In early July, 1958, it appeared that Columbia River temperature at HAPO would be near 24.5ºC by the end of August. River temperature were averaging 4º to 5º above 1957 figures and were 3º to 4º above the ten year highs. It seemed desirable to examine the problem to determine if any corrective measure could be taken, since it was apparent that production losses were imminent.

Hastelloy F is a prime contender as a material of construction for plant dissolvers in the power fuel reprocessing program. Consequently, the fabrication and installation of dissolver was undertaken to delineate any unknown problems associated with the use of Hastelloy F; and, at the same time, to provide a vessel for development studies on the Niflex or the Sulfex processes. The purpose of this report is to describe the actual basis for design as well as to present the problems encountered during the fabrication of the vessels.

The purpose of this procedure is to present a detailed, chronological presentation of the preliminary decontamination and post decontamination steps necessary to fulfill the requirements of the Production Test Authorization IP-239-N. The procedure attempts to present the required operation in sufficient detail to successfully accomplish the intent of the test. Certain procedures involve operations of a standard nature and have not been elaborated upon to any great extent, as it is expected that the reactor operations and radiation monitoring personnel will implement these instructions according to standard operating procedures.

Nuclear safety in the dissolution of irradiated 0.95 U235 enriched fuel has been investigated. In particular, critical conditions of fuel of this enrichment in a 52-inch diameter dissolver crib were studied. Since a crib this size is not safe by geometry, dissolution procedures as well as maximum safe batch sizes were analyzed. Uranium-water lattices are normally studied in systems in which rods are uniformly dispersed in the moderator. The results of such a study for 1.34-inch diameter solid rods as well as I. and E. fuel having a 1.37-inch O.D. by a 0.48-inch I.D. have already reported.

The spatial distribution of thermal neutrons in a reactor lattices cell is of fundamental importance for many reasons. First of all, this information allows the determination of the relative absorption rates in each component of the cell which in turn enables a determination of the thermal utilization, and the diffusion length of the lattice. In addition, the observed spatial distributions of thermal neutrons in the lattice cell is of major interest in testing various approximations to the solution of the transport equation such as the P1 and P3 solutions.

We have recently completed measurement of resonance escape probability in the BNL reactor and in several slightly-enriched uranium light water moderated lattices. In addition, WAPD has made available to us data from other slightly enriched uranium-water lattices, so that about 18 lattices in all have been done.

This report describes the design and testing of a 1.17-Bev/c separated K{sup -} beam designed in the fall of 1958 in connection with a 15-in. hydrogen bubble chamber experiment.

Existing experimental information about the nucleon-anitnucleon interaction is reviewed, and a description is given of a theoretical model, based on the Yukawa theory, which seems able to explain the experimental results.

A method is described whereby many types of flaws lying close to the surface of a metal can be ultrasonically detected, regardless of the time duration of the interrogating pulses. Lamb waves are established in the metal between a flaw and the surface by an ultrasonic beam which impinges at the proper angle of incidence. A suitably positioned receiver transducer picks up the waves to reveal the flaw. In this method the usually troublesome surface echo is eliminated from the receiver by an acoustic barrier, making it well suited for routine and automatic testing. Results of applying the technique to several testing problems are discussed.

Neutron attenuation in the old pile shields is dependent more and more on the slowing down characteristics of the iron as the hydrogen is baked out of the masonite. For neutrons above 1 or 2 Mev, iron does a good job by inelastic scattering. However, below this energy attenuation can be done only by the gradual moderation by elastic scattering to thermal energies with subsequent capture in the iron. Since iron is heavy and thus a poor moderator, there is a good possibility that many neutrons of intermediate energy will leak out of a burned out shield. Also, iron has a large dip in its cross section at 25 Kev which might allow a large burst of neutrons at this energy to leak out. Measurements using a lucite moderator with gold foil detectors indicate a large leakage of neutrons of intermediate energy, but interpretation of these measurements is difficult. These considerations prompted an attempt to get a rough idea of the energy distribution of the leakage neutrons through pure iron using a simple qualitative theory.

