Studies of reflector control for the Advanced Engineering Test Reactor were made. The performance of various parts of the reflector control system model such as the safety reflector and the water jet educator, boric acid injection, and demineralizer systems is discussed. The experimental methods and results obtained are discussed. Four reflector control schemes were studied. The schemes were a single-region and three-region reflector schemes two separate reflectors, and two connected reflectors. Calculations were made of shim and safety reflector worth for a variety of parameters. Safety reflector thickness was varied from 7.75 to 0 inches, with and without boron. Boric acid concentration was varied from 100 to 2% of saturation in the shim reflectors. Neutron flux plots are presented (C.J.G.)
A concentration cell electrode was installed in a fritted glass surface and was used to measure the surface KCl concentration while water was being transpired through the surface into a mixed tank of 0.1 M KCl. The results from the first Fluorox run made with crude UF/sub 4/ showed that 85.3% of the theoretical amount of UF/sub 6/ was accounted for, with 17.9% being collected in cold traps and the remainder being consumed in various side reactions. Installation was completed of apparatus to study the electrolytic removal of nickel on an engineering scale from concentrated HRT fuel solution. An extremely low yield stress of 0.02 lb/sq ft was measured for a thoria-urania slurry containing 2.5 wt.% Al/sub 2/O/sub 3/ and spherical particles of 1.7 micron mean diameter. Preliminary data on the rate of uranium sorption on Dowex 21K from sulfate solutions were correlated with a simple spherical diffusion model. Laboratory scale studies to determine the effect of reflux time, HNO/sub 3/ concentration, and air sparge on chloride removal from APPR-type Darex dissolver product showed that an air sparge is definitely beneficial in chloride removal. Preliminary cyclic cleaning tests showed that boiling 25% caustic solution was adequate for cleaning type F (20 micron) porous metal filter elements fouled in the absence of filter aid with siliceous sludge from Darex solvent extraction feed solutions. Sections of unirradiated stainless steel-clad UO/sub 2/, sheared into lengths of 1/2 in., 1 in., 2 in., and 2 3/4 in., were leached free of UO/sub 2/ with 10 M HNO/sub 3/ in 30, 45, 60, and 75 min, respectively. The abrasive blade cost for the removal of inert end adapters from selected reactor assemblies ranges from 12 for the Gas-Cooled Reactor to 19 for Consolidated Edison. A zirconium dissolution was carried to completion at 600 ction prod- …
A brief survey of decarboxylation reactions and carboxylation reactions that are known or presumed in biological systems will be presented. While a considerable number of amino acid decarboxylations are known, their mechanisms will not be included in the present discussion but will be reserved for a later paper in the symposium. The remaining decarboxylation reactions may be subdivided into oxidative and nonoxidative decarboxylations. In most cases, these reactions are practically irreversible except when coupled with suitable energy-yielding systems. The carboxylation reactions which are useful in the formation of carbon-carbon bonds in biological systems seem to fall into two or three groups: those which exhibit an apparent ATP requirement, and those which exhibit a reduced pyridine nucleotide requirement, and those which exhibit no apparent ATP requirement. Of the first group at least four cases, and possibly six or seven, are known, and one interpretation of them involves the preliminary formation of 'active' carbon dioxide, generally in the form of a carbonic acid-phosphoric acid anhydride. Those exhibiting no apparent ATP requirement seem to be susceptible to classifications as enol carboxylations in which the energy level of the substrate compound is high, rather than that of the carbon dioxide. There appear to be at least three examples of this latter type known, amongs them being the carboxy-dismutase reaction of ribulose diphosphate with carbon dioxide.
Extraction systems involving chelating agents are widely used in radiochemical separations on a control basis due to the economy in time and equipment. In the case of waste streams, the element of interest is, by definition, at a very low concentration in the presence of gross fission product activities. The resultant large sample size necessary for a reliable analytical results creates a marked radiation hazard to laboratory personnel.
The development of new techniques of initiating nuclear weapons has amplified the requirement for an accurate knowledge of the probability of initiation as a function of the various parameters involved. The results of the codes presently used to calculate these probabilities are self-consistent, but have been shown to be quite sensitive to the value of fission cross section used. Accordingly the critical assemblies group vas requested informally by small weapons division to investigate the feasibility of providing a prompt critical assembly which could give an experimental basis for normalization of the code results. A study of the various assemblies which met the requirements of this experiment was therefore conducted to determine which was best suited. The conclusion of the study is that a bare spherical oralloy assembly similar in operation to that of the Los Alamos Godiva would be the most suitable device with which to make these measurements. This conclusion was based on the fact that the prompt critical behavior of such a bare oralloy sphere is well understood and further, this behavior can be calculated by means of the present weapons codes. Such an assembly could be in operation in less than six months and would require an estimated total expenditure of $70,000 exclusive of the oralloy components.. Safety of operation could be insured by proper design of components, interlocks and operating procedures. The proposed assembly would be housed in the West Vault in Building 110 which can be made available in the near future. Much of the present instrumentation in the vault can be used for the new assembly. Some modification to the vault will be necessitated by the higher shielding requirements of the prompt assembly. Cost of these modifications is included in the basic cost estimate.
