This report describes and evaluates the Hanford 105 reactor building structures' ability to withstand an internal pressure increase. The means by which their roof and wall surface would contain a pressure buildup 0.3 psi, and prevent contaminant release which might accompany a nuclear incident are discussed. Prototypes of the B, D, DR, F, and H reactor block wall configuration, the corrugated transite roof of the K reactors, and the corrugated transite walls of the K and C reactors are evaluated. Methods of securing certain building components are described, and a comparison of several candidate sealant coatings presented for consideration. These tests were performed at the request of the Design Operation, as part of CG-791, an existing reactor containment program. This series of tests represents only a part of the overall modifications program. A study of the reactor building containment design criteria is available in a Hanford document, HW-59236, by T. O. Brown.

Introduction: "This eleventh quarterly report of the research and development activities of the Radiological Sciences Department, Hanford Works, as before, includes some items charged to control but included for general interest. Such are identified as "not charged to research". The previous report of this series was numbered: HW-24131 - April 16, 1952."

Summary: "Expressions for the fast and thermal flux in the core of an externally driven, symmetrical infinite slab reactor with a reflector are derived as a function of the reflector thickness. The driver is represented by an inward current of fast neutrons through the outer boundary of the reflector. The flux distribution curve is plotted for several values of reflector thickness. The object is to determine the effect of changing reflector thickness on the flattening of the flux inside of the core of such a reactor."

The Biology Operation, Hanford Laboratories, required plutonium dioxide particles in five different micron and sub-micron size ranges. These were to be supplied as a water suspension in small containers. It was necessary to design equipment, suitable for hood operation, which would give a satisfactory separation of these particles.

This document will be concerned with the design, operation and maintenance of an automatic fuel element stamping machine. In order to take advantage of automatic unloading and length checking at the Acme-Gridley lathes in the 313 Building, automatic stamping of fuel elements is desirable. This, along with the need for a permanent canning line identification, resulted in a request for a machine to automatically stamp HAPO fuel elements on a production basis.

Organic compounds with high boiling points and low corrosion qualities have been considered for use as reactor coolants and moderators. Those being considered, however, have a greater tendency to leak than does water. A program to evaluate mechanical closures to be used for a leak-tight system for organic fluids has been established. Report HW-59446 covered descriptions of the facility used in this study and the primary results of tests covering two liquids, fittings, and instrumentation. This report will cover final results and other pertinent information.

The decontamination of the present Hanford reactors involves a once-through cleaning operation. Considerable interest has also been shown in determining the feasibility of this once-through technique for cleanup of certain portions of the NPR system. This report contains information on a series of tests performed in the 242-B single pass flow facility. Information concerning the following are presented: (1) Efficiency of decontaminating KER Loop 1 and 2 contaminated specimens. (2) Efficiency of decontaminating present reactor pigtails by different methods. (3) Areas that need further study.

Considerable work has been done on decontamination procedures for both corrosion products and fission products. Testing of promising procedures under typical rupture conditions is an important phase of this work. The Irradiated Rupture Prototype (IRP) has been used since August in this evaluation work. This document is one of a series reporting these data.

A part of the development activities of the Process Equipment Development Group, Chemical Development, HLO through the year 1959 has consisted of test evaluation of some of the never products of manufacturers, design and refinement of novel units to do certain specialized jobs and development of improved components for existing plant equipment. Modification of present chemical processing methods and investigation of new basic procedures have made it necessary to utilize different and "exotic" materials, titanium, has given rise to another problem-- that of the selection of adequate and compatible bearings where the titanium is used as moving or rotating part.

An ex-reactor test stimulating the charging of a tubular fuel element in a Zircaloy-2 process tube showed a problem of severe process tube scratching and galling may occur if metal sliding contact fuel element supports are used. Both the fuel supports and the process tubes were autoclaved. Although the conditions of this test were severe in that the sharp edges rather than a flat surface of the supports were in contact with the process tube, it pointed out the potential problem of galling by point on small area contact supports on heavy fuel elements. This difficulty had not been observed with the lighter tubular or cluster fuel elements used in KER irradiation. In addition to the loss of metal and thinning of the process tube after a number of charge-discharge operations, these scratches may lead to localized corrosion or act as points of stress concentration.

This report describes the development of an ultrasonic instrument for resin level measurements in an ion exchange column.

