An enriched I & E fuel element failed during irradiation under Production Test IP-109-AC in December, 1957. Radiometallurgical examination of this slug was requested by the IPD Process Analysis Operation. This report presents the results of the examination work.

The report is divided into: Fissionable materials (2000 program): studies related to production reactors, studies related to separations plants; reactor development (4000 program): Studies related to plutonium recycle program.

In the development of fuel elements for the NPR, one potentially serious fuel element problem -- high temperature uranium swelling -- has not been experimentally investigated. A series of experiments has been proposed in which uranium fuel rod with different amounts of Zircaloy-2 cladding will be irradiated to high exposure at temperatures equivalent to those expected in an NPR. These experiments should show the importance of high temperature uranium swelling as a limiting factor in NPR fuel element behavior. To obtain sample rod temperature of 250 to 300{degree}C on the surface and 500 to 650{degree}C at the center, the rods will be irradiated in aluminum capsules loaded in standard reactor process tubes. The high temperatures will be obtained by restricting the heat flow from the uranium sample to the coolant. The purpose of this test is to determine the validity of the heat transfer calculations used in predicting the temperature drops within the capsule by irradiating one capsule at known flux conditions and measuring the temperature attained by the uranium rod sample. The data obtained from this test will be used in determining the irradiation conditions required for the full scale uranium swelling tests.

The performance of I & E fuel elements under the original tests{sup 1} was such that they are now being charged at C Reactor on a production basis. Data obtained from the original test, however, were not sufficient to indicate long term corrosion effects of the use of these elements, nor to accurately define their operating characteristics. It is the intent of this test, therefore, to provide for long term corrosion monitoring in tubes charged with and without water-mixing pieces; and to obtain the required operating data to refine the operating characteristics of this fuel upon which power limits are based.

The use of organic cooling media for nuclear reactors operating at high power levels predicates the use of a coolant which will not react violently with metallic uranium in the event of a fuel element failure. This report describes the testing, and subsequent examination, of two pieces of irradiated uranium which were immersed in monoisopropylbiphenyl (MIPB) at high temperatures and pressures for periods of time up to twenty-five days. The uranium samples had different irradiation histories and cooling times. Similar experiments had been performed with unirradiated uranium by the Corrosion and Coatings Operation, and it was wished to determine whether irradiated uranium would react with MIPB in a different manner.

The purpose of this work has been to determine the production rate of tritium in fringe Li-Al alloy columns with the degree of precision necessary for economic analyses of such reactor loadings. These results are provided for use in such an analysis. This experiment indicates the production rate of tritium in the outermost fringe tubes to be T = 0.0216 M{sub E} = 0.175 M{sub t} where T = grams of tritium per full length (67 pieces) charge of Li-Al alloy material; M{sub E} = MWD/adjacent ton of E metal; M{sub t} = MWD/adjacent tube of E metal. The above values should apply for fringe loads utilizing greater or smaller quantities of E metal; that is, for isotope production loadings which are over or under-compensated from a reactivity standpoint. In the actual test load it was calculated that one gram of tritium and 13.5 grams of Pu were made for each 21.3 grams of U-235 burned up. During the same time interval the displaced uranium loading would have generated 24.3 grams of Pu and burned up 29.9 grams of U-235. The factor which seems to limit the accuracy with which these data can be interpreted is the ratio of the …

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Following HRT runs 13 and 14, several metallic specimens were removed from the high pressure system and transferred to the Materials Section for examination. Samples of scale accumulation in the high pressure system were also taken after these runs and transferred to the Materials Section. Examination and analyses of these several specimens are still in progress, but some of the results are available and are reported. A possible interpretation of some of these results indicates that a considerable quantity of nickel was contained in the core scale accumulation at the end of run 13, and that part of this nicke1 was dissolved in solution during run 14. The amount of nickel which may have come from this source during run 14 roughly accounts for all of the increase in nickel in solution durirg run 14. A significant amount of uranium was also found in the sca1e accunnulation in the core after run 13. (auth)

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Fabrication techniques for making metal-ceramic fuel elements containing 80 to 90 vol. f UN or UO/sub 2/ in a Type 302B stainless steel matrix were investigated. A hot press-forging procedure was most successful for of theoretical or better. This procedure consisted of sealing the cold-p;ressed core compacts in stainless steel picuture-frame packs, heating to 1900 deg F, and pressing to a total reduction in thickness of 35%. A pressure approximately 50 tsi was used specimens produced by this method were evaluated on the basis of their microstructure, modulas of rupture, electrical conduuctivity, and resistance to thermal shock. Microscopic and macrcscopic examination showed the presence of a continuous metal skeleton even in specimens containing 90 vol. a fuel. The modulus cf rupture at rcom amperature varied from 22,500 psi for a specimen containing; 63 vol. % UO/sub 2/ to 9,200 psi for a specimen containing 87 vol.% UO/sub 2/. Both the electrical conductivity and resistance to thermal shock of UO/sub 2/ were improved by the addition of a small volume of metal. Gaspressure-bonding techniques appear promising for cladding these cores into composite elements. (auth)

A search of the classified and unclassified literature concerning the degradation of a tributyl phosphate (TBP)-kerosene solvent has been conducted. Chemical Abstracts, Nuclesr Science Abstracts, Abstracts of Classified Reports, and the card file of the National Lead Company of Ohio's Library were included in the search. Pertinent information is presented, in abstract form, as it applies to extractant (TBP) degradation, diluent (kerosene) degradation, and solvent clean-up techniques. (auth)

