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Radioactivity Dissemination Near Uranium Processing Mills

Description: This report follows a preliminary survey made to study radioactive contamination of the soil (and to some extent in the air and waters) in the vicinity of seven uranium processing mills.
Date: April 1, 1961
Creator: Feldman, M. H.; Troianello, Emilio J.; Coates, G. K. & Sheehan, W. R.
Partner: UNT Libraries Government Documents Department
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gem-Bis(disubstitutedphosphinyl)alkanes. II. Extraction Properties of Bis(di-n-hexlphosphinyl)methane

Description: From abstract: "Bis(di-n-hexylphosphinyl)methane, HDPM, [(C6H13)2P(O)]2CH2, has been studied as an extractant for a variety of metals. HDPM was evaluated as an extractant for uranium(VI) and compared with tri-n-octylphosphine oxide, TOPO, (C8H13)3PO. In nonpolar solvents, HDPM forms a polymeric-like substance with compounds of uranium(VI). Viscosity measurements indicate that the molecular weight of this polymeric-like substance is about 100 times greater than the corresponding complex with TOPO. Polymer formation occurs only when nonpolar solvents are used as diluents for HDPM and is easily avoided by using polar solvents such as 1,2-aichlorobenzene. HDPM forms 1:1 and 2:1 complexes with uranium(VI) nitrate. Equilibrium constants for these complexes as well as that for the 2:1 TOPO complex were calculated and it was shown that the over-all constant is at least ten times larger for the HDPM complex than for the TOPO complex. The effect of concentration of various mineral acids, extractant concentration, temperature, and diluents on the extraction of uranium are discussed."
Date: April 11, 1961
Creator: Burke, Keith E.; Sakurai, Hiroshi; O'Laughlin, Jerome W. & Banks, Charles V.
Partner: UNT Libraries Government Documents Department
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Hazards Report for the SM-1 Core II Without Special Components

Description: Abstract: This technical report describes the changes incurred in the SM-1 by the insertion of the SM-1 Core II without special components. The SM-1 Core II components were made to specifications very nearly identical to those of SM-1 Core I. The differences consist of europium absorber sections, internal europium flux suppressors in the control rod fuel elements, and low impurity cladding. Each of the SM-1 Core II components with the exception of the five absorber sections new in SM-1 Core I were subjected to a Zero Power Experiment at the Alco Critical Facility. The results of this experiment indicate that the SM-1 Core II will have nuclear characteristics very similar to that of the SM-1 Core I. Since SM-1 Core II will be operated with the same mode of rod control, in the same core support structure, and with the same primary coolant flow conditions, the thermal characteristics should be essentially identical to that of SM-1 Core I. Also, all kinetic characteristics of SM-1 Core II should be identical to those of SM-1 Core I. This report demonstrates that there is no increase in potential for a hazardous situation at SM-1 due to the replacement of SM-1 Core I by SM-1 Core II.
Date: April 19, 1961
Creator: Gallagher, J. G.
Partner: UNT Libraries Government Documents Department
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Project CGC-897--Title I design, fission product storage in B-Plant

Description: A previous study described proposed facilities at B-Plant which integrate future fission product and waste calcination activities. However, in the reactivation of B-Plant in accordance with this study, heavy expenditures, above budgeted funds, would be required at an early date for Phase 1 process changes coupled with general rehabilitation work and facilities for updating of radiological control. Since waste calcination activities in B-Plant are not scheduled until Fiscal Year 1966, the expense of B-Plant rehabilitation items would be borne solely by the Fission Product Program. This report provides the Title I design of Phase 1 fission product facilities at B-Plant which can be provided vith minimum capital expenditures. The facility described in this report accomplishes the overall processing objectives of the facility, namely the recovery and storage of crude strontium-90 and rare-earth concentrates, although certain B-Plant improvements are deferred to later phases of the Fission Product and Waste Calcination Programs.
Date: April 3, 1961
Creator: Caudill, H. L. & Zahn, L. L. Jr.
Partner: UNT Libraries Government Documents Department
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Hanford Laboratories Operation Monthly Activities Report: March 1961

