Effective cutoff energies for point l/v absorbers inside of spherical and cylindrical cadmium filters have been calculated for thermal reactor neutrons. The neutron spectrum was assumed to consist of a Maxwellian plus a 1/E component, and the parameters varied were the thickness of filter, the Maxwellian temperature and the Maxwellian to 1/E flux ratio. Because of the sensitivity of the effective cutoff to Maxwellian flux parameters for thin filters it is recommended that filter thicknesses of about 40 mils be used. Forty mil filters show effective cutoffs at about 0.50 to 0.55 ev for temperatures up to about 500 ction prod- A (or about 0.045 ev). Effective cutoff energies for boron filters were also calculated for purposes of comparison. The cutoffs for cylindrical cadmium filters should be applicable to a properly designed experimental facility. (auth)

A number of organisms have been examined for their ability to produce electron-spin-resonance signals at low temperatures in response to illumination. The efficiency of the response is of the order of not less than 5%, and the wavelength for maximum response is generally slightly on the longer side of the wavelength of maximum absorption, with a minimum appearing at the wavelength of maximum absorption.

The General Electric Company operates eight large filter plants for the Atomic Energy Commission at the Hanford works in the state of Washington. Because of the importance of water to the process, research and development on water treatment has been an important part of the overall Hanford research and development program. The research and development on water treatment has resulted in important capital and operating savings and in the production of better quality water. It is the purpose of this paper to present some of the information developed by the programs. 3 tabs.

The effects of core barrier coatings, void spaces, and surface-cleaning techniques on the quality of Zircaloyclad flat-plate UO/sub 2/ fuel elements prepared by gas-pressure bonding were investigated. Techniques were developed for the application of barrier layers of chromium by a vapordeposition process and of crystalline carbon by a pyrolytic process. These coatings, together with a graphite coating previously developed, were evaluated in pressure-bonded fuel elements for their effectiveness in preventing coreto-cladding reaction and for their general production feasibility. Crystalline carbon coatings 15 to 40 mu in. thick and chromium coatings 25 to 40 mu in. thick were determined to be satisfactory. In the stady of the flow of cladding-plate material into void spaces in the picture-frame assembly, it was established that excessive flow, and consequent thinning of the cladding, can be minimized by individually compartmentalizing the cores with Zircaloy ribs. This design resulted in maximum restriction of the effects of a cladding failure in service. Quantitative data on the maximum amount of void space resulting from manufucturing tolerances or from chipped fuel cores that is tolerable in cladding failure in service. Quantitative data on the maximum amount of void space resulting from manufucturing tolerances or from chipped fuel cores that …

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Materials and specifications for the construction of the N-Reactor buildings are presented.

This monthly document details activities of the Hanford Operations Office during the month of December 1959. (FI)

This monthly document details activities of the Hanford Operations Office during the month of February 1960. (FI)

This monthly document details activities of the Hanford Operations Office during the month of January 1960. (FI)

The failure mechanics of high exposure, in-reactor coextruded fuel rods are quite different from those of defected unirradiated rods. The appearance and corrosion behavior of the high-exposure in-reactor failures suggests that the strength of the coextruded Zircaloy-2 to uranium bond has deteriorated. Notch-fracture tests, in which the strength of the Zircaloy-2 clad to uranium bond is evaluated in a qualitative manner, suggests that the bond strength has not deteriorated to the degree indicated by the failure behavior. It is believed that the irradiation induced property changes of the uranium fuel and not a deterioration of the character of the bond are responsible for the difference in irradiated and unirradiated failure behavior.

A knowledge of the thermal and hydraulic conditions within a reactor fuel channel during an inadvertent flow reduction is needed to establish reactor operating limits. Such limits, which are based on outlet water temperature, are called ``trip-after-instability`` limits by the reactor operating personnel. Laboratory experiments were performed to update the knowledge of such conditions in a BDF reactor type fuel channel while using internally and externally cooled fuel elements (I&E`s) at tube powers up to 1530 KW. In addition to a general extension of previous data, the new data were used to review certain specific details involved in ``trip-after-instability`` limit calculations. It was found that in calculating the limits, the isothermal pressure drop across the fuel elements must be related to flow rate by the exponent 1.8, ({delta}P {proportional_to} F{sup 1.8}), rather than the more convenient value of 2.0. It was found that this method of limit determination is applicable to the high rear header pressures presently attained on the reactors and also applicable to tubes with very low Panellit pressures. And finally, the validity of certain analytical transformations of experimental data, called generalization of hydraulic demand curves, was reaffirmed for the above conditions.