The following report is based on work performed under the auspices of the U.S. Atomic Energy Commission. The purpose of this report is to describe an air-core strong focusing synchrotron and its functions.
Abstract: Design nomographs and equations have been prepared for determining the bending stresses and hoop stresses at the junction of a cylinder and sphere when: (1) Internal pressure exists in the sphere and (2) There is a difference in average temperature between the cylinder and the sphere. A correlation of calculated stresses and photoelastically determined stresses for models with internal pressure is presented.
Tests were conducted to determine methods of removing scale deposits from the HRT heat exchangers. A mockup of the heat exchanger header was cleared of a deposit of iron rust by reverse flushing at a flow rate below 75 gpm. A tube bundle consisting of 109 1/4in. O.D. x 0.049 in. wall tubes was plugged with rust. Approximately 80% of these tubes were unplugged by using a 70 psi water pressure differential in combination with vibration from a pneumatic rivet gun. No mechanical method was employed in the tests which could clear the remaining tubes.
Tests were conducted to determine methods of removing scale deposits from the HRT heat exchangers. A mockup of the heat exchanger header was cleared of a deposit of iron rust by reverse flushing at a flow rate below 75 gpm. A tube bundle consisting of 109 1/4 in OD X 0.049 in. wall tubes was plugged with rust. Approximately 80% of these tubes were unplugged by using a 70 psi water pressure differential in combination with vibration from a pneumatic rivet gun. No mechanical method was employed in the tests which could clear the remaining tubes. (auth)
This document details activities of the irradiation processing department during the month of March, 1959. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering Operation; Employee Relations Operation; and Financial Operation.
This report for December 1958, from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance: Financial operations; facilities engineering; research; and employee relations.
This report, from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; and employee relations.
Flux and dose measurements of fast neutrons, thermal neutrons and gamma rays under perturbing conditions using Godiva as a source are described and discussed. The limitations of single monitoring devices are noted. It is pointed out that in order to improve the accuracy of flux and dose measurements in radiation effects studies, specific detectors must be used at the position of the experimental object. (auth)
The Stack Effluent Radioisotope Monitor is an instrument which continuously measures and records the amounts of each of several different radioactive isotopes in the stack effluent of the Redox Facility. An informal report, HW-56151, describing the theory of operation and the circuits was prepared and is available to Hanford personnel from the technical files. The formal report, HW-56151 Rev, containing essentially the same information, is available from the Office of Technical Services, Department of Commerce, Washington 25, D. C.
The C Reactor was proposed for producing tritium. To evaluate the performance of enriched U-Al J elements and natural Li-Al alloy target (N) elements, 60 charges containing both J and N pieces were irradiated under a variety of conditions in C Reactor. No ruptures were sustained; however Tube 3276-C was discharged because of a suspect. Corrosion rates of J elements were not worse than for natural U irradiated under same conditions. Differences between corrosion of J elements prepared by three different methods were not significant.
To evaluate the self-supported fuel element concept, tests are underway to determine the performance of collapsible bridge-rail supported fuel elements in ribless process tubes under present reactor conditions at B Reactor. It appears expedient, however, to extend this evaluation to future operating conditions in order to establish the relative feasibility of conversion to self-supported fuel elements in ribless tubes in present reactors. Utilization of 1.47% U-235 enrichment will provide fuel element powers comparable to those attainable under proposed future conditions. Since I & E fuel elements of this enrichment have previously attained exposures in excess of 2000 MWD/T at specific powers averaging 75 KW/ft in C Reactor, this test will specifically evaluate the feasibility of the self-supported fuel element concept. The purpose of this report is to present the design of a test to fabricate and evaluate self-supported fuel elements under conditions of comparable severity to those expected for future loadings of this geometry.
