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15 Mile Road/Edison Corridor Sewer Tunnel Failure Study, Detroit Area, Michigan

Description: Partial abstract: "The study consisted of field and laboratory investigations, construction evaluation, and geotechnical and structural analyses to determine the cause(s) of distress and failure of a 2600-ft section of 12-ft 9-in. diameter concrete-lined sanitary sewer tunnel in the Detroit, Mich., area. [...] The report includes summaries of all pertinent construction records, results of all pertinent past and current field and laboratory tests on construction and geotechnical materials, and detailed geotechnical and structural analyses based on observed conditions and measured parameters."
Date: January 1981
Creator: Albert, Dick
Partner: UNT Libraries Government Documents Department

HEAVY WATER MODERATED POWER REACTORS. Progress Report for June 1960

Description: At the end of June 1960, 36% of the construction and 94% of the firm design of the Heavy Water Components Test Reactor (HWCTR) were complete. Revised calculations of transients in the liquid-D/sub 2/O-cooled loop of the HWCTR showed that the safety of the loop was not impaired by recent changes in the location and design of the loop heat exchanger. Preliminary operation of a full- scale mock-up of the bayonet for the boiling-D/sub 2/O-cooled loop of the HWCTR indicated that flow-induced vibrations probably will not be a serious problem in this loop. Irradiation specimens were prepared of Zircaloy-clad tubes of uranium oxide that had been vibratory-compacted and swaged to 91% of theoretical density. The National Research University irradiation of a Zircaloy-clad uranium metal fuel tube was terminated because of mechanical damage to the assembly during an attempted reinsertion into the reactor loop. Tandemextruded joints of Zircaloy to stainless steel were readied for long-term irradiation tests to determine the effects of exposure on the mechanical properties of the joints. (For preceding period see DP-505.) (auth)
Date: October 1, 1960
Creator: Hood, R.R. & Isakoff, L. comps.
Partner: UNT Libraries Government Documents Department

CHEMICAL ENGINEERING DIVISION SUMMARY REPORT, JANUARY, FEBRUARY, MARCH 1961

Description: 8 7 < 6 ; : : = 8 g developed for recovering fissionable and fertile materials from shortcooled reactor fuels. The second laboratory demonstration of the melt-refining process with highly irradiated EBR- IItype fuel pins was completed. A 392-g charge of U-5% fissium fuel pins irradiated to an estimated burnup of 0.4 total at.% and cooled 28 days was melt refined for three hours at 1400 deg C. Data were not obtained on the behavior of fission products. The effect of N concentration on the nitridation rates of unirradiated U-fissium alloys in Ar-N atmospheres was determined. Experiments on the storage of fuel pins at 350 deg C in Ar atmospheres showed that the presence of 5% N lowered product yields only slightly during subsequent melt-refining operations. Supplementary pouring techniques, such as the use of probes and mashers designed to break crusts over the melts, are moderately effective, but are a less desirable solution to the problem of maintaining high yields than the elimination of contaminants in the Ar atmosphere. A liquid metal process is under development for recovery of the fissionable material contained in melt refining crucible skulls produced in the EBR-II fuel cycle. Information obtained in separate studies of the individual process steps is listed. A systematic study is underway to ascertain the influence of atomic size, metallic valence, and electronic configuration on the coprecipitation of various metallic elements with the Cd-Ce intermetallic phase CeCd/sub 11/. Values for the coprecipitation coefficient lambda , defined by the equation log (tracer in solution/ total tracer) = lambda log (carrier in solution/total carrier), were determined for Na, Li, K, Y, Ba, lambda = 0; La, lambda = 1.49; Th, lambda = 1.08; Pr, lambda = 0.63; Ga, lambda = 0.23; Sm, lambda = 0.17; U, lambda = 0.13; Sr, lambda = ...
Date: October 31, 1961
Partner: UNT Libraries Government Documents Department

Failure probabilities of steam generator tubes. Annual report

Description: BNL's efforts focused on the following specific items; the probabilities of failure for perfect steam generator tubes, the probabilities of failure for steam generator tubes containing long axisymmetrically thinned sections, and the probabilities of failure for steam generator tubes containing finite length (relatively short) axisymmetric wastages. (auth)
Date: August 1, 1975
Creator: Reich, M.
Partner: UNT Libraries Government Documents Department

THE TUBE STRESS-STRAIN PROPERTIES OF BRITTLE MATERIALS TO 5000 F. Sixth Monthly Report

