BS>Two-group, two-region criticality calculations were made for 10 and 11 ft diameter tanks. The 10 ft tank required a core radius of 102 cm and the 11 ft tank a core radius of 95 cm for criticality. In the calculations, the fluxes were assumed to go to zero at the edge oi the graphite reflector. The fast group of the two-group calculation was broken down into 3 fast groups. The leakage out of the core and reflector for the 4 energy groups is given. (M.C.G.)

Technology of the Sodium Graphite Reactor. Reactivity calculations were made to study the application of steadystate plutonium feedback techniques to the use of diffusion plant tails for reactor fuel feed material. The performance and design of a twin core SGR power plant are given. Thermal neutron flux distribution measurements are reported for a six-rod fuel cluster and for a hollow uranium rod. A power cost calculation was made for a 1000-Mw SGR Th-U/sup 233/ breeder reactor which starts up on Th--U/sup 235/ alloy. Calculations were made on neutron leakage through the SRE shield. Research on reactor fuel elements and reactor materials are described. Corrosion and irradiation damage data at 5 x 10/sup 7/ r dose (150 deg F) on toluene as the SRE shield coolant indicate that the radioinduced corrosion of Fe, Al, and Cu in the SRE shield should be negligible. Preliminary results are summarized for l-Mev electron ir radiation studies of terphenyls at 400 to 450 deg C. Sodium Reactor Experiment. Progress is reported for various portions of the SRE project: reactor design and evaluation, fuel elements, moderator, reflector, structure, reactor cooling and heat transfer, instrumentation and control, shielding, and reactor services. (D.L.C.)

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.

During the first 15 months of operation of the Redox process, it was clearly demonstrated that in the absence of any pre-solvent extraction treatment of the starting metal solution, ruthenium contributed from 75 to 95% of the remaining fission product activity in both the final uranium and plutonium streams, and that three solvent extraction cycles were able consistently to produce, at best, only marginal quality product. This precarious position was further endangered by a three-fold reduction in the gamma radioactivity specification for recovered uranium shipped from Hanford, and by increased power levels in the reactors, resulting in still higher fission product concentrations in Redox feed solutions. The purposes of this review are to summarize briefly: (1) the chemistry of Ru in the Redox process; (2) the permanganate head-end treatment and its associated problems in plant operation; and (3) alternate possibilities for the elimination or control of Ru, including those which might solve the permanganate process difficulties. It is also the purpose of this document to present a selected bibliography on the subject of ruthenium specifically for those points under discussion herein. 17 refs.

Project CG-588 authorized the design and construction of dissolver and waste neutralizer off-gas scrubbers to remove the ammonia given off during coating removal and waste neutralization steps of the Redox operation. It has always been recognized that the nitrogen oxides in the dissolver off-gases, resulting from the dissolution of bare uranium slugs, could also be absorbed in water under proper conditions to give re-useable nitric acid. Consequently it appeared feasible to provide facilities which would combine these ammonia removal and nitric acid recovery operations. The purpose of this report is to present a scope design for the economical recovery of nitric acid from the Redox dissolver off-gases in addition to the removal of ammonia. It is recognized that acceptance of this scope for project execution would make unnecessary the ammonia scrubbers for the dissolver off-gases of Project CG-588. 8 refs.

In accordance with the policy that a centralized inventory shall be kept of all radioactive waste liquid discharged to ground, it is recommended that the volumes of condensate from the SX-farm, the dates or periods of discharge, and the activity densities of radioisotopes in the condensate discharged be determined by Separations Section and reported regularly to the Radiological Standards Unit. This paper is a description of the condensate system, with recommendations for sampling and analysis.

Boron solution circulated through special poison tubes to achieve more variable control of neutron flattening'' was tested in the 100-DR Hanford reactor. About 2700 gallons of irradiated waste Ink solution from Production Test 105-529-A was discharged to an underground crib at 100-DR, after radiochemical analyses and evaluation of radiation protection aspects by the Radiological Sciences Department. In case the Ink method is considered for production use at Hanford in the future, further biological and biophysical study is recommended to determine whether irradiated waste Ink solution may be disposed of into the Columbia River, into the ground near the river, or into the ground several miles from the river. 10 refs, 2 tabs.

