This document is an addendum to DC 56-8-167 and gives a description of the proposed Navy nuclear seaplane program and the objectives of the X211 engine study program.

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The Uranium Recovery Process (TBP Process) at Hanford extracts and decontaminates uranium from the Metal Waste produced in the Bismuth Phosphate Process. Aqueous waste, approximately equal in volume to that of the Metal Waste itself, results from the process. Although of several years' age, these wastes are still sufficiently radioactive that they must be returned to underground tanks for storage. For several years aqueous wastes of low radioactive content have been discharged to ground at Hanford. Polyvalent cations are strongly absorbed by the soil. Monovalent cations are poorly absorbed if present in solutions of high salt content. Ground waters migrate toward the Columbia River very slowly. These observations point out the desirability of removing, from wastes to be cribbed, those long-lived radioactive constituents which are poorly absorbed by soil. Cesium (Cs-137) and strontium (Sr-90) are the principal constituents of Hanford wastes which possess these characteristics. Strontium, while more hazardous biologically, is of somewhat less concern than cesium because it is better absorbed from high-salt solutions by soils. This report describes research done to develop on inexpensive process for the removal of fission products, especially cesium, from Uranium Recovery Process Wastes. 4 refs., 13 tabs.

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Regarding heat transfer studies of annular spaces around the process tube, there is general agreement among all concerned, that high graphite temperatures are essential at the initial startup. Pile Technology, in studying graphite damage, has developed information that shows little or no expansion occurring when temperatures are maintained at 275{degrees}C. Stored energy levels are less than 5% and K/Ko has saturated in the region of 3.5 to 4. Badly expanded samples (1% increase in length) have recovered 75%. This, of course, is in the pile, since nonlinear annealing is necessary, temperature alone not being sufficient. You may refer to HW-14522, HW-14310 and HW-13117 for further discussion of these points. Based on this assumption we should select some minimum startup temperature, for the edge of the active zone, say 200{degrees}C as a starting point.

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Aluminum caps and cans are thoroughly cleaned, before being used to can slugs, to insure wetting of the metal surfaces by molten AlSi in the canning pot. An acid bath is used, as part of the cleaning operations to remove surface oxide and other surface films from the metal. Two acid solutions are authorized in the standard operating procedure; a 20% phosphoric acid solution for etching both caps and cans, and a 1% hydrofluosilicic acid solution to be used for caps only. It is desired to determine the feasibility of using hydrofluosilicic acid exclusively as an etchant for both caps and cans.

Members of the Working Committees for RDA DC-4 and RDA DC-7 (KE and KW Reactors) have reviewed and approved the design basis being used for the preparation of the 200 Area portion of the Program X preliminary project proposal. This design basis includes the following: plant capacity of 275 tons per month, average flow rate; the basic chemical process used will be the Purex process; the plant will be located in the Southeast Quadrant of the 200 East Area; the plant will contain a double line of canyon process equipment, each line being so designed as to operate independently of the other line; all facilities in the 202-A Area will be designed for earthquake (Zone 2) resistance; in the design of Building 202-A, 276-A, 291-A, and 203-A, consideration will be given to achieving the highest degree of blast protection compatible with economical design, availability of data on the effects of bomb blast, and design and construction schedules; and building 211-A will include chemical storage capacity required for operation of the plant at full capacity for sixty days; building 203-A will include uranyl nitrate hexahydrate storage capacity of three 60,000 gallon stainless steel tanks; Building 241-A, Waste storage tank farm, will …

This document discusses waste disposal criteria were established on the basis that the occurrence of river flow rates which were 72 cents of normal would not cause the effects of waste disposal to exceed limits. Since even the base case exceeds the criterion for the average body burden of phosphorus-32, provision to reduce the output of this radioisotope must be included in any expansion program. Provision to reduce the output of other radioisotopes will be required for most cases where the bulk outlet temperature limit is 105{degrees} or higher. For reactor flow rates exceeding 100,000 gpm it may be necessary to reduce sodium dichromate concentrations as low as 1.5 ppM during periods of low river flow. Heat output was discussed but no limit was set.

A knowledge of the effect of reactor irradiation on the properties of UO{sub 2} is relevant to the design and evaluation of UO{sub 2} fuel elements for the PRTR. Of particular interest is the effect of reactor irradiation on the thermal conductivity of sintered UO{sub 2} and the type, mechanism, and extent of radiation damage in sintered UO{sub 2}. In the latter case the study of fractured surfaces of irradiated UO{sub 2} specimens is a method of determining the extent of radiation damage. It is the purpose of this report to propose an irradiation test of a number of capsules containing UO{sub 2} thermal conductivity and fractography specimens.

It is recommended that the actual intensity existing today above the liquid be measured before recommendations are made on cutting through the concrete seal. A detailed statement of work on this problem follows. The calculations have been filed. In the study of the feasibility of cutting through the concrete seal on the center riser of the waste storage tank the radiation intensity that would be encountered is critical because director operator contact would be required. The question of the radiation intensity to be expected was approached from two standpoints: (a) a study of previous measurements of sludge level and activities in the storage tanks; and (b) direct calculation.

