This report presents the daily production data for the redox process, purex process, and uranium trioxide operations at the Hanford Engineer Works for November 1, 1958 to December 31, 1958. (JL)

The thickness of an oxide film on phosphor bronze helices was determined by first establishing the oxygen content of the helix "as received" and after cleansing with nitric acid. Based on the assumption that the difference between the two values was the oxygen in the film, and that the film consisted entirely of cupric oxide, the thickness of the film was calculated from the density of cupric oxide, weight of film, and surface area of film. A value of 1080 A was calculated as the thickness by this method. (auth)

The technical feasibility of employing uranium carbide aluminun dispersions in aluminum-base research reactor fuel elements was investigated This study was motivated by the need to obtain higher uranium loadings in these fuel elements. Although toe MTR-type unit, containing a 13 18 wt% U-Al alloy is a proven reactor component, fabrication problems of considerable magnitude arise when attempts are made to increase the uranium investment in the alloy to more than 25 wt.%. Au approach to these fabrication difficulties is to select a compound with significantly higher density tban UAl/sub 4/ or UAl/sub 3/ compounds of the alloy system which when dispersed in aluminum powder, will reduce the volume occupied by the brittle, fissile phase. The uranium carbides, with densities ranging from 11.68 to 13.63 g/cm/sup 3/), appear to be suited for this application and were selected for development as a fuel material for aluminum-base dispersions. Studies were conducted at 580 to 620 deg C to determine the chemical compatibility of carbides with aluminum in sub-size cold- pressed comparts as well as in full-size fabricated fuel plates. Procedures were also developed to prepare uranium carbides, homogernously disperse the compounds in aluminum, roll clad the dispersions to form composite plates, and braze …

The five year forecast for operation of the HAPO reactors calls for the achievement of increased process water flows in B, C, D, DR, F and H reactors. The Process Design Operation has initiated a study in support of this forecast whose objectives are: to determine present water plant and effluent system flow capabilities; to provide basic data for determining the ultimate economic optimum flow capability of these plants; and-to provide a basis for scope and development work preliminary to the initiation of any required project action. The present I&E slug program has pointed up the need for such a study of increased flows in order to take advantage of the lower system resistance of the I&E, slugs. Initial studies have indicated that considerable development work and testing is required in order to determine the most economical method of achieving increased process water flows. For this reason, CGI-815 ``Increased Water Capacity, 100-B, C, D, M, F and H`` was initiated. This interim report presents the information on the first goal of the study, namely the present capabilities of the existing water plant systems and equipment. The reactor study program has been reported separately in HW-57737. Conditions which may be encountered …

The two predominate methods of dispersing uranium in graphite are reviewed and evaluated. This study indicated that the most feasible method of dispersing uranium in graphite would be to fabricate a mixture of graphite and U/ sub 3/O/sub 8/ bonded with a thermosetting resin. A commercial type graphite was developed through independent research, and this fabrication procedure was adapted for the manufacture of the TREAT fuel matrix. (auth)

Last year (1948) during the flood period the flow at Grand Coulee fluctuated widely. 2 PM, June 8, 543000 c.f.s.; 4 AM, June 9, 568000 c.f s.; 2 PM, June 9, 543000 c.f.s.; 2 AM, June 10, 573000 c.f.s. A total instantaneous fluctuations of 37,500 c.f.s. was reported. Now there is installed a new control. This control can keep downstream variation within 500 c.f.s. By lowering the lake level prior to the crest period, the drum gates could be used as flood control (1948 high water basis) the drum gate control plus the water turbine discharge (if the lake level had been reduced) could have dropped the crest at Richland three feet. a. Drop in crest at Richland one foot: Electrical loss nominal, b. Drop in crest at Richland two feet: Electrical loss 1 megawatt/foot for six generators. Loss Max possible 13,310 KW each generator, 79,860 KW total (7 days). Capacity 1,170,000 KW Max Loss 6.8% for 7 days to 10 days. c. Drop in crest at Richland three feet: Electrical loss 1 megawatt/foot for 6 generators Max possible 30,100 KW each generator 180,600 KW total 8 days. Capacity 1,170,000 KW Maximum loss 15.4% for 8 to 12 days. Actual …

A maximum leak rate of 0.08% was measured for a piston ring seal assembly which was evaluated for use as the Hallain Power Reactor process tube- grid plate seal. A maximum leak rate of 0.14% was observed after subjection to 10,000 cycles (560 hr) in 625 deg F Na. The maximum leak rate was 0.07% after 25 cycle exposure in 1000' Na. Vertical scoring of both the rings and bore tube was observed. Sodium was observed to remain behind the rings after washing. (C.J.G.)