A common and useful assumption is that the front-to-rear power distribution is a shortened cosine curve. This fact led to the calculation of slug factors'' for the longer charges to be used at K-Pile. It is hoped that the publication of these numbers will save time and prevent duplication of effort by those concerned in any way with slug power, surface temperatures, reactivity effects, etc. For the sake of completeness, slug factors for the older piles are included. Another useful ideal is that tube power distributions can be approximated by the assumption of a cylindrical pile and a cosine distribution of power outside of a central flattened region. This led to another double set of numbers since the values for the 2004 tube piles were again included for completeness. 2 figs., 6 tabs.

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The minimum shutdown time, to permit accessibility, for the Intermediate Power Reactor is estimated to be 38 hours. In case the reactor were shutdown following each rupture this long shutdown period would have serious disadvantages. The desirability of being able to make firm power commitments (independent of slug ruptures) has led to a study of the possibility of continuous operation following a rupture. There is evidence to indicate that, at the proposed water temperature, the rate of corrosion of uranium may be so high that at least a major portion of the rupture products may have entered the system before the reactor can be shutdown. A pushout of the affected column would then be a pushout of only those slugs which are still intact and the problem would still remain of removing the rupture products from the system. The first portion of this report is concerned with the rate of corrosion of a slug following rupture and the possible limitations to the principle of non-shutdown operation. These limitations include a flow stoppage by the ruptured can, undue increase in gamma activity, increased corrosion by the rupture products, and adherence of rupture products to the piping. The latter portion of the …

The objective of this test is to establish that scavenging of first-cycle wastes in the Bismuth Phosphate Plant will give a supernatant liquor, after the precipitate settles, that may be routinely cribbed. This test will also perform the functions: establish an effective scavenging procedure, shakedown the pH monitor, and train operational personnel. This document discusses test procedures and results.

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.

Management of first-cycle wastes from the Bismuth Phosphate Plant using Nickel Ferrocyanide scavenging is described.

This report provides details of a test to determine the feasibility of preventing corrosion of the 107- C retention basins by the installation of a small scale cathodic protection system.

This report describes the direct essential material requirements for processing a standard run through the precipitation separations process in effect on August 1, 1954. Flow sheet conditions are based on a starting maximum batch size of 300 grams of product at a uranium irradiation level of 215 MWD/ton. The essential material requirements are those used to process with 2.5 grams of Bismuth per liter in Extraction, First Decontamination Cycle volumes at 56% of the September 1, 1946 standard, and Second Decontamination Cycle through the Lanthanum Fluoride Product Precipitation volumes at 49% of this standard.

The purpose the production test described in this document was to evaluate by pile irradiation the stability and resistance to rupture of powdered metal compacts.

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From the startup of the TBP Plant until late in September, 1954, when in-line scavenging operations were begun, the wastes from the TBP Plant had been stored after neutralization and concentration in underground storage tanks. Some of this TBP waste has been given a secondary concentration in the waste concentration facilities (first cycle waste evaporators) at the tank farms. Studies by the chemistry Unit have indicated that a further reduction in the volume of waste permanently stored is possible by scavenging these wastes. In this document, a chemical flowsheet is presented for use as a design basis of facilities which will permit scavenging of these stored wastes.

The purpose of this document is to present a capacity study of the existing UO{sub 3} Plant Nitric Oxide Absorption System.

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In numerous instances the graphite growth in the older piles has deformed the horizontal control rod holes with a resultant jamming of the rods and overstressing of the rods and thimbles. In addition, special operating procedures to maintain very low differential pressures are required with the present allowed maximum graphite temperatures to prevent collapse of the thimbles because of loss of strength at this temperature. This is currently a limit to the power level of the H Pile. This report discusses a new rod tip and seal which have been developed to allow the removal of the thimble and permit sealing at the pile face. This will allow advantage to be taken of any future increases in maximum graphite temperature with proportional increases in allowable power.

During the last few months the Redox plant has been plagued with an excessive number of problems which have resulted in its inability to meet forecasted production schedules in November 1953, January and February 1954. The following paragraphs are intended to indicate some of the problems faced at Redox and the steps that are being taken to remedy them. It should be emphasized that the plant has been operating far above original design capacity and this fact tends to aggravate some of the problems. For example, although steps have been taken to increase the throughput capacity of certain of the head-end steps and some of the columns, it is most possible to install a second centrifuge, oxidizer or D-12 waste evaporator in parallel so that when any of these -- one of a kind -- pieces of equipment is out of service the entire plant is unable to operate.