Description: This is the monthly report for the Hanford Laboratories Operation, April 1961. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.
Date: April 15, 1961
Partner: UNT Libraries Government Documents Department
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C-Reactor graphite burnout interim report, 1P-25A(PT-105-532-E)

Description: The oxidation of graphite in the Hanford reactors is of consequence since graphite burnout affects the strength of the moderator. As a means for indication of any highly oxidizing condition within the stack, containers or boats of small weighed samples of reactor-grade graphite are positioned along the length of an empty process channel in each reactor. The rate of oxidation of the monitoring samples, referred to as the burnout rate, is reported as percent weight lose per 1000 operating days (%/KOD). Currently the burnout rate limit is 2%/KOD. This document presents recent burnout data at the C-reactor.
Date: April 5, 1961
Creator: Ryan, B. A.
Partner: UNT Libraries Government Documents Department
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PCCF flow analysis -- DR Reactor

Description: This report contains an analysis of PCCF tube flow and Panellit pressure relations at DR reactor. Supply curves are presented at front header pressures from 480 to 600 psig using cold water and the standard 0.236 inch orifice with taper down stream and the pigtail valve (plug or ball) open. Demand curves are presented for slug column lengths of 200 inches to 400 inches using 1.44 inch O.D. solid poison pieces (either Al or Pb-Cd) and cold water with a rear header pressure of 50 psig. Figure 1 is a graph of Panellit pressure vs. flow with the above supply and demand curves and clearly shows the effect of front header pressure and charge length on flow.
Date: April 26, 1961
Creator: Calkin, J. F.
Partner: UNT Libraries Government Documents Department
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Post-irradiation examination of bumper elements with high in-reactor weight losses (RM-418)

Description: This report discusses three natural uranium, X-8001 aluminum clad, I&E Hanford production fuel elements, which were irradiated in 3363-D as part of PT-IP-262-A, were selected for detailed examination in the Radiometallurgy Laboratory. The three pieces were from the same tube and each had lost about 15 grams of cladding during irradiation. Examination was requested to determine the extent of the corrosion and whether the attack was uniform or localized. Also, measurement of the uranium fuel was requested to reveal any change that occurred during irradiation. Corrosion was general rather than localized and occurred over approximately three-fourths of the surface. In each element. about one-fourth of the surface on one side was virtually unattacked and vas probably the area that lay between the ribs of the process tube during irradiation. In one element localized attack occurred beside two of the bumpers. External aluminum cladding thicknesses ranged from 0.020 to 0.043 inch. About 0.005 inch of the spire surface vas removed by corrosion. Both internal and external dimensions of the uranium increased. The average external diameter was 0.010 inch larger and the average internal diameter vas 0.011 inch larger than the average preirradiation diameter measurements. The growth vas not uniform as ellipticity up to 0.028 inch was observed.
Date: April 3, 1961
Creator: Gruber, W. J.
Partner: UNT Libraries Government Documents Department
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Offsite Extrusion Program summary report

Description: The coextrusion process for fabricating zircaloy-2 clad metallic uranium fuel elements was developed by Nuclear Metals Incorporated, West concord, Massachusetts. When the coextrusion process was selected for fabricating NPR fuel elements, it became necessary to acquire experience in this process. Since there were no extrusion facilities immediately available at HAPO, it was necessary to initiate a development program using offsite extrusion facilities. Two basic objectives of this program were: (1) To acquire training in the art of coextrusion and, (2) to produce fuel material for the HAPO process development program. The extrusion work was performed at Nuclear Metals and at Bridgeport Brass Co., Adrain, Michigan. At Nuclear Metals, a 1000 ton, direct drive, Watson-Stillman rod press was used. At Bridgeport Bass, a 2200/200 ton, water accumulator, external manipulator, three column (Brooklyn Model) Loewy Hydropress was used. A total of 128 coextrusions were made. The date, location, description of material extruded and pertinent reports are summarized. Detailed data, description and evaluation for each extrusion run is reported in Appendices I--IX inclusive. Significant observations and conclusions are summarized in each appendix.
Date: April 1, 1961
Creator: Nickolaus, J. W. & Guay, A. E.
Partner: UNT Libraries Government Documents Department
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Irradiation testing of tubular fuel elements: PT-IP 292A. Final report

