This is the monthly report for the Hanford Laboratories Operation, January 1960. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

A linear induction pump of a special design was used to circulate molten sodium through a mockup of an experimental "overflow" type of sodium-cooled reactor. The distinctive features of this pump are that no seals or moving parts are required: no piping is required to carry sodium to the pump or away from it because the pump is mounted directly on the reactor vessel, with the windings outside of the vessel and the magnetic flux return path inside the vessel. The pump develops 342 gpm at 6.2 psi when pumping sodium at 600 deg F with an efficiency of 4.7%. (auth)

Multigroup calculations were performed to compare BeO-moderated cores with graphite-moderated cores, using various void fractions and core diameters. The core leakages and conversion ratios which were calculated are presented in a series of curves. (auth)

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SNAP II is the designation for a nuclear auxiliary power unit, designed primarily for utilization in the WS117L satellite vehicle. The SNAP II system consists of a reactor heat source, a mercury Rankin engin, and an alternator. Dynamic analysis of the power conversion system was conducted utilizing a comprehensive analog computer simulation. Feasibility of a parasitic load control for numerous system disturbances was demonstrated. (auth)

The characteristics of Rankine space power plants in the zero gravity aspect of the environment of space were lnvestigated. The expected effects of Rankine space power plants are described. Discussions of experimental techniques for studying these phenomena show that this information can be obtained rapidly and economically. Recommendations for a program to supplement SNAP II and slmllar Ranklne space power development efforts in this vital area are made, and consist of: the development and testing of a small system that adequately simulates a complete Ranklne system, first in zero grayity and finally, in the complete orbltal environment; followed by, the development and similar testing of a complete Rankine system using SNAP ll hardware. (auth)

A mechanism for shutting off a power excursion in a homogeneotns reactor by the rapid formation of bubbles was investigated. Equations are derived which give upper and lower bounds for the radius of a bubble, as a function of time, under conditions present in a reactor. Deduction of the bubble nuclei growth rate from observations of void volume and pressure can be made. (auth)

R and D is reported in the following: Reactor and Fuels (PRTR, Pu fabrication pilot plant, KER, NPR, materials); Chemical R and D (Pm recovery, fission products, Purex column, non-production fuels reprocessing, Salt Cycle process); Physics and Instrument R and D (PCTR, NPR, critical experiments, PRTR); and Biology (monitoring, irradiation experiments).

Analytical methods were employed to determine the maximum volume-to- stress ratio for a two-radii-contour diaphragm pump. A proposed failure criterion considers the effect of biaxial stresses on fatigue failure through. the use of the Mises-Hencky criterion for fatigue failure. By use of the proposed criterion, it was determined that an optimunn ratio of the two radii does exist, its value being dependent on the ratio of diphragm thickness to diaphragm deflection. Values for the optimum ratio of the two radii (where the ratio of radii is defined as the radius of the central pcrtion of the diaphragm contour divided by the radius of the outer pontion of the diaphragm) range from 1.94 to 7.33 as the ratio of diaphragm thickness to diaphragm deflection varies from 0.5 to 0.05, respectively. (auth)

This report examines the pertinent features in the hazards analyses of the Hanford Reactors which may be affected by the substitution of zirconium tubes for the present aluminum process tubes. Resized I & E slugs, designed to preserve present pressure drops across the active zones and to minimize corrosion, are used as examples to compare the characteristics of the zirconium tubes reactor with the present.

An advanced sodium-cooled, graphite-moderated nuclear power plant is described which utilizes high-pressure, high-temperature steam to generate electricity at a high thermal efficiency. Steam is generated at 2400 psig, superheated to 1050 deg F and, after partial expansion in the turbine, reheated to 1000 deg F. Net thermal efficiency of the plant is 42.3%. In a plant sized to produce a net electrical output of 256 Mw, the estimated cost is 8232/kw. Estimated cost of power generation is 6.7 mills/kwh. In a similar plant with a net electrical output of 530 Mw, the estimated power generating cost is 5.4 mills/ kwh. Most of the components of the plant are within the capability of current technology. The major exception is the fuel material, uranium carbide. Preliminary results of the development work now in progress indicate that uranium carbide would be an excellent fuel for high-temperature reactors, but temperature and burnup limitation have yet to be firmly established. Additional development work is also required on the steam generators. These are the single-barrier type similar to those which will be used in the Enrico Fernri Fast Breeder Reactor plant but produce steam at higher pressure and temperature. Questions also remain regarding the use of …

The antiproton--proton interaction was studied at three energies, 2.0, 1.25, and 0.98 Bev. Antiprotons produced internally in the Revatron and channeled externally by a system of bending magnets and quadrupoles were selected from background particles by using a gas Cherenkov counter and scintillation counters. At the two lower energies, an electrostatic-magnetic velocity spectrometer was used to reject background particles. A liquidhydrogen target was completely surrounded by scintillation counters so that all charged secondaries from the antiproton--proion interactions could be detected. With the information obtained from these counters, the --p-bar--p total, elastic, inelastic, and charge-exchange cross sections and the angular distribution of the elastic scatterings were obtained at each energy. The total cross section was found to be 80, 89, and 100 mb at 2.0 1.25. and 0.98 Bev. respeclively. The inelastic cross section was about twothirds of ihe total cross section at each energy. It was found that each of the partial cross sections was dropping off slowly with energy. The results were fitted by an optic al-model c alculation. (auth)