This report describes the operation of the fuels preparation department for the month of September, 1957. Manufacturing employee relations, process development, plant improvements and financial operations are described.

This document details activities of the Fuels Preparation Department during the month of December 1957.

This report discusses Production Test 105-516-A which was written to determine the stability by both measurement and visual examination, of thorium slugs as used in flattening columns to exposures in excess of 400 MWD/AT. Both rolled and extruded thorium slugs were used; the highest exposures reached were about 1000 MWD/AT. Also discussed is Production Test 105-551-A which was written to demonstrate the stability of thorium slugs as used in flattening columns at exposures up to approximately 1500 MWD/AT; the supplement authorized exposures up to 3000 MWD/At. Six tubes of 10-66 material already in the pile were chosen; these have been discharged at exposures varying from {approximately}1230 to {approximately}2000 MWD/At.

This report covers the examination of thirteen process tubes seven from F Pile, three from C Pile and three from D Pile. Five of the thirteen tubes were removed because they were leakers, four from F Pile and one from C Pile. One tube from F Pile and one from D Pile were removed. Reactor Section had requested the removal of two from F Pile to check for external corrosion. Two tubes, small diameter old pile annulus, were removed from C Pile under PT 105-519-E, ``Raising Permissible Outlet Water Temperatures of Certain Tubes at C Pile.`` Two tubes were removed from D Pile under PT 105-525-E, ``Effects of Water Quality on Operations.`` Visual inspection was made of the inside and outside surfaces of the tubes before and after cleaning in a cold 10 per cent nitric acid solution. Samples varying from one to four inches in length were taken from each section for metallurgical examination to determine depth of pitting, wall thickness, and spot check the 72-S cladding thickness. These determinations were made at selected points in each section on what appeared to be the area of severest attack.

This document presents a summary of work and progress at the Hanford Engineer Works for August 1954. The report is divided into sections by department. A plant wide general summary is included at the beginning of the report, after which the departmental summaries begin. The Manufacturing Department report plant statistics, and summaries for the Metal Preparation, Reactor and Separation sections. The Engineering Department`s section summarizes work for the Technical, Design, and Project Sections. Costs for the various departments are presented in the Financial Department`s summary. The Medical, Radiological Sciences, Utilities, and General Services, Employee and Public Relations, and Community Real Estate and Services departments have sections presenting their monthly statistics, work, progress, and summaries.

Recent correlations of slug rupture data have indicated that the failure rate increases markedly with specific slug power. After the fact flux traverses have shown that a large share of the ruptures have occurred in conjunction with flux peaks significantly higher than normally expected. If more complete in-core flux information were available, it is expected that a major portion of the ruptures could be shown as having been caused by abnormally high flux peaks. However, at present.there is neither means for continuously monitoring flux distribution nor an operating control system which can be used for effectively controlling the flux distribution in the longitudinal as well as the radial direction. It appears feasible to undertake a program at this time having the specific objective of controlling and lowering the maximum to average slug power in the Hanford Reactors. There is every reason to believe that a marked decrease in the rupture rate would result from such a course of action. The purpose of this document is to systematically examine the justification for longitudinal flux control, the apparent obstacles and limitations to efficient control, and a feasible course of action designed to fulfill the desired flux distribution objectives.

It has been proposed that Pu-238 be produced by irradiating neptunium solution in one or more loops in a reactor and then recovering the Pu-238 in a close-coupled separations plant. Such a scheme could replace the more conventional scheme of solid element fabrication, irradiation, and reprocessing for plutonium and neptunium recovery. This document presents the advantages of such a scheme from the standpoint of product purity and Pu-238 production.

The Process Specifications for Operational Control of the Purex Plant(l) have been revised to define the operational control changes as a result of (a) converting the Purex Plant from three to two cycles, and (b) the use of the Plutonium Anion Exchange Unit (N-Cell) for routine plutonium concentration in place of the L-Cell Concentration Package. The flowsheets specified in Figures I-1, -2, and -3, and Table I fulfill the requirements for nuclear and chemical safety within the process as defined in Purex Process Specifications. The flowsheet also provides adequate recovery and decontamination to meet the product purity requirements itemized in Section 3.04. To clarify the two-cycle flowsheet and to simplify its use, the specifications are listed on graphic flow diagrams with the notation that process control allows a deviation of plus or minus five percent on all specifications except TBP concentration in solvent.

The purpose of this document is to provide a means for estimating the age of ruthenium ground contamination when the ration of Ru 103 activity to Ru 106 activity in the contamination is known.

This report provides recommended tube flow limitations for B and D Pile downcomers.

This report discusses procurement, fabrication, and specifications of aluminium components manufactured by Harvey Aluminium Company in Los Angeles, California.

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