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This is an addenda to the final draft for Program X. It contains two changes one for the storage basin for discharged fuel elements and the other for a fuel element storage facility immediately adjacent to the work area. Specifications for each facility are given.

This report discusses the proposal to expand the existing 224-U Building from an instantaneous rate of 13.5 tons of uranium per day to 18.5 tons per day. This is to be accomplished by the installation of two additional 8{prime} calcining kettles (or pots) available through the Atomic Energy Commission from Luckey, Ohio. It is the purpose of this report to present calculations and recommendations upon which design work is proceeding. This expansion is to utilize existing facilities of 224-U whenever possible and this report forms a record of the methods used in evaluating equipment and systems. Included are analysis of pot ventilation and pot unloading, stress analysis of the new pots, radiation hazard calculations and a composite schematic.

As a part of the five year budget study requested by HOO-AEC, a study of separations waste storage requirements for this period was made. This study took into consideration the variant estimates of amount of irradiated uranium to be processed, and the goals in the waste reduction research and development program. The conclusions of this study were at variance, to some extent, with prior studies. Interest has been expressed in publication of this study to permit independent assessment of its bases and assumptions.

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This report presents an analysis of the projected economies and processing capacity requirements for a power reactor fuel reprocessing industry based on the recovery of fertile and fissionable materials from presently proposed power reactors within tbe confines of the continental United 8tates for the next five to ten years. An analysis of the present general state of development of a technology required for such an Industry is given. A summary of results of power reactor reprocessing chemical and engineering development at Oak Ridge National Laboratory from July 1955 through December 1956 is given. (auth)

Various metallographic techniques were employed to determine the best method for the preparation of Al-clad, Ni-plated U fuel elements for bond studies. The quality of the final results and the speed of the preparation were the most important factors to be considered. The procedure presented was found to yield the most advantageous results in the minimum amount of time. (W.L.H.)

I. An adequate and conservative calculational method for evaluation of the heat generation distribution in the primary sodium system substructural areas was developed. The method was programed for the IBM 704 and the IBM 709. The results obtained from analysis of the gamma heat generation in the primary sodium pipe tunnels and in the intermediate heat exchanger cells are presented. Calculations are outlined, and gamma attenuation coefficients for concrete, sodium, and steel are given. II. Results obtained from analysis of the gamma heat generation in areas where the primary sodium system piping layout was changed from that of the previous analysis are presented. Major changes in magnitude of the hot spot heat generation due to the changes are pointed out. (auth)

S>The results of all laboratory research and development on the process for separation of barium-140 from MTR fuel elements are presented. The steps include caustic dissolution separation of barium and strontium with fuming nitric acid and removal of strontium by the chromate-acetate method. The results of laboratory and pilot plant corrosion investigations and high radiation level flowsheet tests in the Multicurie Cell are also included. ( auth)

Tests were conducted to determine the limit of error of the temperature measuring system for high thermal-stress tests on moderator cans and to determine a means for the calibration of chromel-alumel thermocouples after installation. Results and recommendations are included. (J.R.D.)

The thermal neutron flux distributions for the Sodium Reactor Experiment and the Hallam Power Reacter radial shields are calculated by three different methods. The method giving the highest fluxes is used to calculate conservative values of the heat generation rates in these shields. (T.F.H.)

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The EGCR was designed to accommodate up to four gascooled experimental loops plus several experimental fuel elements in the open core. Two of the loops will utilize 51/2-in.-O.D. stainless steel tubes passing through the core along an axis which is about 17 in. from the central axis of the core. The other two loops will utilize 91/2-in.-o.d. tubes about 68 in. from the central axis. Inherent safety in the design, facility design, primary loop design, auxiliary systems and equipment design, primary and secondary containment design, instrumentation and controls, and special operations are discussed. (M.C.G.)

The phase diagram of the Th-- Mo alloy system has been determined to be of the eutectic type with a eutectoid reaction associated with the Th alpha-beta transformation. X-ray, thermal, electric resistance, and metallographic methods have established the eutectic point at 1380 plus or minus 10 deg C and 7.0 plus or minus 0.5 wt% Mo. A eutectoid reaction is proposed at 1358 plus or minus 5 deg C and less than 0.1 wt% Mo. No solubility of Th in Mo was detected at 1325 deg C. (auth)

The transference number of the chloride ion was determined in pure fused MgCl/sub 2/, CaCl/sub 2/, SrCl/sub 2/, and BaCl/sub 2/ utili zing radioactive chloride ions in a porous quartz membrane cell. On the basis of a two-step dissociation for these salts, MCl/sub 2/ in equilibrium MCl/sup +/ + Cl/sup -/ in equilibrium Ml/sup +/ + 2Cl/sup -/, the ext ent of the second dissociation was qualitatively predicted from a consideration of the relative mobilities of the ions. (auth)

A vertical centrifugal sump-type pump utilizing commercially available impeller and volute designs was selected to circulate the fuel salt in the Molten Salt Reactor Experiment (MSRE). Tests were conducted in water to determine the adequacy of the pump design, to assist design of the prototype fuel pump, and to investigate the effectiveness of xenon removal with high velocity liquid jets contacting sweep gas in thc pump tank. Hydraulic head characteristics were within +1 to -3 ft of manufacturers data for a given constant speed. Adequate and neccssary provisions were devised to control the liquid and gas bubble behavior in the pump tank. The results of priming and coastdown tests are reported. During the gas removal tests, the fuel, xenon, and helium in the MSRE were simulated with distilled water, carbon dioxide, and air, respectively. The best configuration removed carbon dioxide from water at approximately 99% of the ideal removal rate when the stripping flow was 65 gpm and the sweep gas flow rate was 4 scfm. (auth)

Phase acceptance of the Alternating-Gradient Synchrotron Linear Accelerator is theoretically examined and found to agree well with an experimental determination of phase acceptance. Bunching in the AGS linac is also analyzed, the results indicating a satisfactory understanding of this process also. A double buncher that should be more efficient than the present single buncher is discussed, and space charge effects (as yet unobserved in bunching) are analyzed. (D.C.W.)

Information needed to solve nuclear safety problems is condensed into a reference book for use by persons familiar with the field. Included are a glossary of terms; useful tables; nuclear constants; criticality calculations; basic nuclear safety limits; solution geometries and critical values; metal critical values; criticality values for intermediate, heterogeneous, and interacting systems; miscellaneous and related information; and report number, author, and subject indexes. (C.H.)

This document is a user manual for those who are familiar with problems 500 Node Transient Heat Transfer Code.

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Neutron yields for alpha particle reactions with beryllium, boron, carbon, magnesium, and oxygen are calculated. The dependence on energy of the incoming alpha particles is analyzed for yield. Calculations are varied for different sets of cross section data.