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Measurement of Xenon Poisoning in the HRT

Description: Measurements obtained during three periods of HRT operations indicated that the xenon poison fraction was approximately 0.010. The technique used was based on mass spectrographic analyses of the stable xenon isotopes in the reactor off-gas stream. Models proposed to explain the measurements show that xenon, which is formed primarily by decay of iodine adsorbed on the pipe walls, is held up on the walls, out of the circulating stream, for an average period of about eight hours. (auth)
Date: April 19, 1962
Creator: Burch, W.D.
Partner: UNT Libraries Government Documents Department
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Corrosion Behavior of Reactor Materials in Fluoride Salt Mixtures

Description: Molten fluoride salts, because of their radiation stability and ability to contain both Th and U, offer important advantages as high-temperature fuel solutions for nuclear reactors and as media suitable for nuclear fuel processing. Both applications have stimulated experimental and theoretical studies of the corrosion processes by which molten salt mixtures attack potential reactor materials. Corrosion experiments with fluoride salts which were conducted in support of the Molten-Salt Reactor E xperiment and analytical methods employed to interpret corrosion and masstransfer behavior in this reactor system are discussed. The products of corrosion of metals by fluoride melts are soluble in the molten salt; accordingly passivation is precluded and corrosion depends directly on the thermodynamic driving force of the corrosion reactions. Compatibility of the container metal and molten salt, therefore, demands the selection of salt constituents which are not appreciably reduced by useful structural alloys and the development of container materials whose components are in near thermodynamic equilibrium with the salt medium. Utilizing information gained in corrosion testing of commercial alloys and in fundamental interpretations of the corrosion process, an alloy development program was conducted to provide a high temperature container material that combined corrosion resistance with useful mechanical properties. The program culminated in the selection of a high-strength Nibase alloy containing 17% Mo, 7% Cr, and 5% Fe. The results of several long-term corrosion loops and in-pile capsule tests completed with this alloy are reviewed to demonstrate the excellent corrosion resistance of this alloy composition to fluoride salt mixtures at high temperatures. Methods based on thermodynamic properties of the alloy container and fused salt are presented for predicting corrosion rates in these systems. The results of radiotracer studies conducted to demonstrate the proposed corrosion model also are discussed. (auth)
Date: September 19, 1962
Creator: DeVan, J. H. & Evans, R. B., III
Partner: UNT Libraries Government Documents Department
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Spectrophotometric studies of solutions at elevated temperatures and pressures: status and program for FY1961 and part of FY 1962

Description: A program was initiated on the spectrophotometric study of aqueous solution chemistry. The goal is operation at temperatures up to at least 330 deg C and at pressures up to 200 atm, and to near the critical point if this appears to be feasible. A spectrometer capable of operation under these extreme conditions is being designed. (W.L.H.)
Date: July 19, 1960
Creator: Biggers, R. E. & Chilton, J. M.
Partner: UNT Libraries Government Documents Department
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MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT FOR PERIOD FROM MARCH 1 TO AUGUST 31, 1961

Description: The Molten Salt Reactor Experiment design, component development, and engineering analysis is discussed. Materials studies for the Molten-Salt Reactor Program including metallurgy, in-pile tests, chemistry, engineering research, and fuel processing are described. (M.C.G.)
Date: January 19, 1962
Partner: UNT Libraries Government Documents Department
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Election of the President by the house of Representatives

Description: This report discusses the historical statistics of presidential election, the declaration of the results and other reports.
Date: August 19, 1968
Creator: Government and General Research Division
Partner: UNT Libraries Government Documents Department
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EXTRACTION OF NEPTUNIUM FROM ACIDIC SOLUTIONS BY ORGANIC NITROGEN AND PHOSPHORUS COMPOUNDS

Description: Neptunium distribution coefficients from acid nitrate, chloride, and sulfate solutions by several organic nitiogen and phosphorus compounds were measured as functions of several extraction variables, including neptunium valence, acid and salt concentration, and reagent concentration. Extractability by all the reagents varied in the order Np(IV)> Np(VI)>> Np(V). Except for primary amines, all reagents extracted Np(IV) much more strongly from nitrate than sulfate solutions. Among organonitrogen compounds the order of extractability of Np(IV) was: quaternary> tertiary> primary and secondary from nitrate solutions but primary>> secondary> tertiary from sulfate solutions. Neptunium(IV) nitrate extractions with the different extractants passed through maxima at widely different acid concentrations. In most cases, extraction increased when nitric acid was replaced by nitrate salt. Extraction was usually approximately proportional to the square of the reagent concentration. (auth)
Date: October 19, 1961
Creator: Weaver, B.
Partner: UNT Libraries Government Documents Department
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RADIOLOGICAL HAZARDS FROM RUPTURE OF THE SECONDARY COOLANT SYSTEM OF THE 10 Mw ESCR