Description: Additional furnace work using the 25-kw induction power supply indicated reliable long-time performance at 4500 deg F. The detailed check out of the prototype, flat, gas bearing was completed. Bearing operatlon was very stable at 0.0005-in. gap. The design of the spherical bearing was begun. The grips were checked with a single graphite specimen and indicated satisfactory performance with specimen failure at the anticipated ultimate stress. The loading system was assembled, aligned, and operational check out completed. Design study of the optical strain analyzer was continued. (auth)
Date: November 10, 1960
Creator: Pears, C.D.
Partner: UNT Libraries Government Documents Department

INVESTIGATION OF CAVITATION DAMAGE OF MECHANICAL PUMP IMPELLERS OPERATING IN LIQUID METAL SPACE POWER LOOPS. Quarterly Progress Report No. 2, October 1, 1963-December 31, 1963

Description: Water testing of the RI-7C3 impeller in the Pt-2 test stand was completed. Still photographs and movies showed that a vortex was present on the leading edge tip of all blades at NPSH values up to 150 ft at all five test flows of 660, 680, 700, 720, and 740 gpm. Sound data showed a possible correlation with the cavitation performance of the impeller. The sound intensity increased until the tip vortices entered the flow channels and then decreased, reaching a minimum just before head loss occurred. As the head fell off, the sound intensity increased to a level as great or greater than the previous maximum. The TP-1 turbopump detail parts required to complete the assembly of the pump in its modified form were completed. The turbopump was assembled with the tested impeller and installed in the PT-4 water pump test stand. The pump was operated at low speed to assure proper seating of the seals and testing started. The PT-6 liquid metal test stand construction drawings were completed. Construction of the test stand was started by disassembling, cleaning, and acid pickling the entire loop piping. Reassembly of the test loop was started. (N.W.R.)
Date: January 15, 1964
Creator: Kulp, R.S. & Altieri, J.V.
Partner: UNT Libraries Government Documents Department

HEAVY WATER MODERATED POWER REACTORS. Progress Report for October 1959

Description: Continued progress is reported on the design and construction of the Heavy Water Components Test Reactor; 78% of the firm design and 17% of the construction were complete at the end of October 1959. Approximateiy 15% of the firm design for the isolated coolant loops of the HWCTR was also complete. The results of further fabrication tests and irradiation tests of fuel tubes of natural uranium metal are reported. One of the metal tubes failed under irradiation, while other irradiations of metal fuels progressed satisfactorily. (auth)
Date: November 1, 1959
Creator: Hood, R.R. & Isakoff, L. comps.
Partner: UNT Libraries Government Documents Department

PHYSICAL METALLURGY OF UNCOMMON METALS

Description: Incremental couples at 10% intervals across the U-Nb binary system were prepared and diffused. Irradiation damage of nickel single crystals bombarded with 3-Mev electrons from a Van de Graaff generator were studied by Kossel line techniques. It was concluded that most defects anneal out below room temperature and all anneal out below 400 deg C. The cold-rolled texture of tantaium is described by (200) and (110) pole figures. This texture may be approximated by the ideal orientations, {112} <011>, {100{ <011>, and {111} <112>. The directionality of Young's modulus, yield strength, and tensile strengih of tantalum is also presented. The effects of thermal gradients on the transformation kinetics and diffusion in U--10 wt.% Mo were investigated. The alloy U(Fe,Mn) was found to be paramagnetic from 480 to 10 deg K. The remanent magnetization of hematite along particular directions in the (111) plane and along the STA111! direction of a rectangular prism was measured during a complete cycle of temperature chsnge between 488 and 77 deg K. The remanent-temperature relationship snd the thermal hysteresis effect were also measured. The concept of space filling was developed for presenting geometrical relationships of different crystal structures. The structure of the pseudo-binary system ReTi/sub 2/--TiSi/sub 2/ was investigated. (M.C.G.)
Date: April 25, 1961
Creator: Ogilvie, R.E. & Norton, J.T.
Partner: UNT Libraries Government Documents Department

HEAVY WATER MODERATED POWER REACTORS. Progress Report for December 1959

Description: At the end of 1959, 25% of the construction and 85% of the firm design of the Heavy Water Components Test Reactor (HWCTR) were complete. Further safeguards analyses of the HWCTR, done with the aid of analog and digital computers, corroberated earlier data which indicated that the reactor is highly self-regulating, and that the safety system should prevent release of radioactivity outside the containment building. Fabrication tests of metallic uranium fuels and preparation of irradiation specimens of swaged uranium oxide fuel tubes continued. A tube of Zircaloy-2-clad U-2 wt.% Zr failed during a low temperature, low pressure irradiation test to modest exposure. Preliminary examinations of tubular metallurgical joints between Zircaloy and stalnless steel were promising. (For preceding period see DP-445.) (J.R.D.)
Date: January 1, 1960
Creator: Hood, R.R. & Isakoff, L. comps.
Partner: UNT Libraries Government Documents Department

