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 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.

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.

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 May 1960.

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)

An analytical approach is proposed which allows the bunnup (by fission) of uranium required to cause failure in a uranium dioxide-stainicss steel dispersion fuel element to be calculated. The analysis is developed by assuming the matrix of the fuel eicment to be made up of a uniform, close-packed array of spherical UO/sub 2/ particles, each surrounded by and associated with a hollow stainless steel sphere. Equations are then written for the amount of fission gas released into the stainless steel cavity in terms of the UO/sub 2/ particle size and density and the burnup. The release mechanism is by recoil only, since diffusion is unimportant for the particle sizes and temperatures (<1000 tained F) of interest. The gas atoms recoiled from the UO/sub 2/ particle are assumed to diffuse from the stuinless steel shell into the caviiy. The pressure thus exerted in-side the stuinless steel sphere is computed by the application of a real gas law. A suitable failure criterion for an internally pressunized, heavy-walled metal sphere appears to be when the sphere becomes entirely plastic. An equation for the pressure at failure and displacements of the sphere is written in terms of the UO/sub 2/ loading and the yield …

A practical, prototype silicon p-n junction fast-neutron dosimeter, sensitive in the same range as human tissue, was developed, together with sn associated read-out circuit to facilitate the accurate measurement of accumulated dose. From both theoretical and experimental considerations, it was demonstrated that the dosimeter is essentially insensitive to the gamma and thermal components of a uranium fission spectrum. It was shown that accumulated damage effects appear to be environmentally stable up to an ambient temperature of 100 C. A rather raarked reversible temperature dependence of the read-out parameters requires either control of the read-out temperature or temperature compensation in the read-out device. A high degree of reproducibility of dosimeter characteristics from one device to another was not achieved. The lack of reproducibility was attributed to uncontrolled variables in the bulk silicon from which the devices are fabricated, and in the production procedure. (auth)

None

Work on the preparation of graphite spheres containing thorium oxide for use as breeder elements in the Pebble Bed Reactor-Steam Power Plant is described. Graphite spheres of varying density were infiltrated with thorium nitrate solutions, followed by denitration to thorium oxide.

Absolute (p,n) cross sections have been measured for Ti47, V51, Cr52, Co59, and Cu63 at energies between 4 and 6.5 Mov. These data plus earlier measurements of the cross section for inelastic proton scattering have been used to estimate total proton absorption cross sections for V51 and Co59. An optical model calculation using parameters giving a good fit to elastic scattering measurements predicts an absorption cross section in good agreement with the measurements for Co59. For V51, some sets of parameters gave good agreement with the measured absorption cross section, but the fit to the elastic scattering data was only fair.

One of the primary reasons for operating the high pressure KER loops is to obtain information concerning water quality control characteristics for recirculating water cooled reactors. The KER-1 loop is predominantly carbon steel and approximates the water quality conditions specified for the New Production Reactor (NPR).

In answer to the question, "How well do we know the composition of the fuel material for the PRTR start-up tests?", the analytical data on the PRTR fuel elements and other fuel elements which were fabricated by similar processes was gathered and analyzed. The results of this analysis are presented.