An annual report is given on the operation and costs of waste-disposal facilities at ORNL laboratories and operating buildings in the Bethel Valley area. The operations of the hot-chemical and metal-waste systems, the process-waste system, and the radioactive-gas-disposal system which utilized the 250-ft stack located in the Radioisotope area are discussed. The miscellaneous operations which include the SS (source and special nuclear) material control, SS material recovery, off-shift service for research divisions, water demineralization plant operations, and hydrogen liquefaction are included. However, the disposal of cooling water from LITR, off-gases from the Hot Pilot Plant, and the ORNL Graphite Reactor building are not covered by the report.

The circuits described herein and shown functionally in Fig. 1 are to be added to the HRT control circuit to provide control and protection for the revised HRT cooling water system. The circuitry will provide protection against excess pressure in the demineralized cooling water loop and cooling water activity, will initiate action to insure containment of activity in event of an explosion and will provide emergency cooling water from the tower basin when required.

A method was developed for calculating shield induced gamma radiation escaping through openings in a biological shield. The method was applied to the HRT and the results indicated that the contribution to the dose from induced activity in the HRT shield was around 0.1 r/hr and was insignificant in comparison to to other mechanisms contributing to the escape of gamma rays through shield openings.

Five uninsulated semipermanent freeze coils, series connected, can be frozen when submerged in 70 F water with a Freon-11 flow rate of 1.85 gpm at an inlet temperature of -40 F. The refrigeration unit of the HRT-CP is capable of delivering >3 gpm to a similar semipermanent freeze coil system located in Cell C. Therefore the number of F-11 risers required in Cell C of the HRT-CP can be minimized by series connecting this many semipermanent freeze coils where required.

A general program has been written to plot curves on the Oracle curve plotter. A description of the code and complete instructions for preparation of input tapes and operation of the code are given. The code tape is available from the Mathematics Panel or from the author.

The fact that material transport does occur under conditions of finite temperature difference in a flowing molten metal system was established. The rate mass transfer was thought to be either diffusion limited or solution rate limited. It is believed that the mass transfer of structural materials in Na or NaK systems is solution rate limited. The limiting process has not been qualitatively or quantitatively confirmed for the Inconel-Na or Inconel-NaK system. Increasing the maximum system wall temperature increases the amount of mass transfer, at least above 1300 deg F. The effect of the total temperature difference across the system on the amount of mass transfer was determined.

A series of two-group calculations has been made on the Oracle for the purpose of obtaining critical-mass and flux distribution data for various ORR core configurations. The 3G3R code of Bate, Einstein, and Kinney was used, together with the RSP code developed by Nelson. This made it possible to obtain results for the three-dimensional case. The results, which are presented graphically, are intended to serve as a guide for the design of experiments until such time as actual measurements are available. The calculations were performed for the "clean cold" case, and it should be realized that the presence in the core of experiments and of fission products built up during operation will materially alter the flux patterns found. It is believed that the critical-mass data are accurate to within 10%. Within the fuel region it is believed that the thermal-flux patterns are the also accurate to this degree. Comparison of the results with MTR critical experiments, however, indicates that the thermal flux in the reflector in the vicinity of the fuel-reflector interface may have been underestimated by a factor of as much as 1.3. It should also be recalled that in a two-group calculation the "fast flux" is often a …

The OCUSOL-A program (ORNL-CF-57-6-4) for Univac is a modification of the Eyewash (ORNL-1925) multi-group, multi-region reactor code. The group=energy-lethargy-temperature relationship are given in Table A. The element code numbers are given in Table B. The cross sections now on the sigma-tape are given in tables in the Appendix numbered with the element code number. This technical report explains the bases for choosing the cross sections.

An investigation of the causes of observed explosive reaction of zirconium-coated uranium dioxide on dissolution in nitric acid was conducted. It was concluded that such a reaction is to be expected. Possible but unconfirmed methods of alleviating the problem are suggested.

Experiments in the controlled melting of irradiated fuel specimens, particularly of the APPR, STR, and MTR types, have confirmed that prolonged heating in air at temperatures in excess of the melting point results in the release of a large portion of the radioactivity. On the other hand, a moderate amount of heating in air or steam sufficient only to melt a specimen results mainly in the partial volatilization of rare gases, iodine, bromine, cesium, and rubidium. In the presence of air or water vapor, strontium and other fission products are not released. At trace concentration of fission products, slow melting of the APPR plate at 1525 C in air or steam effected the release of 50% of the rare gases, 33% of the iodine, 9% of the cesium, and traces of strontium. After 25% burn-up, the cesium value increased to about 60%. Aluminum alloy of the MTR type, also at trace concentration, upon melting at 700 C released up to 2% of the iodine, 10% of the rare gases, and negligible portions of other fission products. Zirconium alloy of the STR type after 15% burn-up, when melted at 1850 C, released up to 95% of the rare gases, 90% of …

The continuous dissolution of 304 stainless steel and stainless steel-UO2 alloy in dilute aqua regia was studied with subsequent stripping of the dissolver product to remove chloride ion. The process has the advantage of producing, by means of a simple head end treatment, a solvent extract feed in a conventional nitric acid medium so that existing solvent extraction processes, materials of construction and waste disposal methods can be used. The purposes of this study were to investigate the variables affecting the dissolution process and to obtain dissolver scale-up data, and to investigate the removal of chloride from the dissolver product and the variables affecting the stripping operation. A continuous flooded pot dissolver was used. It has the advantages of stability of operation and ease of control in comparison with column dissolvers and requires a minimum of mechanical processing prior to dissolution. Stripping of the dissolver product to remove chloride ion was studied in a 4-in. diameter Pyrex bubble-cap column containing 12 single bubble cap plates. Continuous dissolution rates and dissolver product stainless steel loading were correlated with aqua regia feed composition, acid feed rate and surface area exposed to reaction. Profiles of chloride concentration down the stripping column were obtained …