ARMY GAS-COOLED REACTOR SYSTEMS PROGRAM SUMMARY REPORT ON MATERIALS FOR THE GCRE-II

Description: Investigatiors were made of various materials for development of metal- canned and semi-homogeneous GCRE-II fuel element concepts. The materials were studied for application to development of fuels, grapanite, silicon-silicon carbide coatings, metal claddings, carburization barrier coatings, and graphite joining. A survey of the literature showad that uranium carbide fuels are superior to other types for the applications described and that refractory metal or metal carbide fuel coatings appear superior to other types for use with the types of graphite investigated. Experimental measurements were made of the thermal conductivity, tensile strength, stress-strain reiationships, and thermal expansion of graphite powdsr bonded with baked carbon at a final firing temperature of 760 deg C. Results showed that these materials were stronger and more isotropic at all test temperatures than a standard structure graphite such as ATJ. The thermal conductivity is somewhat lower and the thermal extansion slightly higher than the corresponding properties of ATJ. A silicon-silicon carbide coating was developed as an osidation-resistant coating for graphite. Preliminary air oxidation tests at 1000 deg C showed that the first samples survived 2000 hr with 10% failure. Subsequent experiments showed that it is reasonable to expect better performance in further tests. Tests for compatibility with graphite were conducted on zirconium, Zircaloy-2, "A" nickel, and K-Monel at 1750 and 1850 deg F for 1000 and 1500 hr. Chemical analyses, metallography, and tensile tests indicated that the K-Monel is the material most compatible with graphite; it possesses good strength and ductility with negligible carburization or carbon diffusion. Zircaloy-2 tubing showed a growth of from 3.4 to 3.8% when thermal cycled 100 times between 850 and 1850 deg F. Tests for compatibility with Hastelloy X were conducted on graphite samples coated with molybdenum, niobium carbide, and zirconium carbide at 1750 deg F and 300 psi for 1000 and ...
Date: December 30, 1960
Creator: Carpenter, R. & Del Grosso, A.
Partner: UNT Libraries Government Documents Department

RADIOCHEMISTRY DURING START-UP AND EARLY OPERATION OF THE NUCLEAR SHIP SAVANNAH. Final Report

Description: It was demonstrated that the radioactivity content of the primary system of the N.S. Savannah reactor plant was small and normal during the period of initial criticality and start-up, and during the sea trials and acceptance tests. The principal radioactive constituents (/sup 56/Mn, /sup 41/Ar, /sup 13/N and / sup 18/F) are either intrinsic to the primary system of the pressurized water reactor or are normally found in the coo1ant in concentrations comparable to those observed in this program. The /sup 56/Mn concentrations observed at the various reactor power levels were slightiy higher, relative to those for the other nuclides, than those observed in similar reactor plants. This slightly increased concentration is attributable to the fact that the coolant of this reactor was generally maintained between pH6 and pH7, whereas the primary coolants of the other plants were maintained at somewhat higher pH values. Data for fission product concentrations in the primary coolant indicate that their only significant source is uranium contamination of the reactor core surfaces. The observed concentrations do not represent any significant hazard or potential difficulty in plant operation. The small value of 5.6 x 10/sup -2/ mu g/cm/sup 2/ for the surface density of uranium indicates that no significant contamination of these surfaces occurred during core fabrication. No significant defect in a fuel element cladding was detected during the period in which these measurements were performed. The efficiency of the demineralizer for removal of anionic and cationic radionuclides from the primary coolant was shown to exceed 90%. Volatile radionuclides were the only radioactive constituents found in the demineralizer effluent. Data obtained for the concentrations of gross radioactivity in the waste tanks were maintained below the maximum permissible concentrations for discharge to the environment. On the basis of these radiochemistry studies, it may be concluded that the ...
Date: July 1, 1962
Creator: Battist, L; Winnowski, W S; Dieterly, D K & Koch, R C
Partner: UNT Libraries Government Documents Department

METALLURIGICAL EVALUATION OF FAILED BORAX-IV REACTOR FUEL ELEMENTS. Final Report-Metallurgy Program 7.6.11