Description: This report discusses Zircaloy-2 clad uranium and uranium-2 weight percent zirconium fuel tubes which were irradiated to 3200 MWD/T in a high temperature water cooled loop. The outer clad of one tube split due to swelling of the uranium. Postirradiation examination of the fuel cores included metallography, electron microscopy, density determinations, dimensional measurements, and radiochemical burn-up analysis.
Date: April 1, 1961
Creator: Geering, G. T.
Partner: UNT Libraries Government Documents Department
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Post-irradiation examination of chemically nickel-plated fuel elements from PT-IP-263-A (RM-414)

Description: Two chemically nickel-plated, internally and externally cooled, Hanford production fuel elements, which were irradiated to approximately 800 MWD/T as part of PT-IP-263-A, were transferred to the Radiometallurgy Laboratory in December 1960. The elements were selected for detailed examination because one had incurred a hot spot during irradiation and the other contained some unusual cracks in the nickel plate. Prior to irradiation, both fuel elements had been baked at 300 C to heat-treat the nickel plate. Also, the nickel plate of several unirradiated elements was damaged by scraping, marring, scratching and punching. The elements were exposed for six weeks to 105 C basin water, which was approximately the length of time the irradiated elements were in 105 C basin prior to transfer. Two unirradiated elements were submitted for comparison with irradiated pieces. The examination was requested by Process Engineering, Fuels Preparation Department; and Process and Reactor Development, Irradiation Processing Department, to determine the effects of irradiation on elements with improved nickel plating and to aid in evaluating the nickel-plated fuel element program.
Date: April 17, 1961
Creator: Gruber, W. J.
Partner: UNT Libraries Government Documents Department
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Production test IP-310-A-FP, determination of the dimensional stability of uranium fuel cores classified by the fuel core tester (UT-2)

Description: Since it is now possible to obtain a heat-treated U core that is randomly oriented and has a finer average grain size, it is necessary to irradiate measured, transformed fuel cores over the full range of grain sizes, in order to compare relative dimensional stabilities. An improved ultrasonic tester, Fuel Core Tester-UT-2, is used to test all fuel cores.
Date: April 27, 1961
Creator: Clinton, M. A.
Partner: UNT Libraries Government Documents Department
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40-tube overbore facility location, C Reactor

Description: Possible locations of the projected 40-tube overbore facility at the C Reactor are discussed from the standpoint of obtaining conversion ratio data applicable to a full-reactor overbore program.
Date: April 19, 1961
Creator: Nilson, R.: Nechodom, W. S.
Partner: UNT Libraries Government Documents Department
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Loading and operating conditions for PT-IP-401-A and PT-IP-363-A in KER-1

Description: IP-401-A authorized the irradiation of 18-inch UO{sub 2} elements and IP-363-A authorized 18-inch KSE-3 elements. This document provides specific loading and operating conditions for a charge of one UO{sub 2} element and two KSE-3 elements in KER-1.
Date: April 5, 1961
Creator: Kratzer, W. K.
Partner: UNT Libraries Government Documents Department
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Process Water Flow Tests, C Reactor

