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Corrosion of 2R-2 and 304 Stainless Steel Following the Turco-4501 Decontamination Process.

Description: The build up of contaminated film on the internal surfaces of high temperature in-reactor recirculating water loops has created serious radiation exposure problems to operational and maintenance personnel. A considerable amount of work has been applied to develop an effective decontamination process for the decontamination of these loops and their components.
Date: April 20, 1959
Creator: Larrick, A. P. & Lotsinger, R. J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Effects of In-Reactor test Loops on PRTR Operation and Program

Description: Recently proposals for justifiable additions to the Plutonium Recycle Test Reactor Complex were presented to the Atomic Energy Commission at their request. In addition to a critical reactivity measuring facility in the fuel element storage basin, the following in-reactor loops were proposed: 1. A high pressure, H2O cooled fuel test loop. 2. A rupture loop to investigate fuel element failures. 3. One or more materials testing loops.
Date: March 18, 1959
Creator: Peterson, R. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Heat Transfer Study for Self-Boiling Radioactive Wastes

Description: The temperature characteristics associated with the handling of self-boiling radioactive wastes from the separations extraction processes in the Chemical Processing Department have necessitated several heat transfer studies. Earlier studies 1,2,3 defined the feasibility of self-concentration in existing waste storage facilities by determining the rate of heat generation from the decay of stored fission products and by defining the rate of heat loss from existing storage tanks to the surrounding soil.
Date: April 27, 1959
Creator: Stivers, H.W. & Taylor, H.W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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A Proposed Mechanism for the Corrosion of Aluminum in Water

Description: Data has been previously presented to show that aluminum corrosion in high temperature water may proceed with either a parabolic or a linear dependence on time. The rate of the parabolic process is an Arrhenius function of temperature and essentially independent of alloy composition. More recently several aluminum melts have been tested which corrode by a logarithmic rate process.
Date: March 19, 1959
Creator: Millon, R. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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PRTR Gas-Cooled Loop, Hazards Survey of Preliminary Scope Design

Description: The Atomic Energy Commission has recently developed an enlarged program for the study of graphite moderated, gas cooled power reactor systems. It has been recognized, however, that understanding of radiation damage and radiation induced chemical reactions of graphite at the proposed high moderator temperatures is inadequate and that improved understanding is essential if the design of such reactors is to be optimized. Accordingly, the Atomic Energy Commission requested Hanford to organize a modest research and development program directed toward the study of these graphite problems.
Date: April 20, 1959
Creator: Wittenbrock, N. G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Protection of Stainless Steel Sheathed Thermocouples from Uranium at 500 C

Description: Ceramic insulated, stainless steel sheathed thermocouples have been used to monitor temperatures of encapsulated uranium specimens, both in-reactor and out-of-reactor. No operational difficulties are encountered at low temperatures, but at a temperature of 700 C or greater, a eutectic is formed between uranium and iron. This reaction destroys protective sheath and results in thermocouple failure. A typical example of the phenomenon has been reported by J.W. Geffard of the Fuels Development Operation. Hanford Laboratories. Tantalum was suggested as a barrier between these metals and an evaluation of this system was made at 500 C.
Date: March 30, 1959
Creator: Sake, J.H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Division of Reactor Development Programs Monthly Report - February 1959

Description: Plutonium Oxide Fuels. Mixtures of PWR grade UO2 containing 10, 20, 30, 40, 50, and 60 w/o PuO2 were sintered in hydrogen for 44 hours at 1600 C to get additional data on solubility in this system. Densities of all the pieces were low, approximately 80 percent of theoretical; however, solid solution formation was complete in every case. The low density material should not affect lattice parameter values, but it did slightly reduce the intensity of the reflections.
Date: March 15, 1959
Creator: McEwen, L.H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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A Scintillation Nuclear Incident Alarm Monitor

Description: This report was written to describe the instrument and test results obtained. It is understandably imperative that such alarming devices be incorporated in various areas of the plant to provide an alarm or warning for increasing dose rate from gauss radiation fields.
Date: March 13, 1959
Creator: Spear, W. G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Photographic Observations of the Growth of Uranium Dioxide Crystals by Vapor Deposition

Description: Photographic observations of the behavior of uranium dioxide at high temperatures are of great value in designing and evaluating fuel elements. This paper reports the growth of uranium dioxide crystals by vapor deposition during out-of-reactor and in-reactor experiments.
Date: March 9, 1959
Creator: Bates, J. L. & Newkirk, H. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Simulation of an Electromagnetic Nondestructive Test on an Analog Computer

