This report supersedes ORNL CF-57-2-1, which was an extension and revision of ORNL CF-54-6-72. Details concerning a recommended system of flow-plan symbols and drawing are given. The system is designed to identify the function of all major instrument components and to show schematically the operation of the instrument relative to the particular process. The system is used for identification and designation. The system is a modification of the Instrument Society of American Recommended Practice (RP 5.1).

The new and as yet unsolved problems introduced by the production of large quantities of fission products and radioactive isotopes from fission or neutron capture present mankind a most complex technical, economic, and political problem. On one hand, the possibility of using the fission process to produce energy from an unexploited and abundant natural source is emerging from large programs of research and development. We are also beginning to see the promise of use of particulate and electromagnetic radiation for the good of man. On the other hand, we are presented with the problem of controlling the dangerous products of fission for periods of time measured in terms of many hundreds of years, periods longer than the effective tenure of any political state in history. We must not only devise ways of protecting ourselves in the present and for our lifetime but, in addition, we must establish the basic technical, social, and administrative control of vast quantities of artificial radioactivity that must remain effective for at least ten to twenty lifetimes.

The general Wigner-Eisenbud theory is used to develop a method of analysis for the cross sections of fissionable nuclei. The method is employed in giving a reasonable description of the low energy cross sections in U/sup 235/. The single level fit for U/sup 235/ is known to be unreasonable. Many level expressions for the cross sections are derived--the only approximation to the general theory being the neglect of all but a small group of resonances. It is shown that regardless of the number or definition of the fission channels the many-level expressions require few level parameters: the E/sub lambda /, GAMMA / sub lambda n/, GAMMA /sub lambda gamma / and GAMMA /sub lambda F/ of the single level theory for each resonance and a few additional parameters pertinent to the interference between levels. The interference terms are described and shown to be important. The shape and size of the U/sup 235/ cross sections below 2 ev are fitied to within one per cent using (a) only one negative energy resonance of smaller size than in the single level fits (b) no additional levels to fit the shape other than the observed levels at positive energies (c) three interference parameters …

The Eyewash program, written by James H. Alexander and Nancy D. Given, provides solutions of reactor criticality problems in spherical geometry by means of the group diffusion method. It employs thirty lethargy groups (plus one thermal group) in nine regions. The input consists principally of specifying the geometrical scaling factor, boundaries and compositions of the various regions, and temperature level. The output includes the value of vc that would render the system critical, the relative fission density distribution, fissions, absorptions, and leakages in each lethargy group in each region, and, if desired, an edit of the flux at each space point, each lethargy, and an edit of the macroscopic cross sections for each lethargy, each region. OCUSOL-A is a minor modification and extension of Eyewash. It provides for the computation and editing, on the supervisory control typewriter, of the total absorptions in selected nuclides in the various regions. This information is useful in the computation of breeding ratios and the preparation of detailed neutron balances, and in the estimation of flux-averaged cross sections for use in estimating the rate of change of concentration of the various nuclides with burn-up. The program also provides for saving and transferring the final fission …

Design calculations were made for a system consisting of a pump, one hundred feet of pipe, and a heat exchanger to remove 1 Mw of heat from various aqueous thorium oxide slurries. The rheological properties of the slurries were varied over a range of yield stresses from 0 to 1.5 lb/sq ft and of coefficients of rigidity from 1/2 to 2 centipoise. Two different cases were studied: a heat exchanger having fixed axial and radial delta T in which the tube length was allowed to vary and a heat exchanger having fixed tube length in which the axial and radial delta T were allowed to vary. It was shown that the pump power must be increased by a factor of 15 to 30 in order to maintain satisfactory operation of the heat exchanger as the slurry yield stress is increased form 0 to 1.5 lb/sq ft. However the pump power is essentially independent of heat exchanger tube diameter for any given slurry. The rated capacity of a slurry heat exchange is essentially independent of slurry yield stress and coefficient of rigidity, provided that the tube velocity can be suitably increased as the slurry yield stress in increased.

Measurements of flow rate through the fuel high-pressure system sampler indicate that the average flow rates is about 0.29 gpm (145 lb/hr) plus or minus 50%, which affords an adequate purge of from 12 to 36 volumes through the sample line if the full fifteen minutes of purging is allowed before isolating the sample. The fuel sample cooler was fund to have adequate capacity to reduce the temperature of the fuel solution form about 275 to 70 C, using pre-heated cooling water at 70 C. Uncertainties in temperature measurements make it impossible to estimate an observed over-all heat transfer coefficient.

To supply information about the several letdown valves which have been in HRT service, a typical valve plug examined by the Metallurgy Section confirms the belief that chemical attack on the Stellite #6 was occurring. It appears that most of the corrosion was due to rinse solutions, since this phenomenon has not been noted on letdown valves in the HRT Mockup where over 90% of the operating time has been with UO2SO4 solutions.

A 10% phosphoric acid solution contain 0.2% "Rodine 45" inhibitor is recommended to be used for the chemical cleaning of the HRT seam and closed colling water systems. Wet storage is recommended for both of these systems. The steam system is to be stored with steam condensate containing 100 ppm hydrazine, and closed cooling water system is to be stored with steam condensate containing 1000 ppm potassium chromate.

Radiation level measurements of various HRE fuel system components, made in September, 1956, after 27 moths of storage, are compared with the June, 1954, readings before storage. Measurements were made with a standard cutie pie and results tabulated.

An estimate has been obtained of the heat that should be generated in the HRT core in order to hold the system at operating temperature under no-load conditions. This estimate was made by measuring the feed-water rate to the package boiler during an oxygenated water rung. Results are summarized.

For various reasons it is not possible to leave the uranium or other nuclear fuel in a power reactor until all of it has been "burned up" by fission. In the case of liquid fuel (homogeneous) reactors a small part is continuously bled out, purified and returned. In the case of solid fuel reactors, fuel elements are periodically removed, reprocessed and the "unburned" fuel put back into service. In both cases the purification produces wastes which contain radioactive fission products and transuranic elements, and it is with the disposal of these wastes that we are concerned. For technical reasons, we will limit our consideration to the wastes from the processing of solid fuel elements, and from the processing of the very similar solid "blanket" elements in which fissionable fuel is made from non-fissionable isotopes of uranium and thorium by interaction with neutrons in the outer regions of the nuclear reactor.

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.

Specifications are presented for cleanliness during installation of piping and equipment in the High Level Volatility Laboratory, Bldg. 4507.