Description: The hazards study was made to determine the radiation level from the secondary sodium lines of the l0-Mw ESCR, and to evaluate the corresponding radioactive concentration in the secondary loop and the maximum permissible effective activation flux in the intermediate heat exchanger. The results are presented graphicallyn the radiation level from the coolant lines during normal operation as a function of line diameter and Na/sup 4/ concentration; the effective thermal neutron activation flux level in the intermediate heat exchanger as a function of the Na/sup 4/ concentration in the secondary loop and the ratio of the time spent in the flux field to the time spent in making one cycle; average Na concentration in the reactor room atmosphere for operative and inoperative ventilation systems, assuming that all the sodium in the secondary system is released to the room and burns; and the downwind concentration of Na/ sup 24/ resulting from the release of all the secondary sodium on an open pad area exterior to the reactor building. An analysis of the results shows that in the event of an accident the toxicological hazards are more severe than the radiological hazards. Recommendations are given for minimizing the toxological hazards. (B.O.G.)
Date: January 19, 1960
Creator: Piccot, A.R.
Partner: UNT Libraries Government Documents Department
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Civilian Power Reactor Program. Part II. Economic Potential and Development Program. Heavy Water-Moderated Power Reactor

Description: The reactor design which forms the base for the current economic status of D/sub 2/O-moderated reactors was estimated from developments in several reactor programs. However, since a heavy water-moderated reactor was not operated on natural U fuel at power reactor conditions, considerable improvement from this current status can be foreseen. A summary of improvements is presented concerning the concept which would result solely from operation of succeeding generation plants without a parallel development program, and improvements which would result from the successful completion of the development program as presented. One plant size was used in the evaluation of plant potential, with a 300 Mw/sub e/ nominal rating. The boiling D/sub 2/O-cooled, pressure tube direct cycle plant design was used. The current development program is outlined; this work includes several items leading to the long-range development of the concept. (auth)
Date: August 19, 1960
Creator: Hutton, J. H.; Davis, S. A.; Graves, C. C. & Duffy, J. G.
Partner: UNT Libraries Government Documents Department
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Civilian Power Reactor Program. Part III. Status Report on Large (100 and 300 MWe) Heavy Water-Moderated Power Reactors--as of 1960

Description: An evaluation of 300- and 100-Mwe power plants was conducted using ground rules prescribed by the USAEC for this study. Costs corresponding to two average discharged fuel burnups are: 8.6 mills/kwh (8500 Mwd/ metric ton) and 8.8 mills/kwh (7500 Mwd/metric ton) for the 300-Mw plant. Costs for the 100 Mw plant are 14.7 mills/kwh for an average discharged fuel burnup of 6010 Mwd/metric ton. Estimates of future potential indicate that the 300 Mw/sub 3/ (8500 Mwd/metric ton) plant could produce power for 7.3 mills/kwh in a second generation, full scale plant of the same type. A further reduction to 6.4 mills/kwh should be possible as the result of the recommended ten-year development program. The current development program is adequate for providing the data needed to design and construct a prototype reactor. However, there is no natural U-fueled prototype and no prototype of the chosen reference design scheduled in the U.S. Current technology is sufficiently developed to initiate the design and construction of a pressure tube, boiling D/sub 2/Ocooled, natural UO/sub 2/- fueled reactor prototype plant in the immediate future. This plant would demonstrate the main features of a full scale plant and, in addition. would provide design data which could only be obtained by operation of a natural U- fueled reactor. (auth)
Date: August 19, 1960
Creator: Hutton, J. H.; Davis, S. A.; Graves, C. C. & Duffy, J. G.
Partner: UNT Libraries Government Documents Department
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CHARACTERIZATION OF UO$sup 2$ POWDERS. Progress Report No. 8, May and June 1961

Description: Correlation of a number of physical and chemical properties of 12 different UO/sub 2/ powders was continued. The UO/sub 2/ powders were studied by means of infrared absorption measurements, oxidation temperatures as determined by hot stage microscopy techniques, and B. E. T. surface area measurements. Additional pellets were prepared to study ceramic performance. (M.C.G.)
Date: July 19, 1961
Creator: Carpenter, J.F.
Partner: UNT Libraries Government Documents Department
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INFLUENCE OF RADIATION FROM AN UNSHIELDED REACTOR ON A NATURAL MICROFLORA