Description: The purpose of this report is to show the impedance of an approximate equivalent circuit on a complex plan for various constant frequencies and different metals. The equivalent circuit represents a testing coil driven with an AC voltage and a network to represent the impedance of the metal.
Date: March 9, 1959
Creator: Cameron, W. D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Piping Components for Organic Coolant Systems

Description: Organic compounds have been considered for use as reactor coolants for two primary reasons. First, the high boiling points of the compounds would permit the reactor to operate at high temperature without the need for the high pressure required when water is used as a coolant. Secondly, the compounds are less corrosive than water and would permit the use of carbon steel rather than stainless steel components in the reactor. Unfortunately, the organic compounds proposed for use as reactor coolants have a greater tendency to leak than water and are thus more difficult to seal. A development program was established to evaluate the types of closures required to produce a leak-tight system. This report presents the results of the evaluation. Two proposed coolant compounds, monoisopropylbiphenyl (MIPB) and the eutectic mixture of 55 per cent ortho terphenyl, 25 per cent biphenyl, and 20 per cent meta terphenyl were used.
Date: March 3, 1959
Creator: Floyd, H. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Process Vessel Precision and Accuracy Estimates

Description: The following is an attempt to explain the method by which the precision associated with an observed volume reading in a process vessel (E-12, C-1) should be calculated based upon a regression analysis of cumulative data. There are two types of volume measurement consists of estimating the total volume in a vessel at some inventory time. A transfer measurement consists of estimating the volume between two levels within a tank.
Date: March 2, 1959
Creator: Hough, C. G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Scratch Depth Measurement Methods

Description: Judging scratch depth or surface roughness by unaided visual inspection under controlled conditions, while rapid and popular, is not quantitative. Comparison methods improve reproducibility but are generally not applicable to evaluation of depths of single widely spaced scratches. Stylus-type contour recorders yield valuable scratch contour data but may themselves plow through soft materials and fine details. Depth measuring microscopes are particularly applicable to measurement of pinhole depth but do not graphically reveal profiles and provide only a small field of view. The comparatively large field of view and graphic display of contour provided by profile microscopes make them particularly suitable for evaluation scratch depth as well as surface roughness. A HAPO-constructed instrument has demonstrated an accuracy of +/- 50 micro inches in the range of 50 to 15,000 micro-inches scratch depth. It is a pocket-sized, portable, and can be used on horizontal and vertical surfaces by untrained persons with only brief instruction.
Date: February 26, 1959
Creator: Brenden, B.B.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Detailed Procedure for K Reactors Rear ace Decontamination by Chemical Flush or the Rear Crossheaders, Pigtails and Nozzles as Authorized by the Production Test Authorization IP-239-N.

Description: The purpose of this procedure is to present a detailed, chronological presentation of the preliminary decontamination and post decontamination steps necessary to fulfill the requirements of the Production Test Authorization IP-239-N. The procedure attempts to present the required operation in sufficient detail to successfully accomplish the intent of the test. Certain procedures involve operations of a standard nature and have not been elaborated upon to any great extent, as it is expected that the reactor operations and radiation monitoring personnel will implement these instructions according to standard operating procedures.
Date: February 25, 1959
Creator: Crossman, W.A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Effect of Moderator Height on Reactor and Vertical Flux Distribution in PRTR

Description: Primary control of the PRTR is achieved by regulating the level of the heavy water moderator which is held in the reactor vessel by a helium gas balance system. Emergency shutdown is effected by a gas-balanced moderator dump system which drain the moderator from the calandria at a rapid rate. This report presents a quantitative appraisal of the reactivity effects due to moderator level changes in controlling or scramming the reactor. In conjunction with the reactivity calculations, solutions were obtained which yield an evaluation of vertical flux or power distributions for any positioning of the moderator level. Coupled with the radial distributions for a given fuel loading, this information is useful in obtaining the value of the maximum specific power associated with a given power of operation and moderator height. The calculations were made using VALPROD, a one dimensional, multigroup diffusion theory reactor code programmed for the IBM-650 computer.
Date: March 3, 1959
Creator: Reginmbal, J.J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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The Stability of Three Base Metal Thermocouples in Helium-Carbon Dioxide, Carbon Dioxide and Air at 300 C

