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Tantalum-sheathed plutonium fuel pins were prepared for the first core loading of the Los Alamos Molten Plutonium Reactor Experiment-I. Plutonium--10 at.% iron alloy was prepared by co-reduction and by co-melting methods. After casting the alloy into rods, each rod was machined and finished to a 0.357-in.- diameter piece weighing 175 g. The finished alloy rod was finally placed in a tantalum sheath, then sealed by fusion welding to a tantalum cap. Procedures and equipment used for alloying, casting, machining, welding, and inspection are described. Methods used to prepare rods of other low-melting plutonium alloys also are discussed. (auth)

Reactivity and fuel burnup studies were performed for a 255 Mw(e) sodium- graphite reactor of the advanced calandria core type. This reactor is briefly described. Initial criticality calculations and flux distributions were obtained, using two-group theory for enrichments between 2.0 at.% U/sup 325/ and 4.0 at.% U235. A four-group burnup study was performed for enrichments between 2.5 at.% Uisup nd 3.25 at.% U/sup 235/. Core lifetime, changes in isotopic fuel composition, variations in radial power distribution, and fuel cross sec tions are presented. Reactivity during core lifetime was assumed to be controlled by the presence of a homogeneous poison which simulated the effects of control rcds. The results presentad are useful in determining initial enrichment selection in fuel programming and fuel cost studies. (auth)

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)

Samples capsules containing buttons of materials of interest for SNAP fuel elements were held at temperatures up to 1600°F , in hydrogen, under a 20-tsi load for 200+ hours.

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)

This document presents a summary of work and progress at the Hanford Engineer Works for July, 1969. The report is divided into sections by department. A plant wide general summary is included at the beginning of the report, after which the departmental summaries begin. The manufacturing Department reports plant statistics, and summaries for the Metal Preparation, Reactor and separation sections. The Engineering Department`s section summarizes work for the Technical, Design, and Project sections. Costs for the various departments are presented in the Financial Department`s summary. The Medical, Radiological Sciences, Utilities and General Services, Employee and Public Relations, and Community Real Estate and Services Departments have sections presenting their monthly statistics, work, progress, and summaries.

BS> A method for the measurement of thermal conductivities of electrically conducting materials at temperatures in excess of 1000 deg C with a high accuracy was devised. The spatial boron concentration gradient in boron- doped graphite to achieve maximum p-type thermoelectric output was established. The production of an n-type silicon carbide-graphite composit was studied. Theoretical studies produced a detailed picture of the effect of doping on the electronic properties of graphite. A method was developed for the preparation of rare earth nitrides in a good state of purity. Electrical studies of the materials produced showed that several of the compounds are semiconductors as had been predicted. (M.C.G.)

The purpose of this report is to define the extent of modifications necessary to increase capacities of the 100-B, C, D, DR, F, and H water plants for reactor flows of 90,000 95,000 105,000 and 115,000 GPM, and to provide supporting data for budget studies for increased production.

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Three small samples of zircalloy-2, obtained from a KER tube, were irradiated in the Quickie Facility in F Reactor for one, four, and nine months, respectively. Each sample weighed approximately two grams. One day after discharge from the reactor, the specific activity of each of the samples was essentially the same (about 150 mr/hr/gram at one foot in air). Attenuation measurements taken one day after discharge using lead plates yielded a ten-fold reduction in gamma intensity for each 0.95 inches of lead. This indicates that an enclosure for kilogram samples of zircalloy which have been irradiated for long periods of time at fluxes up to 10{sup 14} neutrons per square centimeter per second should have six inch thick lead walls or the equivalent for gamma shielding. Based on the effective gamma energy and the half-life for the first day, it appears that Zr-89 may be the predominant isotope initially. After a few days, the half-life become long and gamma spectrometer measurements indicate the gamma softens revealing that Zr-95 might become the predominant isotope. However, isotopes were not determined specifically; impurities, other isotopes, or combination of isotopes could yield the same results.

The cap-spire pulsing technique of preheating the cap-spire portion of the fuel assembly does significantly improve the brazing of the cap-spire assembly. The air pocket at the spire wafer junction is fully removed. The cap side wafer is essentially 100% wetted with brazing alloy. Destructive tests show that a 4 to 1 improvement in most quality measurements is achieved over present cap preheating techniques, without using additional cleaning step of spire etching. The pulsing is accomplished by a cammed drive system, using a stroke of three-fourths of an inch with a spring return. The system is driven by an electrical gear reduction motor at a rate of 1.4 pulses per second. A preheating cycle of 21 {plus_minus} 2 seconds is used for the current I&E cap designs. The cap-spire assembly does not require any special treatment other than the normal chemical cleaning.

The accuracies which can be achieved in analog recording systems are for the most part tape limited. While reasonable variations in tape speed do not directly affect the accuracy, such variaticns result in improper tape tensioning, which in turn results in poor head-to-tape contact. The quality of the tape transport mechanism should be such that the error introduced by improper head- totape contact is less than tape errors. Some additional equipment is necessary in setting up a tape recorder for a given experiment. The amplitude of the input pulse spectrura to the recorder should be adjusted for optimum value and, if possible, the counting rate should be adjusted to obtain good tape utilization. An oscilloscope is probably adequate to satisfy both conditions. (auth)

This report details activities of the Fuels Development Operation during the months of July, August, and September 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.

As set forth in the CPD Waste Management Program (HW-63958), new and improved methods of waste processing are planned to be initiated at Hanford. A major part of this program is concerned with the high-level wastes from the Purex Plant, with major objectives being the control of long-term hazards, the insurance of production continuity, and the attainment of positive waste accountability at reasonable over-all costs. Specific plans for the disposal of Purex wastes are based on the use of high temperature calcination for immobilization and volume reduction of the solvent extraction raffinates (1WW). Four major steps are involved, each of which is of basic importance to the program objectives. These are: (1) interim 1WW storage and removal for calcination, (2) calcination, (3) primary containment of the calcined waste, and (4) long-term storage. Immobilization of coating wastes and organic wash wastes is also planned, either by tank solidification methods or possibly by inclusion in the calcination feed stream. This document was dated September 15, 1960.

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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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)

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)

The objective of this test is to correct the fuel depression caused by 3X balls lodged in graphite by charging compensating enrichment in the depression region.

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 program`s objective is to develop the technical foundation for use of projection fuel elements through an accelerated testing program. Two types of projection fuel elements are being considered (1) Self-supported fuel elements for use in ribless aluminum or zirconium process tubes; and (2) Bumper fuel elements for use in ribbed process tubes. A further objective of the program is to determine the most desirable method of fabrication and type of support for projection fuel elements. The program must develop the self-supported concept in time to meet the proposed ribless zirconium process tube installation schedule. Parallel development of the bumper fuel element is an interim program to realize immediate gains in reduced hot spot rupture tendency, during the change over to zirconium tubes. The appended charts indicate the proposed accelerated testing and development schedules being followed. Testing of the self-support concept has indicated that the collapsible supports are adequate to maintain fuel centering, do not introduce additional corrosion problems resulting from the supports, and do in fact greatly reduce the incidence of hot-spot flow patterns. Reductions in rupture potential of at least a factor of 30 have been demonstrated at the 95% confidence level. Bumper fuel testing has indicated that …

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)