Description: The soil microflora and its respiratory activity were measured in soil cores collected at different distances from an unshielded reactor. No direct correlation was found between dose received and microbial counts or respiration. Indication of a correlation between dose and microbial respiration was obtained after eliminating the overriding influence of moisture. Radiation probably affected the soil microflora through damage to the phanerogam vegetation rather than directly. This preliminary study indicates the need to have undisturbed sampling areas close to the ORNL fast burst reactor. Extensive dosimetry, both in and above the soil, in these areas, and a program of long-term ecological descriptibn, should be started before the reactor becomes operational. (auth)
Date: October 19, 1961
Creator: Witkamp, M.
Partner: UNT Libraries Government Documents Department
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Gcr-Orr Loop No. 2 Filter Tests. Part Ii

Description: Tests of Cambridge absolute filters, Model Sl-071, specified for use in the GCR-ORR Loop No. 2 as full-flow, primary coolant fiiters were completed. kD.O.P/ (dioctylphthalate) efficiency tests were performed on three filters in the as-received condition, on two filters following canning and thermal cycling, and on one of the canned fiIters following bsking out. None of the three units met the design criteria of 99.97% efficiency for removal of 0.3 micron particles in the as-received condition. The postthermal cycle efficiencies of the canned fiIters were slightly higher than their respective as-received efficiencies. At the corapletion of testing, the two fiiters canned for installation in the reactor facility had measured efficiencies of 99.855% and 99.93%. These values were judged acceptable for the intended application/su The thermal cycling of the two canned filters and the subsequent baking out of one of these units demonstrated that a limited amount of off-gas products would be given off/su Pressure drop tests were performed on the canned fiiters with instrument air (ambient temperature, atmospheric pressure) over a flow rate range of 150 to 530 lb/hr. Curves of pressure drop across each fiIter versus Reynolds number were plotted for air and He. (auth)
Date: February 19, 1962
Creator: Flint, F. A. & Smith, A. M.
Partner: UNT Libraries Government Documents Department
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CERENKOV RADIATION INTENSITY CALCULATIONS FOR Sr$sup 90$ AND Co$sup 60$ IN WATER

Description: A method for calculating Cherenkov radiation intensity from an initial electron energy distribution is presented. The Cherenkov radiation intensity from 1 curie of Sr/sup 90/ in secular equilibrium with Y/sup 90/ in water was calculated from the beta energy spectrum to illustrate the use of the method for a pure beta emitter. The Cherenkov radiation intensity from 1 curie of Co/sup 60/ in water was calculated from the Compton electron energy spectrum to illustrate the use of the method for a gamma emitter. The steps necessary to obtain the Compton electron energy spectrum from a gamma emitter are indicated. (auth)
Date: September 19, 1961
Creator: Wymer, R.G. & Biggers, R.E.
Partner: UNT Libraries Government Documents Department
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An Evaluation of the Uranium Contamination on the Surfaces of Alclad Uranium-Aluminum Alloy Research Reactor Fuel Plates

Description: Reported radioactivity in the Low-Intensity Test Reactor (LITR) water coolant traceable to uranium contamination on the surfaces of the alclad uranium-- aluminum plate-tyne fuel element led to an investigation to determine the sources of uranium contamination on the fuel plate surfaces. Two possible contributors to surface contamination are external sources such as rolling-mill equipment, the most obvious, and diffusion of uranium from the uranium-aluminum alloy fuel into the aluminum cladding. This diffusion is likely because of the 600 deg C heat treatments used in the conventional fabrication process. Uranium determinations based on neutron activation analysis of machined layers from fuel plate surfaces showed that rolling-mill equipment, contaminated with highly enriched uranium, was responsible for transferring as much as 180 ppm U to plate surfaces. By careful practice where cleanliness is emphasized, surface contamination can be reduced to 0.6 ppm U/sup 235/. The residue remaining on the plate surface may be accounted for by diffusion of uranium from the fuel alloy into and through the cladding of the fuel plate. Data obtained from preliminary diffusion studies permitted a good estimate to be made of the diffusion coefficient of uranium into aluminum at 600 deg C: 2.5 x 10/sup -8/ cm//sec. To minimize diffusion while the plate-type aluminum-base research reactor fuel element is being processed, heat treatments at 600 deg C should be limited to 2.5 hr. The uranium contamination on the surfaces of the finished fuel plates should then be less than 0.6 ppm U / sup 235/ . This investigation also revealed that the solubility limit of uranium in aluminum at 600 deg C is approx 60 ppm. (auth)
Date: March 19, 1962
Creator: Beaver, R. J.; Erwin, J. H. & Mateer, R. S.
Partner: UNT Libraries Government Documents Department
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CORE REMOVAL COOLING SYSTEM-SECTION II. CORE I, SEED I. Test Results T- 641113. Section 2