Under the Atoms-For-Peace Plan, the specification that uranium be limited to 20% enrichment in the U/sup 236/ isotope has necessitated development of a highly-concentrated uranium--aluminum alloy as the fuel material in the composite aluminum plates of research reactor fuel elements. Efforts have been directed to determining the suitability of a nominal 48 wt.% U--Al alloy in relation to previously established procedures for manufacturing platetype aluminum fuel elements. Increasing the uranium concentration from 18 wt. % to the 48 wt.% resulted in increased segregation, higher strength, and loss of ductility, creating additional fabrication difficulties. Nonuniform deformation of the alloy during roll bonding into composite plates caused localized thinning of the cladding which may limit the material to specific reactor applications. Substitution of Type 6061 aluminum for Type 1100 alumium as frame and cladding of the fuel plates improves this condition. A fuel element, containing the 48 wt.% U-Al alloy, was irradiated in the active lattice of the MTR to an estimated burnup of 25% of the U/sup 235/ atoms with no observable damage. (auth)

For the centrifugal pump under investigation, the static head at pump suction, in feet absolute, at cavitation inception was correlated for water and for 1500 F NaK on the basis of the differences of the vapor pressures of the two liquids. The difference between the vapor pressure of water and NaK, for the same conditions of pump speed and liquid flow, was added to the water-test cavitation inception value, and this estimate proved to be a good approximation to the experimental value found for cavitation inception with NaK. (auth)

studied. Acid concentration, bath contamination, and temperature were the parameters investigated. Pickling rates were found to be linear with time. Pickling rates increased linearly with the HF concentration, but variations in the HNO/sub 3/ concentration had little or no effect on the rate. Contamination of the bath with metal ions found in Zircaloy-2 at concentrations below saturation also had little or no effect on the pickling rate. Temperature- dependency studies in the range of 40 to 160 deg F indicated an activation energy of 4.95 kcal per mole for the dissolution of Zircaloy-2 in HF -HNO/sub 3/. The surface appearance of the Zircaloy-2 used in the study was found to be dependent upon both the HF and HNO/sub 3/ concentrations. (auth)

Recent literature concerning uranyl salt complex chemistry and ion exchange was reviewed in an effort to develop the present state of understanding of the equilibria and kinetic mechanisms involved. In the light of this, a development program is discussed which hopefully would lead to further enlightenment. Various kinetic mechanisms of sorption and elution are proposed and a comprehensive mathematical development is given for one such sorption mechanism. (auth)

The evaluation of factors involved in the production of sound uranium castings is the objective of a program of study now in progress. A mathematical model have been developed to calculate the time-temperature relationships in a cylinifical uranium casting during it solidification. Finite-difference equations are used, and the solution is obtained by use of a digital computer. Studies of the effects of such variables as the superheat, rate of pour, mold preheat, mold size, and others are possible, and their effect on casting soundness can be determined. Initial computed results demonstrate that this approach is feasible, and experiments are proposed to verify the method of analysis. The need for accurate values of the thermal properties of the casting and mold material is indicated. (auth)

Experiments in the controlled melting of irradiated fuel speciman, particularly of the APPR, STR, and MTR types, have confirmed that prolonged heating in air at temperatures in excess of the melting point results in the release of a large portion of the radioactivity. On the other hand, a moderate amount of heating in air or steam sufficient only to melt a specimen results mainly in the partial volatilization of rare gases, iodine, bromine, cesium, and rubidium. In the presence of air or water vapor, strontium and other fission products are not released. At trace concentration of fission products, slow melting of the APPR plate at 1525 d C in air or steam effected the release of 50% of the rare gases, 33% of the iodine, 9% of ihe cesium, and traces of strontiuun. After 25% burn-up, the cesiuun value increased to about 60%. Aluminum alloy of the MTR type, also at trace concentration, upon melting at 700 d C released up to 2% of the iodine, 10% of the rare gases, and negligible portions of other fission products. Zirconium alloy of the STR type after 15% burn-up, when melted at 1850 d C, released up to 95% of the rare …

In order to achieve the high exhaust gas temperatures, which are desirable if the full potential of nuclear fission as an energy source for rocket propulsion is to be realized, it seems essential that the fissionable material be maintained in a gaseous mixture with the propellant. It is then necessary to separate the fissionable material from the propellant before discharging the latter, since the loss of fissionable material is prohibitive otherwise. This report presents an analytical evaluation of the characteristics of a vortex tabe which achieves the desired separation by means of a centrifugal field. Propellant is fed into the tube tangentially, at the periphery, and diffuses radially inward through a cloud of fissionable gas, picking up the fission heat as it goes. The fissionable gas is held against this radial propellant flow by the centrifugal vortex field generated by the tangentially entering propellant. The analysis involves several assumptions, the most important of which are that the flow is laminar and that it is inviscid. A set of non-linear first order differential equations is obtained which is sufficient to describe the fissionable gas concentration, temperature, and pressure distributions in the tube. These equations have been integrated numerically for a very …

Radioactive solids will be entrained by tbe off-gas from the hydrofluorination step of the Volatility Process. The Volatility Process is being developed as a method for recovering uranium from zirconium matrix fuel elements. It is proposed to scrub these solids from the off-gas in a packed bnd using refluxed liquid Hf. A development program has been planned and a unit designed for the experimental studies. (auth)

This report discusses the design and budget requirements for an enclosed firing facility. The use of a steel sphere installed in a tunnel facility is recommended. The system would incorporate handling equipment, ventilation, refrigeration and ice handling.

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