Description: In support of an in-reactor thermocouple stability testing program now in progress, laboratory studies were made relative to the stability of thermocouple bundles similar to those projected for in-reactor use. Two sets of chromel-alumel, copper-constantan and iron-constantan thermocouples were held for extended periods at 300 C in successive atmospheres of 75% helium-25% carbon dioxide. 100% carbon dioxide and air. Thermocouple stability was verified by measuring emf's at the freezing point of lead. The results of these tests are to be used to evaluate data obtained from like thermocouples subjected to condition similar to the laboratory tests with the added facet of reactor irradiation.
Date: April 14, 1959
Creator: Sako, J. H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Random Loading of E-Metal Dissolver

Description: Nuclear safety in the dissolution of irradiated 0.95 U235 enriched fuel has been investigated. In particular, critical conditions of fuel of this enrichment in a 52-inch diameter dissolver crib were studied. Since a crib this size is not safe by geometry, dissolution procedures as well as maximum safe batch sizes were analyzed. Uranium-water lattices are normally studied in systems in which rods are uniformly dispersed in the moderator. The results of such a study for 1.34-inch diameter solid rods as well as I. and E. fuel having a 1.37-inch O.D. by a 0.48-inch I.D. have already reported.
Date: February 25, 1959
Creator: Ketzlach, M.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Quarterly Report- July, August, September 1958 Plutonium Fuels Development Plutonium Metallurgy Operation

Description: Examination of Al- 1.65w/o Pu and Al- 12 w/o Si- 1.65 w/o Pu capsules irradiated 55- 60% burnout of the plutonium atoms revealed a 1.4% volume increase and no apparent microstructural changes. A four rod cluster containing Al-8 w/o Pu and Al-12 w/o Si-8 w/o Pu alloy cores is currently under irradiation in Loop 3 of KE Reactor at a water temperature of approximately 230C. A second cluster has been fabricated an is scheduled for charging late in 1958. Two seven-rod clusters for irradiation in KER are also being fabricated.
Date: February 24, 1959
Creator: Wick, O.J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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The Use of Tetravalent Uranium and Hydrazine as Partitioning Agents in Solvent Extraction Process for Plutonium and Uranium

Description: In solvent extraction purification processes such as are used at Hanford, the fuel elements or "slugs" from the reactor containing uranium, plutonium, and fission products are dissolved in nitric acid, adjusted to the required feed composition, and pumped to the solvent extraction columns. Figure 1 in a schematic diagram of such a solvent extraction process. In the A column, the uranium and the plutonium are extracted into an organic phase while the bulk of the fission products remain in the aqueous phase and leave as waste with the column raffinate.
Date: February 1, 1959
Creator: Buckingham, J.S.; Colvin, C.A. & Goodall, C.A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Radionuclides in High Temperature In-Reactor Loops

Description: A progress is in progress at the Hanford atomic Products Operation to obtain basic information about radionuclides found in in-reactor recirculating loops. This paper is concerned with the identity, amount, and source of radioisotopes found in Hanford loops.
Date: 1957-XX
Creator: Perrigo, Lyle D. Jr.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Concentration and Final Purification of Neptunium by Anion Exchange

Description: It is anticipated that neptunium will be recovered in the Purex process by solvent extraction or ion exchange methods as a nitric acid solution of greater than 0.1 g. Np/1 and containing varying amounts of fission products, plutonium, uranium, and thorium, including Th234 (UX1). At the present time this solution is thermally concentrated in the Purex L-cell package to several grams of neptunium per liter. In this operation the solution is contaminated rather badly with plutonium and stainless steel corrosion products. The present specifications are for the neptunium final product to contain less than 0.1 weight percent plutonium, to be relatively free of gross metallic contaminates, and to be low enough in fission product game activity and Th234-Pa234 (UX1-UX2) beta activity to be handled without resorting to remote techniques.
Date: February 10, 1959
Creator: Ryan, J.L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Radial Thermal Flux Traverses in Natural Uranium - Graphite Lattices

Description: The spatial distribution of thermal neutrons in a reactor lattices cell is of fundamental importance for many reasons. First of all, this information allows the determination of the relative absorption rates in each component of the cell which in turn enables a determination of the thermal utilization, and the diffusion length of the lattice. In addition, the observed spatial distributions of thermal neutrons in the lattice cell is of major interest in testing various approximations to the solution of the transport equation such as the P1 and P3 solutions.
Date: May 25, 1959
Creator: Nilson, R. & Oakes, T. J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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