considerations. This document is being issued with the understanding that these items are not yet firmly established in all respects, and therefore, the numbers put down here are subject to change. Consideration of the spectrum of probable changes that might be made leads to the conclusion that the numbers here are close to the final ones and are satisfactory as a basis for the initial stages of design. Some numbers are omitted because of insufficient data at this time.

Seven enriched seven-rod cluster elements, three with thirty mil Zircaloy-2 jackets and four with twenty mil Zircaloy-2 jackets, will be irradiated at high temperature to an exposure of 4500 MWD/T. Coolant temperatures in the internal and external flow channels will be measured during irradiation.

Discussed is reactivity measurements on fast metal systems and moderated systems (graphite, BeO) used in rocket propulsion, pulsed neutron techniques used in graphite moderated systems, and BeO-U{sup 235} systems.

This report provides justification for a metal loader for ``C`` elevator for the 100-B,D,DR,F, and H Areas. This proposed work is justified on the basis of increased irradiation capacity for plutonium production and miscellaneous improvements in efficiency of reactor front face work.

The following calculations have been made at the request of H.R. Schmidt to determine the probability that self-radiation may play a substantial role in the decomposition of the ruthenium tetroxide in the reflux scrubber, section of the Redox Ruthenium Oxidizer (H-4). The validity of the derived data necessarily depends upon the correctness of the basic assumptions made with regard to process conditions and to possible radiation-activated mechanisms of disintegration. It is estimated that the extent of solid ruthenium formation in the tower which results from radiation effects should not exceed 100 micrograms per batch. This rate is negligible compared to that now found in the presence of stainless steel packing, or to that which might be expected from thermally-activated disintegrations alone.

A meeting was held February 5, 1954 in the 2704-Z Building to discuss the ruthenium problem in the Redox Plant and to decide on a course of action to correct the problem. The following persons were in attendance.

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This report discusses research and engineering work on the purex and redox processes at the Hanford Engineering Works in May, 1959. (JL)