A correlation between the optical densities if the Pb-Sn alloy system and UO2 with respect to Co^60 gamma radiation has been determined. This enables one to fabricate density standards of whatever geometry may be desired for one in the gamma absorptiometer by simply casting a Pb-Sn alloy of the proper composition to correspond to the density required.

The primary purpose of a cleanup system in a recirculating water loop is to maintain the best possible water quality conditions. This is normally accomplished by continuously purifying all or a portion of the coolant. A secondary objective of the cleanup system is to help maintain the system pH at a constant value. A system that will satisfactorily accomplish both of these objectives is at times difficult to obtain. Generally the pH control characteristics are sacrificed in favor of the more important cleanup requirements. A somewhat new approach to the problem pf cleanup system design appears to offer a solution to this problem for high pH systems.

In some engineering applications it has become necessary to operate equipment containing small diameter rods or tubes oriented parallel to flow stream. In the case of several nuclear reactors such as the Plutonium Recycle Test Reactor and the S. S. Savannah Maritime Reactor, bundles of small diameter rods are used as the fuel elements. The situation also has application to heat exchangers. A method for accurately predicting the pressure drop characteristics of various bundle configurations would be helpful in the design and selection of equipment. Some progress has been made toward gaining a greater knowledge of the bundle characteristics under low temperature conditions. This report includes the development of relationship for the effect of temperature on pressure drop.

During normal operation the NPR will function as a closed system and the coolant will be maintained at a high degree of purity. Periodically, however, the system will have to be opened for fuel element discharge and/or decontamination. During these periods the reactor will be operated on single pass cooling. The use of deionized and deoxygenated water as the single pass coolant is not economically feasible due to the large quantities of coolant that will be required. At present it appears that filtered water will be used for this purpose. Since filtered water contains substantial amounts of dissolved solids (approximately 100 parts per million), and since it is saturated with air, this type of operation will have to be investigated with regard to corrosion problems. A particularly serious problem will be present after the system has been decontaminated, since the high temperature oxide films will have been removed and the bars metal surfaces will be exposed to the coolant.

The first series of tests in the KER-REP-1 loop proved that a fission product contaminated loop could be decontaminated to a safe level for contact maintenance. Since a good decontamination process was available, there was much that could be improved about this process. Further testing of this process and several variations of other processes have been scheduled. The evaluation of these processes includes specific decontamination factors, process corrosion, and loop activity reduction factors (loop decontamination factors).

Information was desired on applicability of seismic methods of geophysical exploration in helping to better define and refine hydrological and geological factors affecting ground disposal of liquid radioactive wastes. To date most of information on the area geology has been obtained from surface mapping and well drilling, although it has long been recognized that geophysical techniques may greatly assist in defining the geology of the region. Although direct measurements are generally preferred to indirect measurements have proven extremely valuable in many applications. Careful analysis of field data from geophysical measurements, and correlation with known geologic conditions, should provide significant knowledge of earth formations. Geophysical methods do not directly determine depth, type and composition of underground earth materials, but rely rather on detecting differences in physical properties (elastic wave velocity, magnetic susceptibility, density or electrical resistivity) of such earth materials.

This is a report of what may be an invention in the preparation of uranium dioxide of a quality particularly useful for the fabrications of fuel elements for nutronic reactors.

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.

In considering the problems associated with the measurement of SSNM content of Non-Production Fuels, two distinct problems are apparent. The first is the technical problem of obtaining highly accurate measurements in a complicated chemical and physical system. The second is the administrative problem of choosing a measurement system which provides a basic of financial settlement acceptable to both the seller (reactor operator) and the buyer (Commission).

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.

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.

Series of isothermal transformation studies for improvement in fabrication and reactor performance of uranium fuel elements.

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.

The following document describes data collected from a 14-month period of analyzing crustaceans containing radioactivity. Provided is information based on the change of radioactivity through seasonal changes.

An improved method for the determination of boron in plutonium is reported. Precipitation of plutonium (III) acid oxalate prior to color development with curcumin results in increased precision, greater speed, and lower costs. Results are presented of a statistical study involving all variables.

Three types of stainless steel (309 SCB, Carpenter 20, 304-L) were exposed to both the liquid and vapor phase of each of three types of solutions. The time of exposure totaled ten days during which the specimens were cleaned and weighed periodically. The three synthetic D-12 waste solutions contained respectively chrome-high chloride, chrome-low chloride and high chloride sans chromium. The solutions were renewed at each weighing.

Introduction: "The problem of determining the properties and necessary controls on the TBP Plant waste disposal to eliminate transport line plugging and to permit higher concentration prior to return of this to the tank farms was studied. A report of preliminary findings was made to L. W. Finsh on January 31, 1953. This report compiles information in the aforementioned report and findings since its issuance. Previous studies related to this problem are summarized in the U. R. Technical Manual, HW-19140."

Report describing the results of the cupferron extraction-copper spark method in determining the impurity elements in plutonium metal. This was required the 234-5 Project analytical program.