Description: The fuel elements for the Borax-IV boiling water reactor consisted of ThO/sub 2/--6.36 wt% U0/sub 2/ pellets lead-bonded in X8001 aluminum alloy tube plates with silicon-bonded end closures. When the reactor was brought to power on February 19, 1958, after having been operated for approximately one year, over one-third of the fuel elements were found to have developed leaks, as evidenced by significant fission gas release into the process water. Since no fission breaks had been detected during the last previous operation of the reactor on December 5, 1957, it appeared that the lesks had developed during the 21/2-month shutdown. A detailed metallurgical examination was made of two elements containing several failed tubes to determine the causes of failure and the amount of fission gas released from oxide pellets in unfailed tubes. lt was found that the upper portion of each tube, where a void space had been intentionally left above the pellets, had collapsed under the reactor operating pressure. The reverse bending which occurred in the tube walls caused local cracking. The most probable cause for the lesks which developed during shutdown is considered to be corrosion in cracks and crevices of the collapsed tubing. The performance of the silicon-bonded end closures appeared to be quite adequate. Measurements and analyses of gas samples taken from unfailed tubes showed that the pellets released an average of 3.4 percent of the total fission gas yield, although they had operated at relatively low temperatures. The release is attributed to open porosity which existed in the pellets. (auth)
Date: May 1, 1961
Creator: Reinke, C.F.; Neimark, L.A.; Carlander, R. & Kittel, J.H.
Partner: UNT Libraries Government Documents Department

THE MANUFACTURE OF SUPPLEMENTAL DEPLETED FUEL RODS FOR FCF STARTUP

Description: Approximately 2000 supplemental rods were made for use in EBR-II Fuel Cycle Facility startup tests. They were made in the same manner as Core-I fuel rods but using partially depleted pins instead of fuel pins. A duplex'' or double melting operation was used for Core-I production. The alloys were first melted together and cast in and ingot mold. The ingot was then remelted and injection castto produce fuel pins. In order to simplify the operation, a single melt, or simplex'' operation, alloying and injecting casting in one step was tried. This operation was unsatisfactory because of uncontrollable gas evolution from the ingredients of the charge. The interior parts of the furnace became coated with condensed metal to an extent that threatened mechanical and electrical failure of the furnace. A thermocouple head was developed for use in the injection casting furnace. It had increased accuracy and reliability, and was more easily remotely replaced. The improvements were due to unit construction and improved cold-junction contacts. A statistical analysis was made of a sample of 412 rods. The analysis produced (1) and equation for predicting sodium levels through the selection of sodium loads, and (2) evidence that jacket-preassembly classification is necessary under existing specifications for sodium level. (auth)
Date: September 1, 1963
Creator: Carson, N.J. Jr.; Jelinek, H.F. & Shuck, A.B.
Partner: UNT Libraries Government Documents Department

DEVELOPMENT AND MANUFACTURE OF FUEL, BLANKET, AND THERMOCOUPLE RODS FOR THE EXPERIMENTAL BREEDER REACTOR I, CORE IV

Description: A description is given of the development and manufacture of Core IV for the Experimental Breeder Reactor I (EBR-I). A total of 420 rods and 10 thermocouple fuel rods containing plutonium-1.25 wt% aluminum fuel slugs were made. In addition, 120 blanket rods and 5 thermocouple A description is given of the developrnent and manufacture of Core IV for the Experimental Breeder Reactor I (EBR-I). A total of 420 rods and 10 thermocouple fuel rods containing plutonium-1.25 wt% aluminum fuel slugs were made. In addition, 120 blanket rods and 5 thermocouple rods were assembled with depleted uranium slugs. Both the fissile and fertile slugs were NaK bonded in Zircaloy-2 jacket tubes. The fuel alloy was made by induction melting plutonium and aluminum. The centrifugal casting technique developed for EBR-I, Core-II was tried without success with this alloy. The gas pressure injection casting method developed for the EBR-II fuel cycle was used to cast the Core-IV rods. These were parted into rough slugs which were roll-sized and coined to final shape. After coining, the slugs were heat-treated and inspected. A loading technique was developed to protect the outside of the jacket tube and the weld zone from plutonium contamination. A connector fitting, with an integral tube for NaK filling, was welded to the top of the jacket tube. A measured quantity of NaK was injected through the fill opening by a hypodermic needle and syringe. The filling tube was then welded closed by the capacitor-discharge technique. The rods were then heat treated, leak detected, eddy-current bond inspected, x-rayed for NaK level, and gaged. All rods were surveyed for alpha contamination of the external surfaces and were shipped in approved birdcages to the reactor site. (auth)
Date: September 1, 1963
Creator: Burt, W. R., Jr.; Hins, A. G.; Mayfield, R. M. & Shuck, A. B.
Partner: UNT Libraries Government Documents Department