The majority of two-phase flow problems involving equations of state are solved by use of point-wise utilization steam table values. In this manner, problems involving the use of the various flow equations of continuity, momentum and energy are generally forced into iterative solutions. Considerable effort towards the development of an analytical expression for the state equation seems indicated so as to simplify the analysis of two-phase problems, particularly the transient cases. The question of instability of state and mixture condition is particularly apparent in the analysis of systems undergoing phase transformation as demonstrated by the significant difference between simple theory and experimental critical flow determinations. The assumption of homogeneous, equilibrium mixtures is indicated as a first attack upon the problem.

On July 31, 1959, an autoclave ruptured while being used for thermal cycling studies of Plutonium Recycle Test Reactor (PRTR) plutonium-aluminum fuel elements. Since stand-in materials were being used in this test, no contamination was involved. This accident could lead to inference of greater hazards associated with PRTR fuel designs than had previously been postulated. An ad hoc technical investigation committee was appointed by the Manager Reactor and Fuels Research and Development, to study the autoclave failure. The committee was charged with developing a sound technical explanation of the accident and/or recommending experimental programs to test hypotheses of the physical and chemical processes leading to the rupture of the autoclave.

The utilization of gloves wherein heavy element additives have been included is a significant means of reducing exposure to hands and, consequently, increasing allowable working time in operations which require direct contact with low energy isotopes - such as plutonium. The number of such gloves available has been limited considerably by the feasibility of fabrication and the practicality of application. The following study was made on two basic gloves - the first, a heavy zinc-neoprene coated glove and the second, a lead loaded neoprene glove available in two thicknesses. Included in the study was a standard 30 gauge neoprene glove.

The chromium-nickel-manganese alloys are a group of austenitic stainless steels which were developed during the Korean War to conserve nickel. These alloys are very similar to their corresponding 300 Series grades in mechanical, physical and corrosion properties. A portion of the nickel in the 300 Series grades has been replaced by approximately 2% manganese for each percent of nickel replaced. Two compositions, AISI 201 and AISI 202, are recognized as standard grades. Two other compositions, AISI 204 and AISI 204-L, have been produced in limited quantities to replace AISI 304 and AISI 304-L. Experience with the 200 Series steels indication they are every bit as good as the grades for which they were once alternates. In some shapes, such as rod and sheet, the cost per pound is considerably lower than the corresponding 300 Series grades.

A preliminary evaluation of the autoclave corrosion behavior of a series of U-low C alloys and a series of U-low Zr alloys prepared by Fuels Fabrication Development Operation has been made. The corrosion testing was conducted by Coatings and Corrosion Operation by the experimental methods and procedure outlined in HW-61378.

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.

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).

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.

Basic Studies. It has been reported previously that a reduction of PuO2 to a suboxide does not occur when a powder sample is heated for one hour at 1450 C. To investigate this anomaly, the present hooded facilities were converted from full air flow to an argon atmosphere to prevent oxidation of a possible suboxide. Five grams of PuO2 powder were heated in dry hydrogen to 1500 C for times of one and eight hours. Immediately after discharge, they were mounted and transferred to a helium atmosphere diffractometer hood. The resulting x-ray diffraction pattern consisted only of the single FCC PuO2 phase.

It is well known that neutron damage generally causes increases in the yield and ultimate strength and a decrease in ductility of a metal. There is a continuing program at HAPO to determine the extent of these changes in Zircaloy-2 as functions of integrated neutron exposure, irradiation temperature, and reactor atmosphere. Three investigations from this program will be described and the results summarized. The first investigation deals with both annealed and cold worked Zircaloy-2 irradiated at approximately 50 C. and the other two investigations deal with annealed Zircaloy-2 irradiated at approximately 100 and 280 C respectively. In each investigation tensile testing was performed at room temperature.

On a project of the magnitude and complexity of the Plutonium Recycle Test Reactor is was considered prudent engineering practice to obtain an independant design review by a competent off-site group that had not participated in any way in the development of the design. Accordingly, the Atomic Power Equipment Department of the General Electric Company was requested to make such an engineering review, and an Assistance to Hanford contract authorizing this work was approved by the Atomic Energy Commission in November, 1958.

