A pump was developed for pumping carbon dioxide in a closed loop without introducing impurities. This pump will give flow rates of up to 3 liters/min and will develop a working pressure of over 70 mm Hg. No wear was observed after 2000 hr of testing. It is felt that this pump is more desirable for this application than those developed by other experimenters for two reasons: relatively inexpensive construction of the pump and the associated electronic circuit and the low coefficient of friction between the piston and the cylinder wall. Various modifications are suggested which will make this pump satisfactory for other applications. (auth)

It is requested that a design be initiated for the primary piping static test. This test is necessary to provide information as to the reliability of the pipe subjected to reactor operating conditions. The test conditions are as follows: temperature - 2000 F (isothermal), pressure effective - 42 psi, and test time - 10,000 hours. It will be necessary to test two sizes of pipe as shown on the preliminary piping layout (2.250-inch O.D. x .095-inch wall and 3 1/2 SCH. 10 pipe). The test specimens shall be jacketed in an inconel containment vessel. The test rig should be similar to the design of the 4-inch pressure vessels (T-1030244). In addition an outer containment vessel constructed of stainless steel must be provided around the clam shell heaters and the inconel containment vessel. This is to provide an inert atmosphere for the inconel vessel. Provisions should be made in the design for a 1/4-inch clad thermocouple. It is planned to use the pipe test as a vehicle for studying experimental Tc's (Cb-Mo and W-W.26% Re).

Laboratory-scale tests on methods for recovering uranium and thorium from graphite-base reactor fuel elements are reported. The 90% HNO/sub 3/ process, which involves simultaneous disintegration and leaching in 21 M HNO/sub 3/, is applicable to all fuel elenments which do not contain coated fuel particles. Leaching of irradiated (0.001% burnup) fuels containing 3 and 12% uranlum recovered approximates 99.3 and 99.9%, respectively, of the uranium in two 4-hr leaches with boiling acid. The graphite residue retained > 50% of the long-lived fission products. Three successive leaches of fuel containing uranium and thorium recovered approximates 99% of both elements. Uranium recoveries by combustion in oxygen followed by dissolution of the ash hn nitric acid or fluorlde-catalyzed nitric acid are quantitative only when the fuel is not coated, does not contain Al/sub 2/O/sub 3/-coated fuel particles, and is free from impurities such as iron. During combustion up to 95% of the Ru-106 was volatilized from irradiated specimens. Recoveries, by leaching with 70% HNO/sub 3/, from fuel specimens containing Al/sub 2/O/sub 3/-coated fuel particles were greater than 99% when the specimens were ground finer than 200 mesh to ensure crushing of the fuel particles. (auth)

It is requested that the design be completed for a full-scale fuel element soak test. The test assembly must be designed to permit the fuel element test specimen to be submerged in a lithium bath under a pressure of 60 psi. The maximum temperature of the lithium is to be 2000 F. A total of four test units will be required to complete the test program. Two specimens will be exposed to a thermal cycle between 2000 F and 1400 F with the remaining two specimens being exposed to a thermal cycle between 2000 F and 1000 F. Heating will be done at the rate of 200 F/hour preceded by a 150 hour soak at 2000 F. Heat will be supplied by clam-shell type heaters. The test specimen - lithium system will be contained by a Cb-lZr vessel which will be surrounded by a 310 steel container. The heating units will be mounted on the outside of this 310 S.S. container. A bottom fill line is requested in order to insure a lithium system free from gas pockets. A slow lithium fill will be made up through the specimen to a level indicated by a probe in the expansion tank.

The object of the production test described in this report is to evaluate the operational aspects of a proposed method for reducing minimum downtime. The excess xenon poisoning, which occurs during the first 32--38 hours after the shutdown of a reactor from present equilibrium levels, will be partially overridden by a central enriched zone whose added reactivity contribution would be compensated during normal operation by means of poison splines.

Work is being carried out to develop and demonstrate on pilot plant scale integrated processes for treatment and disposal of radmoactive wastes. High-level waste calcination, low-level waste treatment, economic and hazards evaluation, engineering evaluation, disposal in deep wells, disposal in natural salt formations, Clinch River studies, fundamental studies of minerals, and White Oak Creek basin study are discussed. (M.C.G.)