A preliminary report on development of filter paper for use in A.E.C. operations is presented. Filters for use in removing harmful dusts or radioactive matter from the air discharged from various operations or for other uses are described. The main fiber furnished for the paper is a specially treated wood pulp with an addition of asbestos. Filters of higher or lower efficiencies with corresponding changes in static resistance can readily be made by modifying the manufacturing formula. (J.R.D.)

A Task Force was established in the Irradiation Processing Department to examine the K-Reactor Water Plant to (1) review the operating and maintenance experience with the water plant as improved since startup, (2) identify major plant additions which could further improve reliability, and (3) estimate the costs of any such additions. The K-Water Plant basically consists of the electrically driven primary cooling system with power supplied by the BPA system, electrically driven secondary or backup cooling system powered by a steam driven emergency generator pair, and a ``last ditch`` system consisting of hydraulic cross-ties between the two K-Water Plants. This report summarizes information developed in the course of the Task Force deliberations.

This report discusses the interim program for the production of neptunium oxide at HAPO which applies the following processing steps: isolation of neptunium from the Purex process streams, using Purex flow sheets specially adapted for this purpose; purification of the neptunium nitrate by an ion exchange process carried out in one of the Redox laboratory (222-S) multi-curie cells; and precipitation of neptunium oxalate and conversion of the oxalate to oxide in laboratory-type equipment. The process, being still in the developmental stages, is as yet subject to extreme fluctuations, both conditions and results.

None

Based on the activation of gold foils in an hydrogeneous medium, the neutron flux incident on the rear wall of the discharge area of the KE reactor is estimated to be 6000 M/cm{sup 2} sec. The effective energy of the neutrons is estimated to be approximately 4 Mev. Neither of these values confirm order-of-magnitude estimates of the neutron flux and neutron energy expected to exist in the discharge area.

The 0.045 m UO/sub 2/SO/sub 4/, 0.036 m CuSO/sub 4/, 0.025 m H/sub 2/SO/ sub 4 solution (HRT fuel composition) was chemically stable during 1,866 hr of operation at 280 C and 1500 psi. The system was pressurized by boiling a 0.4 gpm stream of the fuel in a titanium heat exchanger at 313 C. During cursions were made to pressurizer temperatures above 330 C where two liquid phases were formed. These tests indicated that heavy phase began formation at 325 C (vapor pressure equilibrium temperature) as evidenced by loss of fuel from the circulating stream, Good heat transfer excluded the possibility of the missing material depositing in the form of a scale in the heat ex hanger. In each test the original fuel composition perature was lowered below 325 C. The generalized stainless steel corrosion rate during operation at 280 C and 1500 psi was 0.6 mpy for the first 700 hr and 1.6 mpy for the next 1166 hr. The average rate during the period when excursions were made into the two-phase region was 3.0 mpy. The apparent increase in corrosion rate is not easily explained because no unusual attack could be found on inspection of the stainless …

for filtration of plutonium oxalate slurries. A scalpel produces a slit in the filter precoat, leading to increased filtration in this slit, and the oxalate is removed by a doctor knife; this technique results in prolonged blowback cycles and more uniform delivery of filtered oxalate to subsequent processing steps. Several types of filter media were tested, and rigid porous aluminum oxide was found to be the best one. (D.L.C.)

Two coextruded, Zr-2 clad, natural uranium, seven rod cluster fuel elements were irradiated to a calculated exposure of 1250 MWD/T in the KER Facility and discharged 1-16-59. The fuel elements were NPR candidate fuel and examination was requested to determine the behavior of coextruded, Zr-2 clad, natural uranium irradiated at core temperatures of approximtely 425{degree}C. The elements were transferred to the Radiometallurgy Laboratory 2-25-59. The elements demonstrated excellent in reactor performance with no significant changes in either the fuel or the hardware. Detailed examination of the central rod and two peripheral rods from one of the clusters showed no microcracks in the uranium. Moderate growth was observed in the fuel at the unrestricted rod ends.