Description: The power calculator at C Reactor and the process pump flowmeters at 190-C use orifice plates as primary sensing elements. These orifices were designed for considerably less flow and at the present higher flows, the reactor riser pressure is appreciably reduced as the result of the inefficiency of the orifices. Only a portion of the differential pressure across an orifice is recovered. The amount is dependent upon the ratio of the orifice diameter to pipe diameter. The permanent loss across these two orifices is 16 psi. The permanent pressure loss resulting from properly sized venturi elements would be 1.6 psi, for a resultant decrease in system pumping resistance of 14 to 15 psi. This would permit higher process water flow and reactor power levels.
Date: April 11, 1961
Creator: Hamilton, W. D.
Partner: UNT Libraries Government Documents Department
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In-tank solidification of intermediate-activity wastes

Description: Solidification of intermediate-activity wastes is a major goal of the CPD Waste Management Program. Plans are to reduce the wastes, by evaporation, to salt cakes in existing tanks, thereby insuring safe, long-term storage of contained fission products regardless.of tank integrity. Initiation of these plans at an early date is necessary to offset the expected increase in tank failures and to provide space for future wastes. Major decisions of the program relate to selection of the evaporative method to be employed. The requirements of in-tank solidification were therefore reviewed to determine if the choice of evaporative systems can be made at this time. The relative potential of Bentube evaporation and submerged combustion for meeting these requirements were analyzed on the basis of available information, including actual performance of the Bentube facility at the Savannah River Plant (SRP).
Date: April 3, 1961
Creator: Campbell, B. F.
Partner: UNT Libraries Government Documents Department
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Processing data for NIE and NIN KER Loop charges (PT 377)

Description: The following data represents processing conditions used in fabricating prototypic (except for supports) natural and enriched NPR inner fuel elements for Production Test 377. The prototypic NPR inner fuel elements were inserted into zircaloy-sleeves prior to charging. In this way it was possible to simulate, in the KER loops, the conditions under which NPR fuel elements would be subjected under irradiation. The purpose in documenting the data is to provide a permanent record of processing conditions and dimensions which may be referred to for post irradiation analysis and possible future process development work. Post irradiation results will be issued by the Fuels Development Operation, Hanford Laboratories Operation, and the test loop operating conditions will be issued by Process and Reactor Development Operation, Irradiation Processing Department, as outlined in the Production Test Procedure.
Date: April 19, 1961
Creator: Kusler, L. E. & Hays, D. D.
Partner: UNT Libraries Government Documents Department
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Proposal for charging GEH-10-43 NINI interim measurement test in the ETR

Description: The objective of this test is to irradiate an NPR fuel element inner tube in the ETR 3 {times} 3 Loop. the NPR fuel element is a tube in tube type. Both tubes are Zircaloy-2 clad uranium metal. The inner tube is identified by the Hanford designation NINI. In GEH-10-43 we plan to measure an NINI tube after each cycle for four cycles. An NPR inner tube with slightly larger diameter has been run in the ETR Hanford 3 {times} 3 Loop. Two NINI elements are now running in the 3 {times} 3 Loop.
Date: April 7, 1961
Creator: Geering, G. T. & Heck, E. N.
Partner: UNT Libraries Government Documents Department
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Supplement G, production test IP-314-A, measurement of fuel element temperature changes as the result of film deposition

Description: Irradiation of a third thermocouple train with a thermocouple element of the same design as used on the previous two trains, but with different train design and heater elements, is authorized to an exposure no greater than 1000 MWD/T. A third decontamination of KER-1 is also authorized.
Date: April 25, 1961
Creator: Kratzer, W. K.
Partner: UNT Libraries Government Documents Department
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Supplement A to PT-IP-183-A-98-FP: Evaluation of projection fuel elements for use in K process tubes

Description: The objective of this supplement is to authorize charging of ten tubes of ``bumper`` fuel elements and controls into each K Reactor. The test is designed to reevaluate the reduction in hot-spot incidence associated with fuel alignment within K Reactor ribbed process tubes for both natural and enriched uranium I&E fuel elements of the KIV geometry.
Date: April 10, 1961
Creator: Clinton, M. A.
Partner: UNT Libraries Government Documents Department
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