In the Sulfex process, stainless steel cladding, generally 304-L, is removed from fuel elements by dissolution in boiling four to six molar sulfuric acid. When cladding removal is complete, the decladding solution is removed is removed to waste. The exposed fuel cores are rinsed and dissolved in nitric acid.

A prototype instrument has been designed and built which demonstrates an ultrasonic pulse echo technique of measuring metal thickness in the range .010 to .080 inches. Extension of the range of measurement on the thick side appears straightforward, however, transducer frequency response considerations may limit the extension of the method for very thin sections. The instrument is based upon the principle that sections, when driven by pulses of vibratory energy, resonate at frequencies dependent upon the thickness of the section.

At the present time, high-radioactivity-level wastes at Hanford are neutralized and stored as liquid in underground tanks lined with mild steel. This method of storage is relatively inexpensive and is satisfactory on a short-term basis. However, on a long term basis, liquid storage is less desirable than solid storage because of the greater mobility of the liquid. In addition, storage as aa solid would significantly reduce the volume of waste stored. Consequently, various research and development studies have been undertaken in an attempt to develop a practical waste solidification.

A reliable spline disposal system is necessary for full utilization of the Poison Spline Control System. The spline choppers were not reliable enough so a spline coiler was developed and designed to replace the choppers. A Development Test, IP-306-X (HW-63820), was issued authorizing on-pile testing of the spline coiler at X area. This report closes the Development Test and discusses the use of spline coilers.

At present, emphasis is being given the application of nondestructive testing to the inspection of nuclear fuel in conjunction with complete product specifications. The objective is to quantitatively specify fuel quality parameters, and to measure the entire fuel production against these specifications nondestructively. Under this concept the acceptability of a given fuel element becomes a well-defined factor readily determined by appropriate nondestructive tests.

Two types of enclosed hot junction arrangements are possible with sheathed mineral insulated thermocouples. They are the integral junction and he isolated junction. The integral junction is formed by welding, soldering, brazing or forging the two wires into the end of the sheath. The process forms a good thermal and electrical contact between the three materials.

It is the purpose of this report to explain the bands and uses of a digital computer program for the calculation of steady state steam-water pressure drop and critical flow in pipes. The program has been coded for the IBM 709 computer. While so attempt has been made to obtain improved models of two-phase flow, it is believed that the numerical integration technique used in the code will permit more accurate calculation where pressure drops are large compared to the upstream pressure. The technique should lend itself readily to any improved correlations which appear in the future.

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.

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).

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.

Continued operation of F reactor with high water collection rates during the past 12 years has resulted in numerous detrimental effects. In addition to promoting external corrosion tube leaks, water leaks have corroded the Gunbarrel to the biological shield donut assemblies and cast iron thermal shield blocks, thus preventing the majority of tubes in F reactor from unrestrained thermal expansion. Fatigue of the Van Stone flange under the resulting compression loads leads to eventual failure in some cases. In addition, excessive compression loads exerted against the nozzle gasket result in plastic deformation and eventual failures of the gasket.

A disposal method is needed for wastes generated from the decontamination of the NPR primary coolant loop. The limitations imposed by facilities design criteria for the disposal of NPR wastes preclude direct river release of the spent cleaning solutions because of the anticipated quantities of radioactive material in these wastes. The soil at a 100-N Area trench or crib should not be relied on for removing radionucleotides by ion exchange or filtration because of the high salt content of the wastes and the presence of solubilizing reagents. Permanent or long term storage of large volume of decontamination wastes would be expensive. A waste treatment is sought for concentrating the radioactive materials to volume suitable for long term storage and which would permit dispersal of the excess liquid to the environs.

Over the past ten years, there has been considerable interest and expenditure on the recovery of ionium (thorium-230) from certain residues accumulated from the World War II uranium process at St. Louis, Missouri. Most of these efforts were reported in the classified literature. However, an unclassified report was issued recently by the Mallinckrodt Chemical Works on the plant design for ionium recovery. A review of the earlier efforts shows that consideration of ionium recovery was limited to the St. Louis residues. At that time the development of a substantial U. S. uranium ore milling industry had not yet been achieved.

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.

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.

It is being considered that the in-reactor portion of the test section of the PRTR Gas-Cooled Loop Facility consist, in part, of two concentric Inconel-X tubes. The outer tube is the main stress member, with 3.5 O.D. and a wall thickness of 0.065. The inner tube is 3.0 O.D. and has a wall thickness of 0.030. The inner tube provides an annular gas space to serve as thermal insulation.

Designing a dissolver safe for up to 5 per cent U.235 enriched NPF's and having a one a ton per day capacity is a fundamental problem of Hanford NPF Program. Two basic design concepts have so far evolved, both of which employ a recirculating system in which the fuel elements are placed in the critically safe geometry separated from the bulk of the solution in another vessel. In one concept the fuel elements are charged to a set of geometrically safe cylinders or tubes and solution is circulated through them/ In the other concept the fuel elements are placed on a flat tray while solution is sprayed over them. In the latter system, nuclear safety is obtained by avoiding submerged dissolution and the resulting neutron moderation by the solution.

Decontamination of stainless steel and carbon steel used in high temperature recirculation systems is currently being studied to obtain an effective and economical decontamination process for use in these systems. This report presents the preliminary investigation process which has demonstrated very effective decontamination and is low in cost.

In formulating this general diffusion theory expression which represents the neutron balance in a nuclear chain reactor the following assumptions were made : (1) the medium through which the neutrons are diffusing has a low neutron capture cross section, (2) the region in which the flux distribute is being described is two or three mean free paths from strong sources and sinks or from a boundary. Certainly, is going to the lattice cell of a receptor, both of the above conditions are violated; fuel elements have a large absorption cross section and most lattice cells are only two or three mean free paths to is over-all sites

A means for decontaminating the primary system of recirculating type reactor is necessary to insure operation and maintenance. This recirculation system can be contaminated by fuel element rupture products and induced corrosion product activities.

Plutonium Fuels Development Basic studies. Experiments to determine the effect of plutonium dioxide additions on the sinterability of UO2 have continued. PuO2 has been added to ball milled PWR grade UO2 as a physical mixture, and in the form of the mixed crystal oxide.

The transfer or circulation of raw dissolver solutions containing gross particulate matter presents many problems not easily overcome by standard pumping equipment. In April of 1956 the HAPO concept of a modified Archimedes screw pump was developed. Two basic models, externally powered and driven through extended shafts, were constructed and tested. Operation of these preliminary models was so satisfactory that a third unit, integrally formed into drive motor, was built and placed in extended life test. This report describes construction and testing of the third and final model.

A new package type facility for corrosion decontamination studies has been designed and fabricated. This report gives a description, drawings, and comments on its operation. It has been in operation since December, 1958.

Zirflex cladding waste is to be neutralized to pH 10 before transfer to waste storage tanks. This treatment causes the precipitation of zirconium oxide or hydroxide, which may lead to flow difficulties during transfer. The purpose of this investigation was to determine the physical properties and flow characteristics of the neutralized slurry to assist in the selectin of satisfactory transfer equipment and storage conditions.