Graphite reacts with such gases as CO{sub 2}, O{sub 2}, or water vapor to form gaseous oxides of carbon. In the case of CO{sub 2}-graphite interaction, the reaction rate is not significant until about 550 C. Water oxidizes graphite, very roughly, three times faster than CO{sub 2}. Air will oxidize graphite appreciably at temperatures below 500 C. Graphite removal from Hanford reactors is very important, since graphite is used both as a structural support and a moderator for neutrons. Griggs has shown that small graphite samples oxidized to 10 per cent weight loss had only about one-half their original compression strength. Hence, the longevity of the reactors depends to a great extent on maintaining a low graphite oxidation rate. A means of monitoring the extent of graphite loss, i. e., the burnout rate, is necessary to establish future reactor operational standards. Presently, weighed samples of reactor grade graphite are placed along the length of an empty process channel in each reactor. Thus, a sample is exposed to the reactor`s ambient conditions of power level, moderator temperature, and gas composition. This program was initiated in the vicinity of June, 1953 by Woodley. This report presents data on graphite burnout obtained from …

This is the monthly report for the Hanford Laboratories Operation, February, 1960. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

An IBM-704 computer code for the harmonics method of criticality calculation for two-region spherical reactors is described. In the harmonics method, the criticality condition corresponds to the vanishing of a certain infiniteorder determinant; in practice, this condition is replaced by equating a finite-order approximating determinant to zero. By hand, the calculations can be performed conveniently only for second-order approximating determinants. The approximating determinant with the described code is customarily of the seventh order. Losses of significant figures prevented the use of larger determinants. The machine running time per case is generally about 30 sec. (auth)

Operational hazards associated with the use of a radioisotope-fueled auxiliary power unit for a satellite mission are evaluated. The entire fabrication-to-flight and/or retrieval and disposal sequence is examined and safe handling procedures suggested. The design and operation of the Task 2 thermoelectric generator is discussed. (C.J.G.)

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>Technical progress on the research and development program being performed in connection with the design of the Sioux Falls Power Reactor is reported. Areas covered in detail include fuel element research and development, reactor mechanical studies, nuclear analysis, chemistry, instrumentation and control, plant safety analysis, feasibility studies, and steam plant and reactor auxiliary systems design. (For preceding period see ACNP-5924.) (W.D.M.)

Promethium is purified from inactive and radioactive long-lived fission- produced rare earths by the use of Dowex 50 or Nalcite HCR cation exchangers and organic eluants. Americium and promethium, having hydrated ionic radii of the same size, are contained in the same fraction of the eluate. Promethium is purified from americium by adsorbing both elements on Dowex 1 (thiocyanate form) and eluting promethium from the resin with ammonium thiocyanate solution. Equllibrium studies were made in order to determine distribution coefficients of the long-lived radioactive rare earths. Elution curves based on analyses of solutions removed from anion and cation exchangers verify the relative values of the distribution coefficients. From Dowex 1 resin, rare earths 58 through 63 elute with ammonium thiocyanate in the order of increasing atomic number. Conditions are established for the expansion of the present 50- to 100-curie- level processing to levels of 1000 to 5000 curies. (auth)

Two types of aluminum-clad uranium alloy fuel elements, a square (parallel flat plate) and a circular (concentric cylindrical shell) were investigated to determine their relative suitability for use in the Piqua Reactor. Nuclear, thermal, and mechanical data are given, and considerations leading to selection of the circular element are presented. Design dimensions are listed and reactor thermal design and operating conditions are given for the proposed element. (auth)

The fuel surface temperatures and flow rate effects which might be produced by plugs in SRE fuel channels were investigated. It was found that plugs which locally insulate more than 20 to 30% of the surface of a fuel rod are dangerous. Heat transfer through the mcderator would not significantly affect the sensitivity of the outlet thermocouple to change in mass flow through a fuel channel, provided the reactor power is above 2 Mw and the channel is not completely plugged. If a 5% or greater increase in a fuel channel DELTA t is taken to indicate that a plug exists in the channel, about 80% of all possible plugs will be detected and roughly 95% of all dangerous plugs will be detected by this criterion. (C.J.G.)

S>Several viewing devices were built or purchased and various viewing aids were developed to permit examination of the core vessel of the Homogeneous Reactor Test. The hole in the core vessel was viewed and photographed from both the inside and the outside. A number of simple periscopes were designed and fabricated. A small television camera was purchased, and manipulators were built to insent it in the core or in the blanket. A radiation-resistant 21-ft-long periscope was procured which is 7/8 in. in diameter and has a 180 deg field of view. (auth)

This is the monthly report for the Hanford Laboratories Operation. Metallurgy, reactor fuels, physics and instrumentation, reactor technology, chemistry, separation processes, biology, financial activities, employee relations, laboratories auxiliaries, radiation protection, operation research, inventions, visits, and personnel status are discussed. This report is for March 1960.

Three tests have been run in the ETR in high temperature, high pressure, recirculating water. In one test, previously unirradiated fuel elements were used and in the other two the fuel was irradiated to 2400 MWD/T at HAPO prior to insertion in the ETR. Failure was initiated by shearing off a projection on the surface of one rod of a fuel element, thus opening a 25-mil hole through the cladding. The projection was sheared off by a hydraulically operated chisel controlled from outside the reactor. The first test was operated seven hours after the defect was opened with no failure. Failure is defined as having occurred when sufficient uranium oxide has formed to split open the cladding and release large amounts of fission products into the loop water. The second test was operated for fourteen hours after the defect was opened with again no failure. The third test was operated for only 33 minutes after the defect cap was sheared off before fission product activity in the loop water caused the test to be terminated.