Description: A test was performed on June 19, 1959 to determine the capacity of the Core Removal Cooling System for removing reactor decay heat under split-flow'' conditions. The system operated satisfactorily during this test; the pumps developed a flow of approximates 73 gpm at a total head of 254 ft water, as compared with their rated capacity of 75 gpm at a total head of 250 ft water. (D.L.C.)
Date: May 19, 1961
Partner: UNT Libraries Government Documents Department
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SM-1 Reactor Vessel Cover and Flange Stress Analysis

Description: The maximum stress calculated for the SMl-1 reactor vessel closure studs occurs during operation at full power. This value is 27,180 psi of which 19,800 psi is tension and 7380 psi bending. This stress does not include a stress concentration factor for effect of threads. It was eonservatively assumed the studs were initially tightened to a code allowable stress of 20,000 psi as specified in the ASME Code rather than the lesser stress obtained by the normal operating procedure. The maximum calculated stress occurs at the outside surface of the cover where the stress ranges from 318 psi in tension to 90,660 psi in compression. The alternating stress is 50,000 psi. According to the Navy Code for a stress range of 50,000 psi, the eover material ean safely undergo a maximum of 1600 cycles. It was estimated that the SM-1 will go through approximately 000 startup and shutdown cycles during a 20-yr life period, so the calculated stress is regarded as safe. For a transient eondition of 30 deg F/hr during heat-up, approximate temperature differences between the inside and outside surfaces of the cover were obtained. Temperature differentials between the inside and outside surfaces of the cover are increased by roughly 10%; above the steady state condition. More exact calculations of the transient stresses did not appear necessary siuce they would be not more than 10% greater than the steady state thermal stress. (auth)
Date: February 19, 1962
Creator: Sayre, M. F.
Partner: UNT Libraries Government Documents Department
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PREPARATION AND FABRICATION OF ThO$sub 2$ FUELS

Description: Dense partricles of ThO/sub 2/-UO/sub 2/ were prepared by a sol-gel process and vibratorily compacted into metal tubes to a density approaching 9.0 g/ cc. The steps in this method are all simple and can be carried out behind shielding, which is necessary for refabricating U/sup 233/ fuels. The sol-gel process consists of preparing a hydrous thoria sol, adding the U/sup 233/ as nitrate solution, evaporating to a gel, and finally calcining to almost theoretically dense oxide particles at orly 1150 deg C. The sol-gel-prepared oxide, after being sized, was compacted with a simple, inexpensive pneumatic vibrator. (auth)
Date: June 19, 1962
Creator: Ferguson, D.E.; Arnold, E.D.; Ernst, W.S. Jr. & Dean, O.C.
Partner: UNT Libraries Government Documents Department
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RADIOISOTOPE AND RADIATION APPLICATIONS. Quarterly Progress Report No. 13

Description: Research was continued during the report period on the use of isotope neutron sources for producing short-lived radioisotopes. Experiments with a newly constructed betacounting cell are reported in which a 50-curie Be--Po neutron source was used. Study of the radiation chemistry of polymers was continued concerning the effects of polymer structure on free-radical formation. Free-radical formation in several additional polymers was studied. Preliminary work is also reported in an investigation of internal irradiation effects on the chemical activity of catalysts. (J.R.D.)
Date: July 19, 1962
Creator: Sunderman, D.N. ed.
Partner: UNT Libraries Government Documents Department
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Research and Development Studies on Waste Storage Process

Description: The basic objectives of this program were the determination of the thermal stability of various fission product oxides and nitrates, and an investigation of the gas phase decomposition of ruthenium tetroxide. To accomplish these objectives, a literature survey was flrst made of available physical and chemical data for the oxides and nitrates of Cs, Sr, Ru, Zr, Nb, and Ce. The data were supplemented by a calculation of thermodynamic functions for RuO/sub 4/ vapor from the experimentally determined infrared spectrum and the theoretically calculated raman-active fundamentals. Data are presented graphically. (C.H.)
Date: May 19, 1961
Creator: Ortner, M. H.; Anderson, C. J. & Campbell, P. F.
Partner: UNT Libraries Government Documents Department
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