The Atomic Energy Commission has assigned Hanford Atomic Products Operation the responsibility of reprocessing some of the slightly enriched uranium fuel elements from nuclear power reactors. Some stainless steel clad fuel elements contain sodium or a sodium-potassium alloy as a heat transfer medium between metallic core and outer sheath. The nature of the reaction between water and sodium is well known; however, the reaction between sodium and an oxidizing acid is more energetic and not so well known. The unknown factor of interest is the pressure generated by the reaction between the next transfer medium and the acid used for fuel dissolution prior to solvent extraction. The scouting studies discussed in this report were made to determine hydraulic pressures developed when sodium is exposed to cold concentrated nitric acid beneath the liquid surface.
Property changes to common graphites resulting from reactor irradiation depends strongly on the temperature of exposure. Radiation affects at 30 C and 400 to 500 C are quite well known; however, at intermediate temperatures, results are not well known although a large volume of graphite in most reactors operate in this range. Production test, IP-22A, was reactivated, after a dormant period of many months, to study radiation damage between 30 and 400 C. This report concerns mainly the experimental details and evaluation of the equipment necessary for the controlled temperature irradiation.
Uranium dioxide pellets sealed in Type 316 stainless steel containers with a helium gas were irradiated in helium and in C0/sub 2/ in thermal fluxes or the order of 1 x 10/sup 13/ n/(cm/sup 2/)(sec). Cladding-surface temperatures were reportedly between 1200 and about 1800 F. The hot-cell examination performed by BMI showed that there were no obvious effects of the irradiation on the specimen tested in helium. However, the specimen irradiated in the presence of C0/sub 2/ exhibited severe cladding-CO/sub 2/ reaction and possible central melting of the UO/sub 2/. Although comparisons between pre- and postirradiation data were difficult because of involved fabrication history of the specimens, the tests did further establish the fact that helium is a satisfactory coolant gas for stainless steel cladding material at a temperature of 1200 F. The data obtained from the specimen tested in the presence of C0/sub 2/ indicate that at temperatures in the range of 1600 to 1800 F Type 316 stainless steel is not compatible with C0/sub 2/. (auth)
The development of bakeable high-vacuum flanges for the ORNL PIG Facility is reported. The general design approach for this type flange is to obtain a bakeable vacuum seal by first welding thin metal rings to a set of heavy metal flanges, and then edge-welding the rings together. This design sllows the option of O-ring sealing for nonbaked operation. A number of flange designs are discussed together with fabrication inspection, testing, and installation and maintenance information. (auth)
In connection with the program of the measurement of eta for U/sup 233/, several critical experiments were performed with light water solutions of uranyl nitrate in an essentially bare sphere 27 inches in diameter. Results of two multigroup-diffusion-theory calculations for the above experiments are presented. Assumed cross sections, material concentrations, detailed neutron balances, and a comparison with elementary theory are included. The agreement between the calculated and experimental multiplication constants is excellent for the multigroup calculation but only fair for the elementary calculation. The latter method overestimates the fast leakage so that the computed multiplication constant is less than that found experimentally. (auth)
In connection with the program of the measurement of eta for U233, several critical experiments have been performed by R. Gwin and D. W. Magnuson of ORML with light water solutions of uranyl nitrate (highly enriched in either U233 or U35) in an essentially bare sphere 27 inches in diameter. This report presents the results of two multigroup-diffusion-theory calculations for the above experiments performed by C. B. Mills and associated at Los Alamos. Assumer cross sections, material concentrations detailed neutron balances and a comparison with elementary theory are included. The agreement between the calculated and experimental multiplication constants is excellent for the multigroup calculation but only fair for the elementary calculation. The latter method overestimates the fast leakage so that the computed multiplication constant is less than that found experimentally.
Fused uranium dioxide has desirable characteristics for fabricating high density nuclear fuel elements by swaging. Methods of preparing dense uranium dioxide particles being investigated at HAPO include are arc fusion, and sintering and crushing of low density powder. This paper reports the results of exploratory experiments to prepare fused uranium dioxide by induction heating of low density powder in graphite crucibles.
A study has been made of comparative costs of disposal of radioactive wastes at sea and by burial, taking into account such factors as loading, storage, and transportation by various means. (auth)
This report is a comparative cost study of radioactive waste disposal for the Lawrence Radiation Laboratory (UCRL). In particular, it compares the costs of sea disposal in depths of 1000 fathoms and of 2000 fathoms off the California coast with land burial of the wastes at the Hanford Atomic Products Operation (HAPO), Richland, Washington, at the National Reactor Testing Station (NRTS), Idaho Falls, Idaho, or at the Nevada Test Site (NTS), Nevada. In the comparison, the cost of utilizing a commercial waste-disposal firm is also shown as well as a tentative cost of using the Military Sea Transport Service (MSTS).
Several hollow uranium cores of "C" size I & E diameters were fabricated by cold extrusion (550 to 750 F) at Hunter Douglas Aluminum Corporation. Results show diameter control and reproducibility are excellent. Preferred orientation induced by this process is completely removed by a single standard beta heat treatment.