A study was made to determine pressure differentials across coolant system components by considering the blowers' influence during the transient state following a rupture in one hot leg of the Experimental Gas Cooled Reactor Coolant System. Results are presented in tables and graphs. It was concluded that a large instantaneous rupture of the reactor coolant system results in lifting the core when the entire graphite area is exposed to the differential. The forces depend on the type of gas seal used to prevent bypass of reactor coolant around the graphite columns. If the gas seal is located at the periphery of the column, the entire graphite area is subjected to the pressure differential. The use of a separate gas seal at each fuel channel results in only a portion of the graphite column cross-sectional area exposed to the pressure differential thereby reducing the upward force on the core graphite. The incorporation of quick-closing'' valves in the reactor coolant system also reduces the differential pressure across the core and limits the lifting force to which the core is subjected. (M.C.G.)

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To find the critical size of the proposed shield test facility based upon available data and present construction concepts.

The objective of this irradiation is to verify the corrosion rate of a cluster-type fuel element under conditions of high specific power and central core temperatures. The test will also be used in the development of rupture detection instrumentation and decontamination procedures as a necessary part in the development of the NPR. In this test a fuel element irradiated to 3200 MWD/T will be ruptured, the rate of rupture product released will be determined and the gamma spectrum from fission products released into the coolant will be observed. Permission is requested for charging three 7-rod cluster fuel elements (one previously irradiated to 3200 MWD/T at the Hanford projects, the other two unirradiated) into the GEH-R33P7 loop of the ETR. The irradiated element will have attached to it a hydraulic mechanism for opening a defect in one of its fuel rods. The other two elements are to serve as heaters to enable the loop to operate at desired temperatures.

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The test detailed in this report was designed to obtain irradiation data on the brazed closure on a tube-tube geometry. A 1.6% enriched, 2 w/o zirconium core was chosen for this test to gain a faster exposure rate. The brazing alloy for this test was composed of 84 Zry-2 + 4 Be + 12 Fe.

A test was performed to determine the radiation levels of the thermal barriers contained in the rod-drive-mechanism housings. Of the six thermal barriers examined, the two-slot type instrumented barriers exhibited higher radiation levels (350 to 2450 mr/hr) than did either the plain type instrumented barrier (1350 mr/hr) or the three noninstrumented barriers (275 to 725 mr/hr). (W.L.H.)

High tank line cleaning and low flow tests at B, D, F, and H Reactors have been completed and results analyzed and related to high tank backup support. Interim provisions have been completed at the old reactors for bleeding and temperature measurement of high tank cooling water. This letter supplies a recommended guide for maintaining high tank temperature control.

A survey was made to select the construction materials for the SDR-1 reactor core vessel and grid plates. Hastelloy C was selected for the reactor vessel, top grid plate, and bottom grid plate. Inconel X was selected for the core hold-down springs. (C.J.G.)

The solution photochemistry of chlorophyll and chlorophyll analogs is described. Many cases of electron transfer to or from the porphyrin macrocycle have been found, but in no case has any very large degree of energy storage been achieved. Because of the very rapid back-reaction for products with a {Delta}F of approximately -30 kcal, some solid state models in which such an energy storage might be achieved are described and their possible relation to the natural photosynthetic apparatus is given. We can see that while the solid state model (phthalocyanine) allows an approach from a somewhat different point of view, the net result is the same as what was sought, but so far not found, when we looked at the solution chemistry of chlorophyll (and chlorophyll model substances), namely, the transfer of an electron, or hydrogen atom, from the excited porphyrin to an electron acceptor at a high reduction level which can be used to reduce the ultimate carbon dioxide reducers, followed by the donation of an electron ultimately from water to the remaining radical ion, or lattice, which produces the net results of the transfer of the hydrogen from water to carbon dioxide.

Using a simplified model, calculations were made of the possible effects of voids upon the power density in ORR loop experiments. It is concluded that the power density may be markedly increased if voids and channels are plugged with moderator material such as graphite or beryllium. (auth)

Initial heat-treatment studies are reported in which the effects of quenching rate upon room temperature ductility of iron-aluminum alloys were investigated. Quenching media included bciling water, oil at room temperature, and oil at 100 deg C. Experimental results are tabulated. No definite trends in tensile properties were observed with differences in quenching rate. Electropolishing techniques for use in preparation of specimens are discussed in relation to determination of microcrack effects on tensile tests. (For preceding period see AECU-4528.) (J.R.D.)

The fuel value of U/sup 233/ was calculated for five thermal reactors (Dresden. Yankee, Carolinas-Vlrginia, Hallam, GCR-II). Relative to a U/sup 235/ value of per gram, pure U/sup 233/ had a value that varied from .2 to .2 per gram. U/sup 233/ contained in once- and twice-recycle uranium from an initial U/sup 233/-Th cycle had a value slightly in excess of the value of pure U/ sup 233.. The value of U/sup 233/ in recycle uranium from an initial U/sup 235/- Th cycle was less than that for pure U/sup 233/ and decreased with each succeeding cycle. (auth)