One of the problems encountered in the design of the EGCR has been the possibility of bowing in the outer six rods of the seven rod cluster. The power distribution in an outer rod is asymmetrical with respect to the center of the rod. This causes a variation in the cladding temperature, and might result in unven thermal expansion, or bowing.

The design of NPR shows large quantities of carbon steel pipe and fittings in the front and rear face areas. These are also the area where the most work must be done during an outage, and consequently lower radiation levels are needed here than elsewhere in the piping system. A single pass decontamination of the front and rear piping without going through the entire loop is desirable. Because of the quantities of chemicals and available down time, an effective shorty interval process is needed. This report gives some recent laboratory data on this subject.

High speed data handling and computing systems are coming into increasing use in the automatic control of complex industrial processes. One of the next significant developments in reactor design is believed to be the application of high speed data handling and control systems to nuclear reactors. This document is a study of such a system intended for use with the Plutonium Recycle Test Reactor.

Report discussing information relative to the pyrophoricity of uranium for the purpose of reactor core fire safety. The report includes physical and chemical properties of uranium.

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.

To insure a full evaluation of the effects of in-reactor loops with respect to all phases of the Plutonium Recycle Program, a separate study of each loop has been undertaken. An initial study was carried out which analyzed the effects of in-reactor loops using the design criteria for the gas loop as a basis. As soon as the design criteria for the H2O high pressure loop became available, a more detailed evaluation was completed for that loop. Recent completion of the scope description of the PRTR fuel element rupture test loop now permits an individual evaluation of this loop.

Accurate sampling of particulate in gaseous effluents is required for auditing and control of radioactive wastes released to the environment at installations generating or processing radioactive materials. Over the years a variety of sampling probe configurations, sampling velocities, lengths and sizes of sample delivery lines have been used locally to obtain these samples. Efforts to properly evaluate the validity of samples in most sampling installations have been hampered by a lack of basic knowledge of particle behavior under the various conditions imposed. Important variables about which minimal knowledge is available are those of particle impaction and retention on and re-entrainment from the walls of sampling lines. In some sampler configurations the effect of these parameters on sampling validity may be made negligible, but in lines already in place. or where long sampling lines are required, the degree to which deposition affects the sampling validity must be established. Deposition and rebuild up, then rapid release to the air stream. Research has been initiated in Hanford Laboratories Operation to provide the required data relative to deposition and retention on conduit walls and to eventually arrive at approved sampling system configurations.

Although the corrosion of resistance of Zircaloy--2 clad fuel elements in high temperature water is excellent, some concern has been expressed in regard to the resistance against fretting corrosion. Since this form of corrosion can proceed quite rapidly it could induce fuel element failure through penetration of the cladding. Fretting corrosion if it occurred, would take place either where the fuel element contacted the process tube or where portions of the fuel element contacted each other, e. g. wire-wrapping around a cluster of cylindrical fuel rods. This form of corrosion is defined as corrosion occurring at contact areas and as such, fretting of Zircaloy-2 could take place in many reactor concepts. This report details the results of testing performed to induce fretting corrosion of Zircaloy-2 and the direction of the future tests.

This document reports the circumstances surrounding the failure of sheathed thermocouples during a test being conducted in the KAPL-120 Loop. The report is prepared to provide a record of these events and to acquaint those who use these thermocouples for testing of other applications with the difficulties that are sometimes encountered.

Thermal oxidation of graphite in flowing CO2 is being studied at 650 to 850 C, in a single-pass gas system at atmospheric pressure, by observing weight loss rates. The method is used to provide comparative data for candidate reactor graphites. The effects on oxidation rates of graphite purity, structure, coke type, graphitization temperatures and other manufacturing variables are determined. In addition, the effects of gas flow rates and graphite surface to volume ratios are observed.

Over the last decade, several computational methods have been developed and used to examine reactor flow transients caused by pump outage. The variations in system character which have been analyzed are sufficiently diverse that it appears worthwhile to compile them into a single report.

Operating plant experiences in restoring to useful service failed equipment contaminated through exposures to radioactive processes is of international interest to the nuclear industry. In accordance with a request from the Hanford Operations Office of the United States Atomic Energy Commission this report of the various measures taken in rehabilitating for re-use expensive engineered equipment has been compiled by the General Electric Company. This report is a review and summary of the disposition of contaminated major processing equipment at Hanford Atomic Products Operation during 1958-1959 and a listing of 1) equipment buried, 2) equipment stored, to be buried or decontaminated, and 3) equipment decontaminated and returned to service.