Summary. At the time this project was initiated, all brazing had been confined to small retorts of ten cubic feet capacity or less. Larger assemblies were scheduled which required retorts of over 100 cubic feet capacity. Hydrogen atmospheres had given the best results, however, there was considerable reluctance to use hydrogen in these large retorts from a safety standpoint. It was thought that thru the use of PMC 2252, an argon - 2 1/2% hydrogen gas atmosphere which in non-explosive, sufficient cleaning action might be attoined without the inherent hazards encountered with hydrogen. An investigation of the argon - 2 1/2% hydrogen gas as a brazing atomosphoer
This report describes the design of the plant that was construrcted to remove insoluble fission and corrosion products from the circulating fuel solution of the Homogeneous Reactor Test (HRT). (auth)
It has been corfirmed that uranium monocarbide can be prepared by reacting uranium metal powder with methane at about 650 C, and that sintered carbides of greater than 90% of theoretical density can be made. The cast density of carbides has been found to vary from 13.56 g/cm/sup 3/ for UC to 11.7 g/cm/sup 3/ for UC/sub 2/. The electrical resistivity of cast carbides has been found to vary from about 40 microhm-cm for UC to about 90 microhm-cm for UC/sub 2/ . About 500 ppm of iron or silicon has been found to decrease the quality of arc- melted castings of uranium monocarbide. Oxygen reacts with molten carbide to produce metal and carbon monoride. Nitrogen seems to react with molten carbide to displace carbon from the monocarbide and produce some dicarbide. Diffusion rates in the carbides have been found to be appreciable at 1600 C (D = 3.6 x 10/sup -8/ cm/sup 2//sec) and very rapid at 1980 C (D = 2.2 x 10/sup -6/ cm/ sup 2/sec). The activation energy for diffusion in this system is very high (92,000 cal/mole). Additional studies are in progress to determine suitable chemical methods of preparing carbides from available raw materials, to develop techniques for making highdensity compacts of carbide powder, to cast large shapes of carbide by melting techniques, to measure the mechanical and physical properties of cast uranium carbide and its alloys, to determine the chemical compatibility of carbides with various metals and reactor coolants, and to obtain information on the mechanism by which carbide-type compounds are damaged by fission fragments during neutron irradiation. (auth)
Memorandum presenting the effects of wing modification and stability augmentation on the computed longitudinal behavior in the pitch-up region and probable pilot opinion of the pitch-up characteristics of two current fighter airplanes. An exploration of the addition of a wing-leading edge is included. Results regarding computed pitch-up behavior and probable pilot opinion are provided.
BS>Some of the fundamental relationships in a nuclear rocket engine are discussed. The equations required to calculate the performance of the rocket are presented. The problems associated with these calculations are also pointed out. (auth)
Relatively long time cycles are required for the standard nitric acid dissolution of enriched uranium using safe batch size techniques for criticality control. Upwards of forty eight hours are required for the total dissolution of solid of I and E enriched slugs in nitric acid. The dissolution capacity may be increased by: (1) using a dissolver critically safe by virtue of its geometry or, (2)increasing the rate of uranium dissolution. Dilute agua regia -- a mixture of hydrochloric and nitric acids -- is known to dissolve uranium much more rapidly than nitric acid. This reports present a flowsheet for the rapid dissolution of uranium slugs in dilute agua regia.
From abstract: This paper presents the results of a study of the error in radio interferometer measurements of near sources attributable to the use of asymptotic formulae, valid for distant targets.
In accordance with the NPR technical criteria which require that provisions be made for useful isotope production in the control rods, both isotope-producing and non-production rods have been designed. Design Analysis has been requested by Reactor Plant Design to specify the number and location of the isotope rods for the initial installation. This choice, however, cannot be made without knowledge of the fuel element characteristics and a prediction of the reactor operating techniques. Some of the factors affecting the use of isotope-producing rods are discussed in this letter, and the following recommendation is made.
The purpose of this test is to obtain data that may be used as a basis for revising certain Process Standards applying to the cooling or existing reactor shields. The data may also be useful in connection with NPR thermal shield design work. The fact that one of the H reactor bottom thermal shield cooling tubes presently has a flow blockage makes possible an immediate opportunity for test data without undue risk other than that already encountered. Procedures, costs, and hazards of the test are given.
This report was written to describe briefly several possible variations of such systems along with probable cost estimates. Previous work has been shown that the application of scintillation detectors is the simplest and most reliable means for such monitoring. By proper detector selection and measuring methods, the system can be of an approximate gamma dose-rate with gamma energy independence above about 100 Kev. The same system using detector change only is directly applicable to beta, gamma, slow neutron and fast neutron detection. Any type of alarming device desired can be incorporated.