Structural evaluations of the upper head of the EGCR pressure vessel were made. The configuration throughout the cluster region in the vessel was found to be structurally adequate. The primary and primary-plus-secondary stress intensities for the burst-slug detection and gas outlet nozzles were found to be within the allowable limits. However, the complete design evaluations of these units cannot be made until the temperature distributions are known. (auth)

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A method is presented for the determination of uranium (as U/sup 238/) in uraniuni-loaded graphite fuel elenients by a non-destructive, direct radioactivity analysis technique. A 200-cbannel pulse-height analyzer, equipped with a 3 in. x 3 in. NaI(Tl) crystal, is used to measure the Np/sup 239/ radioactivity of the neutron-irradiated samples. The amount of U/sup 238/ in the test samples is deterimined quantitatively by comparing the Np/sup 239/ radioactivity induced in each sample with the Np/sup 239/ radioactivity induced into known standards of U/sup 238/ processed under the same conditions as the test samples. The limit of detection for U/sup 238/in samples of normal uranium composition for this method is about 1.0 x l0-4 ug. (auth)

Channel-to-channel flow distribution within Type 3 (SM-2, stationary and control rod fuel elements modified for use in the SM-1, SM1-1A, and PM-2A core support structures and control rod tubes was measured in single element flow testing. Plots of channel-to-channel flow distribution and element pressure drop at various element flow rates are given. Flow distribution for the top-orificed SM-1A and PM-2A stationary elements was within the plus or minus 12% deviation from element average utilized in previous thermal analyses of these cores. Testing of the bottom-orificed SM-1 stationary element and the SM-1, SM-1A, and PM-2A control rod assemblies showed flow distribution exceeded plus or minus 12% devation from average. Simple modifications to the SM-1 stationary element indicated the possibility of improvng fiow distribution in that element. (auth)

Preliminary design concepts are presented for a wire scanner for experimentally evaluating spatial variations of neutron flux in the SM-l reactor core. Results of a literature search and determination of optimum criteria for flux mapping the core in minimum time dictated requirements for design concepts and specifications. The utility of both manually instrumented and automatically instrumented wire scanners was analyzed with respect to rapidity of measurement, selectivity of detector location, cost, value of data, plant downtime, and additional factors. (auth)

Theoretical and experimental studies indicated that the effective permeability coefficient for graphite is lowered by a helium stream in-sweeping through the graphite pores. This phenomenon was considered in the design of HTGR fuel elements. A portion of the helium gas which is drawn into each fuel element as a purge stream may enter through porous wall sections, supplementing the purge gas entering at the top of each fuel element. The purge stream leaves each fuel element through a header system which carries the purge gas to an external fission product trap. The flow rate through the trapping system determines the upper limit of the average in-leakage through the fuel element walls. In the case of the HTGR, a graphite having a helium permeability of 1.1 cm/sup 2//sec at 350 psia and 700 deg F (approximately 0.1 cm/sup 2//sec at 14.7 psia, 70 deg F) would result in 100% of the purge flow entering through the wall sections of the fuel element. A lower permeability graphite, with most of the purge flow entering at the top of the fuel element appeared to be more desirable for maintaining optimum purge flow conditions. (auth)

It is noted that a uniform flat-top beam spill-out cannot be obtained in the Cosmotron by manual control. A servo system is proposed which will control this spill-out by sensing the external beam intensity, and correcting the magnet voltage to keep this intensity constant. This servo must operate through the transfer function of the main ignitron system and the flat-top filter. An analysis of these special transfer functions is presented. (J.R.D.)

The purpose of this Process Improvement Transition Authorization (PITA) is to authorize full pile loading bumper fuel elements in the fringe zone will be reviewed and, if desirable, recommendations to curtail fringe loading may be made based on economic considerations.

This document details activities of the irradiation processing department during the month of October, 1961. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering Operation; Employee Relations Operation; and Financial Operation.