Data on waste volumes and heat generation of several reactor fuels which may be reprocessed in the Power Reactor Fuel Reprocessing Pilot Plant at ORNL are tabulated. (auth) l6876 A tabulation containing information on the power of existing and proposed U. S. and U. S.-built reactors of 10 kw or greater thermal power is presented. Estimated fuel reprocessing loads for irradiated fuels are also iucluded. (auth)

The objective of this production test is to determine the changes that occur in the uranium-Al-Si bond during irradiation at bond temperatures between 255 and 285 C. Twenty-five M-388 jacketed dip canned depleted uranium solid fuel elements will be irradiated to an exposure of 500 MWD/T in high temperature water. The location and size of unbonded areas on the fuel elements will be measured by ultrasonic mapping before and after irradiation to show the changes in bonding resulting from irradiation at high temperature.

A method is sought for predicting reactivity differences between fuel elements attached to control absorbers and fixed fuel elements. The approach described is based on a logic that is approximate, and which should be subjected to experimental check. (auth)

The solubility of lithium hydroxide in water was determined at 220 to 650 F. The literature furnished data for temperatures below 200 F. A maximum in the curve was found at about 240 and a minimum at 480 F. The variations in solubility, however, were relatlvely small. At 40, the solubility is 12.7 g LiOH per 100 g H/sub 2/O, while at 240, it is 17.7, and at 650 F, it is 16.5. The vapor pressures of 4.76 wt. % (2.09 molal), 8.59 wt.% (3.92 molal), and saturated (approximately 6.25 molal) lithium hydroxide solutions were measured as a function of temperature. At about 685 F, the more dilute solution showed a depression in vapor pressure of about 130 psi, the intermediate 154 psi, and the saturated 158 psi. The more dilute solution showed a greater deviation from Raoult's law than did the other two. Vapor-pressure data for sodium hydroxide solutions were compared with those for lithium hydroxide of similar concentration by weight and molality. (auth)

The objective of this Supplement described in this report to Pt-IP-250-A is to d enriched tube-and-tube elements will develop pitting corrosion on the Zircaloy-2 jackets when irradiated in pH 10 water. The measurement of dimensional changes in the fuel elements and the observation of the effect of irradiation on the uranium and bond area are also objectives of the test, but secondary in importance to identifying a pitting corrosion problem in NPR quality water, if one exists.

Calculations indicate that single-transfer line concatenated pulse column systems can be operated with static pressures that are not excessive if a sufficient number of vessels are employed in the system. The required number of vessels can be attained by using a series of short columns or by using holdup pots in conjunction with a limited number of columns. General equations for calculating pressure drops and power requirements are presented. (auth)

A discussion is given of the role of thermodynamics in the fused-salt volatility process, particularly as it applies to oxidation-reduction reactions affecting zirconium, uranium, nickel, chromium, ruthenium, and other elements present in the hydrofluorination head-end step. (auth)

The objectives of this study were to demonstrate that the pump speed control system will respond as defined in the equipment specifications and to determine optimum values of controlling variables that will minimize the oscillations that occur in the Na flow rate when transient signals are imposed on the pump speed control system. (W.L.H.)

The failure of defected Zircaloy-2-clad uranium and uranium -2 wt.% zircorium fuel specimens in high-purity high-pressure water at 200 to 345 deg C was observed in a windowed antcclave. Time-lapse color motion pictures were taken to provide a record of the progressive changes ending in the complete disintegration of the core material in the specimens. Continuous measurement of the pressure increase caused by accumulation of hydrogen served to monitor the progress of the reaction when clouding of the water by corrosion products made visual observation impossible. The nature of the attack of all specimens was similar, although the time at which different stages occurred varied. Following an induction period, the first evidence of attack was the slow formation of a blister in the cladding area surrounding the defect. Eventually, a copions evolution of hydrogen occurried at the base of the swollen area. In general, a crack could be seen in the cladding at this stage. Catastrophic failure of the specimen followed swiftly. The time required for each phase of the reaction was reduced as the temperature was raised. Initial swelling occurred after about 24 min at 345 deg C but only after 8 hr at 200 deg C. Diffusion-treated uranium2 …