A method for the direct determination of uranium in a cyclohexane solution of tri-n-octylphosphine oxide (TOPO) is presented. The adduct, UO/sub 2/ Cl/sub 2/ x 2TOPO, that is formed when uranium(VI) is extracted from hydrochloric acid solutions by tri-n-octylphosphine oxide absorbs light in the ultraviolet region. This absorbance is measured at 230 m mu vs. a TOPO-cyclohexane solution that was contacted with hydrochloric acid of the same concentration as that in the test aliquot. The molar absorbance index is 5500. The method is not selective; of the elements that are extracted by TOPO from hydrochloric acid, iron(III), zirconium, molybdenum, tin and thorium, only thorium can be tolerated. (auth)
The neocuproine colorimetric method was applied to the determination of copper in yttrium metal. From 5 to 80 - g of copper was determined in the presence of as much as 5OO mg of yttrium with a coefficient of variation of 2%. Copper was reduced to copper (I) with hydroxylamine and complexed with neocuproine (2,9-dimethyl-1,10-phenanthroline). The colored complex was extracted into chloroform and its absorbance measured, after dilution with ethanol, at 457 m - . (auth)
Production of Pu from separations plants was only 58% of April commitment because of Purex difficulties. UO{sub 3} production, shipments met schedules. Pu shape production and shipments exceeded forecast by 14%. Purex HS column, repaired Oct 1958, developed another leak and was bypassed April 18, resulting in Pu and U that required reprocessing. A Palm recovery run at Purex with all- reducing flowsheet, resulted in 87% recovery and excellent decontamination of product. The prototype dual-pass silver reactor in Purex C-cell plugged with offgases. Processing of unclarified feed through Purex solvent extraction continued. Redox dissolver batch sizes for E-metal processing were increased from 1.75 to 2.0 tons. Testing of first extraction cycle acidic flowsheet at Redox continued, with Np losses to HAW being below detection limit. Ru in 1AFS stream increased 10-fold F.P. activity but was removed in acid deficient U cycles. A sulfamic acid process is being explored for dissolving Pu metal. Scope design of Redox dissolver and RMA line replacement at Z plant was completed. Shielding effectiveness of medium and high density x-ray lead glass was compared.
Pu production from separation plants was only 65% of the monthly commitment owing to Purex difficulties. UO{sub 3} production and shipments both met schedules. Although unfabricated Pu metal production was reduced, all shipping commitments were met on schedule. Purex equipment responded satisfactorily to decontamination. 860,000 Ci of Ce{sup 144} were recovered from Purex Conc. IWW. The all-Ti L-3 concentrator loop was installed in the Redox Pu Concentrator. The safety of the slag and crucible dissolver in Finished Products Operation was improved by adding cadmium to each batch. Engineering studies of Palmolive facilities are reported. An emergency water supply for the Purex 241-A waste storage tank farm will be installed. A study was made on casks for NPR fuel shipment. (DLC)
Pu production from separation plants during May was 116% commitments. UO{sub 3} production and shipments met schedules. Button output and shape production was 97 and 121% of schedule/forecast. Recuplex (product recovery) operated at record rates. Processing at Purex was carried out with the HS column bypassed. Palm processing resulted in excellent product quality but with low yield. A sample of fission products was prepared for Curtiss-Wright. Piping modifications were made to the Purex Pu ion exchange units. One Redox feed batch was prepared with dichromate oxidation; the U and Pu streams increased (Ru) as anticipated. Containers and casks were designed for fission product recovery. Design of installation for subassembly of Pit 65 weapon components was begun.
The November 1959 monthly report for the Chemical Processing Department of the Hanford Atomic Products Operation includes information regarding research and engineering efforts with respect to the Purex and Redox process technology. Also discussed is the production operation, finished product operation, power and general maintenance, financial operation, engineering and research operations, and employee operation. (MB)
This document details activities of the irradiation processing department during the month of April, 1959. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and engineering operations; production and reactor operations; facilities engineering operation; employee relations operation; and financial operation.
This document details activities of the irradiation processing department during the month of August 1959. A general summary is included at the start of the report, after which the report is divided into the following sections: research and engineering operations; production and reactor operations; facilities engineering operation; employee relations operation; and financial operation.
This document details activities of the irradiation processing department during the month of July 1959. A general summary is included at the start of the report, after which the report is divided into the following sections: research and engineering operations; facilities engineering operation; employee relations operation; and financial operation.
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