It has been suggested by many workers in the power reactor field that a more efficient utilization of natural uranium may be obtained in thermal reactors if the Pu produced in a fuel cycle were used to enrich a subsequent cycle in which the irradiation-depleted-uranium would be reirradiated. The work described here was done for the purpose of evaluating the reactivity value of the Pu enrichment under the assumption that all plutonium produced is, after chemical separation, fabricated into separate fuel elements and not alloyed with recycled uranium. It has been suggested that the reactivity value of the Pu decreases with exposure to such an extent that highly burned plutonium should be discarded. We, therefore, wish to look at two of the variables affecting the limiting exposure: the reactivity value of the fuel as a function of exposure and temperature and the fraction of potential fissions which would be discarded as a function of exposure. Though the residence time of the Pu fuel before reprocessing, refabrication, and recycling depends upon the relative decrease in specific power, decrease in reactivity due to fission product build-up, the lifetime of the fuel elements before failure, and the cost of reprocessing, only the first …

Average coefficients of thermal expansion of the temperature range 25 C to approximately 425 C are reported for numerous types of graphite. The dependence of thermal expansion on crystal orientation and crystallite size and the effects of oxidation and cold test hole irradiation are discussed. An empirical relationship between thermal expansion and the initial rate of physical expansion under cold test hole irradiation is formulated.

Apparent excessive concentrations of atmospheric contamination which were reported to be associated with an inert-gas-shielded consumable electrode arc wielding operation were studied to evaluate the potential health hazards. A study was made of the concentrations of metal fume and gaseous products of the operation as well as the spectrum of ultraviolet radiation. Recommendations for necessary controls were made.

This report is a record of experimentation conducted intermittently over a period of two years and directed toward preventing deposition of transport corrosion products on fuel elements and other critical components in high-temperature, circulating water nuclear reactor. It includes the postulated mechanism for deposition, a description of experimental equipment, experimental data, results obtained from the experiments, and recommendations for future study.

A spectrometer has been used with a hollow prism electrically heated to 70 C to measure the refractive indices of solutions of uranium hexafluoride in bromine trifluoride and uranium hexafluoride in bromine pentafluoride. The refractive indices of the two binary systems have been expressed as a function of composition.

Report documenting the suitability of Texas for plant locations to produce synthetic liquid fuels, based on raw materials, water sources, and local interest.

This technical report describes a plant-scale procedure for concentrating plutonium and americium in slag and crucible (metal reduction residues) and other recovery solutions by a calcium oxalate carrier precipitation Conclusions from exploratory laboratory tests on the variables affecting the efficiency are included.

A modification of the specimen support for the Norelco diffractometer permits X-ray diffraction studies to be made of specimens prepared for metallographic examination. The unit has been successfully used to identify phases present in specimens of plutonium-uranium alloys.

Strubl's polarographic method for the determination of uranium in the presence of iron can be used for the determination of uranium in plutonium solutions. A hydroxylamine hydrochloride solution (2M) is used as the supporting electrolyte. This reduces the plutonium to the plus three oxidation state without reducing the uranium. The uranium may then be determined polarographically by measuring the height of the uranyl reduction wave (UVI to UV) at -0.35 volts (vs. S.C.E.) applied potential. Solutions with a U/Pu weight ratio as low as 3.5 x 10(-3) were analyzed with an error of less than 2%. This error increased to 6% for solutions of 1 x 10(-3) U/Pu ration and is considerably larger for smaller U/Pu ratios.

A preliminary report discussing a diamond-drill exploration in the southern part of the Uravan District.

This paper concerns work done by the U.S. Geological Survey on behalf of the Division of Raw Materials of the U.S. Atomic Energy Commission.

Discussing the occurrence of uranium in the Kelso Valley area in Kern County, California

A buncher and pre-accelerator for A-12 is described that has 360 degree acceptance and could handle a current of one ampere from a low voltage source.

An R.F. joint tester was designed to determine the relative heating and tendency toward arching at high RF current density (100 amps/linear inch) of a 12 inch diameter circular copper butt joint in vacuum with several types of inserts and at various amounts of axial loading, ranging from a very small contact pressure to one of 1400 pounds per linear inch of joint.