The Atomic Energy Commission has assigned Hanford Atomic Products Operation the responsibility of reprocessing some of the slightly enriched uranium fuel elements from nuclear power reactors. Some stainless steel clad fuel elements contain sodium or a sodium-potassium alloy as a heat transfer medium between metallic core and outer sheath. The nature of the reaction between water and sodium is well known; however, the reaction between sodium and an oxidizing acid is more energetic and not so well known. The unknown factor of interest is the pressure generated by the reaction between the next transfer medium and the acid used for fuel dissolution prior to solvent extraction. The scouting studies discussed in this report were made to determine hydraulic pressures developed when sodium is exposed to cold concentrated nitric acid beneath the liquid surface.

The NPR design calls for carbon steel to be the major constituent in the reactor coolant piping system. The piping and its associated fittings will, in all likelihood, be exposed to atmospheric weather conditions during the period of reactor construction. This type of exposure causes rusting. From experience gained during the startup of KER Loop 1 it is expected that there will be initially high NPR coolant activity levels. The high activity during the startup of KER Loop 1 was partially caused by the activation of rust that was eroded from pipe walls. Prevention of rusting on the carbon steel prior to its introduction into the coolant system would reduce the initial activity levels.

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.

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 development of a miniature probe was desired for measuring approximate single nuclide beta dose rate in solution and in various animal organs. This probe designed for biological experiments, was to have maximum possible sensitivity to detect low levels of nuclide concentrations. The desired dimensions of the light pipe were to be approximately one-fourth-inch diameter with lengths of three to twelve inches.

Facilities shall be provided in the 221-U Building for storing the metal solution product of the dissolution step in existing tankage from U, T, and B Plants until a reprocessing campaign is scheduled through Redox. This section shall provide a sampling tank for fuel accountability sampling and a pump tank from which the solution will be pumped via a cross-country pipeline to Redox for further processing.

The Plutonium Recycle Test Reactor is heavy water moderated with a heavy or light water reflector contained by a complex aluminum vessel called a Calandria. ( See Figure 1). Construction of this vessel started in August, 1958, at a large West Coast vendor's plant and was completed at Hanford in December, 1959. The fabrication problems associated with a high integrity welded aluminum vessel were generally unrealized prior to this period. This report covers the fabrication of the Calandria and lists recommendations for improving the design and reducing the cost.

The wrist badge provides a dosimeter that is useful in estimating the radiation dose to the hands and forearms. Its new shield system gives good gamma and slow neutron dose discrimination with duPont 552 film packets. The film can be evaluated using the present technique and equipment. Several attempts to develop hand dosimeters have been made. Finger rings using film have been used routinely but have not been entirely satisfactory for all situations. The wrist badge was developed to provide improved gamma and slow neutron dose measurement of the upper extremities under certain appropriate conditions. The wrist badge dosimeter is not a substitute or alternate for finger ring dosimeters but is a necessary dosimeter for some extremity exposure situations.

Pressurized water nuclear reactors may be designed based upon either of two concepts: (1) pressure vessel, wherein the entire core is placed in a large, high strength fuel channels within a low pressure container. The Plutonium Recycle Test Reactor is a pressure tube type reactor. Selection of this basic type of pressurized water reactor depended to an appreciable extent upon the availability of suitable pressure tubing.

In October 1959, a combination testing station developed at HAPO was reported to the Sheath committee. This testing station consisted of electronic instrumentation and mechanical scanning equipment to check coextruded fuel elements (rod and tube) for clad thickness, clad integrity, bond, and core integrity. The clad tests are performed by eddy current methods and the other are ultrasonic.

Ultrasonic cleaning uses high-frequency sound waves to induce cavitation within a cleaning, medium. During cavitation, millions of small bubbles form and collapse, resulting in agitation proportional to the energy put into the solution. The making and breaking of these bubbles produce the scrubbing action associated with ultrasonic cleaning.

Exponential and critical approach type measurements have been made to determine the critical mass, material buckling, and in a few cases, the extrapolation length for the lattices. This report attempts to list all measurements on water-uranium heterogenous lattices made to date at Hanford. All lattices were water moderated hexagonal arrays loaded with uranium of enrichments up to 3.15.

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).