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In the reference letter and later letters the author made recommendations pertaining to the programs of AEC National Laboratories and other AEC contractors in direct support of the Savannah River project during FY-1956. He believes that these recommendations should now be revised in view of schedule changes ad in the light of accumulated experience. The estimated cost of the programs recommended earlier totaled $2,083,000 for FY-1956. The estimated cost of programs currently recommended total $2,048,000 for FY-1956. The programs mentioned in succeeding paragraphs are concerned mainly with the development of processes for producing extended surface fuel elements and, to a lesser extent, with the development of separations processes. The author considers these particular supporting programs important to the rapid and successful achievement of development objectives at the Savannah River Plant and accordingly recommend that the Commission sustain them. Two of the programs, specifically the work by Sylvania and by Horizons to develop processes for reducing thorium oxide to metal, are not now associated with meeting future production objectives at the Savannah River Plant. The author believes that these programs should be brought to an orderly conclusion, as previously planned, to avoid any loss of the art developed to date.

This production test was designed to evaluate the suitability of low hydrogen dingot uranium as routine process material. Nine tubes of I and E fuel elements (6 dingot, 3 ingot) with 32 fuel elements in each tube, have recently been discharged at the C Reactor and this document contains the results of analyses made on the dimensional stability properties of this material.

Nuclear safety rules for transporting plutonium compounds between the 231 and 234-5 Buildings were established in Documents HW-30328 and HW-31465. Provisions for the co-transporting of loaded sample cans, filter boats and/or product samples were not included in the above documents. In the interest of economy and efficiency the rules have been reviewed and revised according to the provisions set forth in this document. The paper describes the present transportation limits and the revised limits for transporting plutonium nitrate, plutonium oxalate, and other plutonium samples. The limits are in keeping with the policy of preventing the accidental accumulation of unsafe masses of Pu.

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This report discusses the events from a chemical standpoint following a total loss of coolant disaster will not be altered in the melting reactor by the introduction of N metal. The interdiffusion of uranium and aluminum will be the dominating reaction, causing the blockage and tying up of the lithium in UAl{sub 3} which does not melt until after the uranium does. Pressure from the swelling UAl{sub 3} will extrude uranium-aluminum and lithium into graphite weep holes and block interfaces. The migration of lithium by vaporization will not became appreciable until well over 2000{degrees}C, well beyond the time when uranium and UAl{sub 3} have melted. The eventual result will be a diffuse distribution of uranium, lithium, and aluminium in the lattice. The E-N pile has a larger excess over required control capacity than the uranium provided the large reactivity poison tied up in the lithium is not lost. Compared to the natural uranium pile, the gain of reactivity on loss of coolant is less and the net temperature coefficient in the dry pile remains negative to higher exposures. Furthermore, permanent pile poisoning during meltdown is accomplished via two mechanisms both lithium and uranium redistribution in the lattice produce large negative …

A stable carbon arc operated in controlled atmosphere is described. The arc was designed to serve as a light source during lifetime studies of the B/sup 2/ SIGMA state of the CN molecule. The CN radiation from the plasma of the arc was investigated and found to have a brightness temperature of 5500 icient laborato K at lambda 3883 A. This is considerably higher than an estimate of the value required for lifetime measurements. The stability of the carbon arc under various conditions is discussed. For successful lifetime measurements, the light source employed must have a high brightness temperature (intensity). A method for the determination of the brightness temperature of a light source at a specific wave length is described. The method was used for determining the brightness temperatures of some available light sources. Sodium, thallium, and mercury discharge lamps, a medium-pressure mercury arc lamp, and the carbon arc were studied. (auth)

This study was undertaken to determine the most feasible and economical process or path for closing the uranium fuel cycle at HAPO, and to establish what benefits, other than improvement in FPD`s competitive position, would result from the selected closed fuel cycle. The study was separated into four phases; Phase I includes the selection and organization of plausible processes and the establishment of a realistic and effective evaluation procedure, Phase II will include evaluation and selection of an optimum path based on a medium range and a short range approach (5 to 10 years), Phase III will include evaluation and selection of an optimum path based on a long range approach (> 10 yrs.), and Phase IV will include refinement of previous work and issuance of a final report summarizing the study and the conclusions or recommendations which develop from the study. Phase I of the study has been completed. The purpose of this paper is to document the assembled data and the completed work.

The composition of dissolver off-gas was determined for seven different dissolvings of uranium in nitric acid (both with one and two dissolvers in operation) at irregular intervals during February, March, and April, 1958. Samples were taken at the Purex 293-A facility absorber inlet, scrubber inlet, and scrubber discharge. The absorber was operated with and without reflux, and the scrubber was operated with water or with approximately 10% sodium hydroxide. This memorandum describes the sampling procedures and analytical methods and gives the gas analyses found.

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The Dry Chemistry Process of the RM Line provides the means for converting plutonium oxalate to plutonium tetrafluoride. This process consists of drying the oxalate cake and then converting to the tetrafluoride by passing HF gas at elevated temperature through the filtered cake. Improvement to or replacement of the present Task II system is being investigated because of the high maintenance costs and unreliability of the present system. This study is one of several being conducted under the Chemical Processing and Reduction Design and Development Program for improving Task II operation. This report describes a Task II system using round filter boats and vertical hydrofluorination furnaces.

