Discussion is given on the welding fabrication of tungsten, molybdenum, niobium, and tantalum. Properties which make the four refractory metals important are tabulatcd along with titanium which is given for comparison. Extensive evaluation was conducted using the gas, tungsten arc welding process employing both manual and machine welding. Design data were obtained exclusively from machine welded sheet materials. Flash welding, resistance spot welding and brazing, electron beam welding, and high frequency resistance welding processes were also applied to molybdenum alloys. The oxidation of molybdenum, tantalum, and niobium in flowing air at 2000 deg F is also given. (P.C.H.)

The temperature, geometry, and flow conditions which exist in the EGCR were duplicated in a mock-up designated as the EGCR Burning Rig to establish the combustion conditions in the reactor. The conditions under which the EGCR Burning Rig will ignite were established and an analytical model was developed which predicts these conditions. Because the Burning Rig cannot exactly dupIicate the reactor situation the final prediction of the safety of the EGCR must rest on computer calculations employing the above analytical model. No advantage in retarding combustion was found in using silicon carbide coated fuel sleeves. The negative results of these tests are due both to the particular geometry of the EGCR moderator and sleeves as well as to the fact that all sleeves tested contained imperfections in the coatings. Chlorine was demonstrated to be an effective agent for extinguishing graphite fires. Concentrations in air of about 1% were observed to extinguish graphite fires at temperatures as high as 1000 deg C. (auth)

The intermediate energy neutron flux existing outside the biological shielding of reactors has not been studied to any great extent previous to this time, because of the lack of an instrument capable of detecting neutrons in the intermediate energy range. The instrument used at the MTR utilizes polyethylene spheres of various sizes to give different amounts of moderation and absorption to the impinging neutrons. A procedure for the approximate determination of the relative number of intermediate energy and fast neutrons is given. By graphical methods the following information is obtained: (1) fraction of intermediate neutrons, (2) fraction of fast neutrons, and (3) the approximate average energy of the fast neutrons. Since the instrument described can be used to determine the thermal neutron flux independent of intermediate and fast fluxes, only one instrument is required to determine the neutron flux in all three energy ranges. Dose calculations indicate the intermediate range neutrons give a dose greater than the dose delivered by fast neutrons around the MTR-ETR reactors under normal operating conditions. (auth)

An investigation was conducted to flush the 1BD coolant purification system demineralizer of resin and to measurc the radiation level at pornts on the resin discharge line. It was found that the system demineralizer was satisfactorily flushed of resin. It was also found that the amount removed is measured by the amount required to recharge the demineralizer. (J.R.D.)

The use of oversized piping in an air lift for transferring solutions at rates less than 5 liters per hour was proven feasible with certain limitations. Reliable operation was also obtained with air lifts containing a lateral run inserted between the point of air injection and the final discharge point. Discharge of the air lifts, especially at low liquid flows, was very erratic under the conditions studied. (auth)

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Progress in the fields of nuclear chemistry, isolation and chemical properties of synthetic elements, chemical separation of isotopes, radiation chemistry, organic chemistry, chemistry of aquecus systems, electrochemistry of corrosion, nonaqueous systems at high temperature, and chemical physics for the year ending June 20, 1961, is reported. Separate abstracts were prepared for each topic. (M.C.G.)

The growth of nonprotective, crust-like oxide films was encountered in high-temperature alloy systems that contain molybdenum, vanadium, or tungsten as strengthening additions. The cause of accelerated oxidation in such alloys appears to be associated with the characteristically low melting temperatures of oxides of these refractory elements. The factors that contribute to a breakdown of oxidation protection in these systems are outlined and remedial methods which may be used to avoid catastrophic oxidation are discussed. Commonly encountered service failures that have resulted from catastrophic oxidation are also described. (auth)

Neutron flux measurements were made in the Oak Ridge Research Reactor beam hole HN-l shield plug. at low reactor power (N/sub L/) with three fuel configurations. The purpose of these tests was to determine the most favorable fuel arrangement in the region of the experimental hole in order to permit minimization of exposure time of an in-pile slurry loop experiment using pure thoria. It was found that the perturbed thermal neutron flux decreased by factors of 2, each 1.4 in., at the forward end of the beam hole. Maximum and average fluxes observed for three fuel configurations were: high, 9.7 x l0/sup 13/ , 5.6 x 10/sup 13/; intermediate, 8.0 x 10/sup 13/, 4.7 x l0/sup 13/; and present operating, 7.4 x l0/sup 13/, 3.8 x 10/sup 13/. In the high and intermediate configurations fuel elements were located in the outer row of the lattice adjacent to the beam hole. Cadmium ratios were generally high (22 to 111) implying low available epi-cadmium flux under any of these configurations. (auth)