Fifty-seven papers and l7 abstracts of papers are presented in the report. Fifty two of the papers are abstracted separately; in addition, a single abstract is written to cover the section on Plasma Physics Theory, which contains 3 papers and 8 abstracts of papers. The two brief papers not abstracted separately are concerned with fast neutron detection, and homogeneous critical assemblies of 3%enriched UF/sub 4/-paraffin systems. (T.F.H.)

Solutions of the Boltzmann equation for monoenergetic neutron transport in spherical geometry are derived which are respectively singular and regular at the center of the sphere. A few specific partial singular solutions are presented. The regular solutions in spherical geometry are constructed by superposition of solutions in plane geometry which belong to the same k. Finally, the solutions are compared with their representations by a series of spherical harmonics. (D. L.C.)

The 15,000 MW nuclear economy assumed for the long range study of pot calcination costs reported earlier was used as a basis for calculating optimum fill volumes. An algebraic expression was developed for cost as a functmon of the normalized radius of the central void space in a partially filled vessel. Minima of this expression were found for acmdmc and neutralized wastes in 6, 12, and 24in.-diameter vessels. Optimum fill volumes decreased as vessel diameter increased, varying for acidic wastes from 99.8% for 6-in.-diameter vessels to 92.5% for 24-in.diameter vessels. Decreases in costs by using optimum fill volumes instead of the 90% fill volume assumed for all cases in the long range study were small, the largest being an 8% decrease for neutralized wastes in 6- in.-diameter vessels. (auth)

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The combination of a large reactor core with a sufficiently high flux level results in complications of the dynamic behavior of the core due to asymmetric xenon and temperature instabilities. In both phenomena, a local increase in neutron flux causes an increase in neutron multiplication in a surrounding zone. We consider here the instability associated with the delayed production of Xe{sup l35}, an isotope with a very high thermal neutron cross section. In order for this instability to result in flux oscillations., the neutron production in one part of the core must be independent of conditions in another part. Said in another way this means that the neutron migration area must be considerably less than the square of any core dimensions. Thus, even though the total reactor power is held constant, it may be possible that neutron leakage from one part of the reactor to another is insufficient to counteract the tendency of the local flux to continue increasing in the part of the reactor where it is already high. An asymmetry in the distribution of reactor power may thus tend to increase in After some initial confusion., it has been found that close monitoring of the local power trends …

The percentage of helium which will bypass the core if the reflector system shown is used is predicted. It is estimated that nominally about 0.1 to 0.2% of the total flow will bypass the core, which is not considered excessive. The most difficult parameter to determine was Z, the gap between the sealing surfaces. The method used to predict Z is presented. The effect of bowing due to a temperature gradient across the seals is discussed. (auth)

A conceptual design and economic evaluation of 300 and 40 MW/.sub e/ steam-cooled fast breeder reactor power plants were performed. A reactor core composed of U-Pu oxide rod-type fuel elements clad with Inconel-X and surrounded by a blanket of depleted UO/sub 2/ fuel was studied in some detail. Reactor breeding ratios of from 1.27 to 1.5 and overall system doubling times of from 20 to 30 years are achievable. For the near term (1967) 300 MW/sub e/ plant, an energy cost of 7.6 mills/kwh is estimated, based on AEC ground rules for privately financed plants and utilities. This cost may go down to 5.7 mills/kwh by 1975. For the 40 MW/sub e/ plant corresponding energy costs are 19.5 and 13.7 mills/kwh, r -spectively. The R&D program required for this reactor concept is estimated at million with an additional million for improvements leading to the 1975 reactor. Investigation of the operational and safety aspects of the reactor indicated that satisfactory procedures can be used for startup, shutdown, and emergency cooling of the reactor. An increase in reactivity upon flooding can be prevented by incorprating small amounts of high resonance absorption material in the core. Preliminary calculations indicate a substantial increase in …

Studies carried out on corrosion associated with the hydrofluorination- dissolution phase in the fused-salt Fluoride Volatility Process are summarized. Corrosion for hydrofluorination-dissolver vessels used in bench-scale and semiworks-scale process development at ORNL is discussed. The results of a study on construction materials for the dissolution phase are presented. Corrosion studies at ANL are described for comparison purposes. A full-size hydrofluorinator dissolver is described. (M.C.G.)