The isotope Mn{sup 56} is the major short half-life constituent of reactor effluent water. It has half-life of 2.59 hours and decays by emission of a 2.86 mev maximum beta to stable Fe{sup 56}. This report describes a rapid method for the quantitative determination of Mn{sup 56} in a radiochemically pure form for use as a routine control procedure. A carrier-precipitation procedure has been developed which successfully separates Mn{sup 56} from other radioisotopes in reactor effluent water. It involves addition of Mn{sup ++} carrier to the sample followed by oxidation to MnO{sub 4}, removal of radioactive cations by stirring with Dowex 50 cations resin, and precipitation as MnO{sub 2}{center dot}xH{sub 2}O. The MnO{sub 2}{center dot}xH{sub 2}O is dried and weighed and its radioactivity measured by beta counting. A self-adsorption and self-water and self-scatter curve for Mn{sup 56}O{sub 2}{center dot}xH{sub 2}O was prepared by using a Mn{sup 56} spike prepared from reactor effluent water. For samples of interest a correction for self-absorption and self-scatter will be unnecessary. 2 refs., 2 figs., 3 tabs.

Neptunium-237 is currently produced in the Hanford reactors at a rate of approximately .003 gms/MWD via the following reactions: (a) U{sup 238} (n,2n) U{sup 237}{sup {beta}}{yields} Np{sup 237}. (b) U{sup 235} (n,{gamma}) U{sup 236} (n,{gamma}) U{sup 237}{sup {beta}}{yields} Np{sup 237}. In order to calculate the buildup of Np{sup 237} via reaction (a), which accounts for the greater share of the formation of Np{sup 237}, the n,2n cross section for U{sup 238} must be known. An old value quoted by Arnold of 5.2 millifermis for an {open_quotes}effective{close_quotes} 2200 m/s value is not large enough to account for the observed Np{sup 237} yield by about a factor of two. Recent n,2n cross section measurements for U{sup 238} permit a newer calculation and the result is 11.2 mF, effective 2200 m/s value.

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In this report, the equation (HW-55219-RD) expressing tube-wise rupture rate as a function of exposure, power, and temperature has been adjusted to include 1958 data.

This report describes the operation of the fuels preparation department for the month of July 1957. Manufacturing, employee relations, process development, plant improvements, and financial operations are described.

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The slow neutron transmission of a sample of enriched uranium containing 3.193 gm/cm{sup2} has been investigated with a resolution width of 1 microsec/m (1 microsec/m is the full width of the triangular resolution function at the base) and with points spaced 1/3 microsec/m. The results of these transmission measurements are shown in Fig. 1. The solid line represents the new measurements while the dashed line represents the previous measurements (CUD-4) on this sample using a resolution width of 1.4 microsec/m and with points spaced 2/3 microsec/m.

It is desired at this time to present in summary form a factual history of the Building 234-5 operations, encompassing not only the immediate pre-start-up period from April to July, 1949, but including the major obstacles encountered in the succeeding interval to the end of 1949. This report is intended to serve not only as a review of 234-5 operations during this period but also as a documentary evidence of the difficulties experienced and the manner in which they impeded the start-up. This report presupposes a certain familiarity with the background of the 234-5 Building. To one unacquainted with the scope of this project, or for those desiring a background review, reference should be made to Document HM-253, ``234-5 Building Program Review``, by D.D. Streid, dated April 22, 1949. It must be realized that any attempt to discuss in detail all the circumstances and difficulties contributing to the start-up period would be disconcerting to the reader, and would very likely be repetitive. Therefore, although it is far from the writer`s intent that a cursory review will suffice, the presentation will be as nearly as practicable, chronological and comparative. For the sake of clarity it will be necessary to present several …

In the P234-5 Building operation, Task I consists of converting plutonium nitrate solution received from the primary separations plants to a plutonium oxalate slurry. The slurry is filtered and the filter cake is transferred to Task II while the filtrate is treated and returned for further processing in either Recuplex, Redox, or Purex. The basic Task II process consists of calcining the plutonium oxalate cake to form plutonium dioxide. The plutonium dioxide is then fluorinated to form plutonium tetrafluoride which is transferred to Task III for further processing. Various alternate schemes for increasing production have been studied by the Technical and Design Sections during the past several months. The results of these studies are available in references (10) and (11). It is the purpose of this document to describe in detail the new equipment required for replacing Task I and II and to serve as a basis for detailed design of the proposed project.

The purpose of this report is to summarize briefly the work which has been done to date and to acquaint the reader with the results of the ``Hot Spot Investigation Program `` as well as some of the problems encountered.

Early in the development of an extended surface fuel element for use in the NPR, several 7-rod cluster fuel elements were irradiated to determine the dimensional stability of such geometries at high burnups. These elements were fabricated from small diameter uranium rods clad unbonded in stainless steel tubes and assembled in a rod cluster geometry by various support devices. Zircaloy clad fuel rods were not yet available, the stainless steel clad rods therefore served as a suitable material which would withstand high temperature water over a long period of time and maintain relatively high strength properties. The purpose of the irradiation detailed in this report was to determine the effect of high exposure on the swelling, dimensional stability, microstructure, and physical properties of uranium rods restrained unbonded in stainless steel. At the same time, this test was designed to evaluate the effect of fuel rods operating in a cluster geometry, to monitor the central core temperature of the uranium, to determine the stainless steel-uranium interface heat transfer bond coefficient, and to determine the average specific power of the assembled element. Goal exposure for this irradiation test was 3500 MWD/t.