The HRT was operated experimentally during run 2l at powers up to 5 Mw to explore the limiting conditions of fuel stability and to demonstrate the reliability of the system. The effect of core pressure on fuel stability was investigated over the range from l250 to 1750 psig. Stable operation at 5 Mw (2.6 Mw in the core) was demonstrated at 1250 psig. At 1600 and 1750 psig, fuel instability accompanied by rapid loss of reactivity occurred at powers down to 2.5 Mw. The threshold power for reactivity loss at intermediate pressures was raised by increasing the fuel acid/sulfate ratio from 0.28 to 0.34. In other studies the fuel temperature was varied from 240 to 275 deg C at sev eral different pressures. In some instances the reactor appeared more stable at the lower temperatures. The effects of suspended solids and oxygen concentration were examined briefly without conclusive results. At times during operation at low pressure and high power, an increase in reactivity, indicating deposition of uranium on the core tank, was observed. During an experiment to investigate this effect, a hole was melted in the core near the equator. The reactor was shut down for examination and modifications to …

The raw water export system forms the last ditch water supply system to the ``O`` and ``C`` type reactors; in the event of electrical and steam power failure, the export system is designed to supply enough raw water coolant. After the original export orifice was modified twice, the export system was retested.

A block-loading of striped columns and tubes simulating a blanket loading were analyzed for product (Pu, tritium, E-metal) after irradiation in IP-255-A-9-FP. Results are rationalized to full-pile values; pile conversion ratios and pile gains are given.

Tests in which aluminum-jacketed, Al-Si bonded uranium fuel elements were baked at various temperatures have shown there is a time-temperature relationship for Al-Si layer decomposition. For heat transfer and secondary coolant barrier considerations, the extent of bonding layer deterioration during fuel element irradiation is important. Currently, Al-Si bonded fuel elements show evidence of spire bond separation, and to a lesser degree, can-bond separation following irradiation. Such evidence has aroused concern for the ability of the currently produced Al-Si bonded fuel elements to withstand future reactor operating conditions. Several potential uranium fabrication and canning process improvements are being developed to further advance fuel element stability and performance. Optimization of process conditions based on these improvements may provide the necessary margin of safety for good bond layer integrity, but before a decision can be made to continue improvement of the present process or convert to a new canning process, more information on the stability of the present fuel element bond is needed. This report presents the irradiation phase of a test which was designed to more fully evaluate Al-Si bond integrity under anticipated future reactor conditions.

The oxidation rate of the moderator in D-Reactor has been monitored from samples placed along the length of process tube channel 3478. During the interval from August 8, 1960, to August 10, 1961 were very high, up to 40%/KOD (1000 operating days). From the shape of the front-to-rear burnout profile, the oxidant appears to be oxygen and/or water.

105-KE downcomer pressure data are tabulated.

Irradiation of the initial test in this program involving ten tubes of alternately charged nickel-plated C-64 alloy clad test elements and X-8001 alloy control elements has been successfully completed. The test indicated that the nickel-plate spalling problem has been resolved as no significant spalling or flaking was observed during the post-irradiation examination. The second test in this program will be to verify the performance of nickel-plate with a pilot loading (up to 100 charges) of fuel elements which have been nickel-plated on a production basis. The objectives of this test are to demonstrate with a larger scale test that nickel-plate performs satisfactorily and that reducing the nominal plate thickness from .6 mil to .2 mil will not affect the performance of the nickel-plate fuel element. This test authorizes the irradiation of up to 100 columns of OIIN, chemically nickel-plated, C-64 alloy jacketed fuel elements to 200% of normal goal exposure to extend the evaluation of nickel-plated fuel elements on a pilot scale at DR Reactor. Seventy columns will be plated to a nominal thickness of .6 mil and thirty columns to a nominal .2 mil thickness. Twenty measured columns, ten representing each plate thickness, will be charged to monitor the …

An analytical program was carried out to obtain an accurate single point measurement of neptunium-37 content versus exposure for E-metal. Two large volume dissolver solution samples representing E-metal from the KW, KE, C, and DR reactors were obtained from the Redox Plant for the basis of analysis. The neptunium-237/U ratio was determined by direct analysis, and the exposure was estimated from the measured Pu/U ratio and the uranium-235 burn-out.

To determine the temperature coefficient of reactivity when the station is operating at power.

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.

The purpose of this review is to report the operability status of the reactor control instrumentation. The status of the instruments was determined twice during the first six months of this year, April 1 and June 1. The information contained in this report is not intended to be a complete description of the control instruments, but only as they apply to reactor control. The assigned Pile Physicist at each reactor reported the status of instrumentation at his reactor. Chart I summarizes the operability status of the various instruments. Chart II shows the relative range of reactor power over which these control instruments apply. Appendix II contains a functional description of the instruments and Appendix III lists how each instrument is used during reactor operation.

The objective of this irradiation is to determine the differences in irradiation behavior; typified by dimensional changes, surface roughness, and overall distortion; resulting from elements of similar fabrication history but different beta heat treating schedule. The fuel will be the inner tube only of an NPR fuel assembly. Both elements were heated in chloride salt at 730C; one was rapidly quenched and the other air cooled to obtain a wide variation in grain size and structure and residual stress.

It is necessary to provide accountability tables and equations for use in routine product buildup predictions for the duration of the E-N and blanket loads charged under PT-IP-350-C. Recommended values are given.

The purpose of this report is to present experimental data concerning the heat transfer and fluid flow conditions within a C-overbore geometry process channel for the cases of steady state operation, flow plugging incidents, and inlet piping failure incidents.

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