The electron paramagnetic resonance (EPR) of color centers in additively colored KCl crystals is measured to observe the effects of optical bleaching at room temperature. Earlier measurements on the F-center are confirmed and the susceptibility is measured at 78 and 300 deg K over five decades of power, including the very low power region. The width and the saturation properties of the individual multiplets are studied in detail and the technique of making E PR measurements on inhomogeneously broadened lines is discussed. A calculation is presented which shows that a slight departure from a Lorentzian multiplet shape can account for the saturation data. The bleached crystals show a resonance which has a width of 35 gauss and a different rate of saturation than the F- center. This resonance is associated with the B-band which appears in the optical absorption. (auth)

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)

A test was carried out to determine the adequacy of storage capacity and operating procedures of the radioactive waste disposal system during a normal reactor plant warmup. The capacity and operating procedures were found to be adequate. It was impossible to perform a complete material balance based on existing level instrumentation and using the data required by the test procedure. Approximately 21,290 gal. of waste were received in the system and 13,210 gal. were discharged to the river with a total activity of 1200 mu c. A quantity of 6,670 gal. of reactor coolant effluent was processed. Approximately 634 lb of combustible waste were incinerated. (M.C.G.)

Plant drawings for the final design for the Army Reactor (PL-2) are presented. Two hundred and twenty-eight figures are included. (M.C.G.)

>The basic fuel element consisted of a uniform dispersion of fuel in a 1 1/2 inch diameter graphite sphere. Ceramic coatings for the retention of fission products were studied. It was found-that molecularly deposited'' ceramics such as alumina, siliconized silicon carbide, and pyrolytic carbon were excellent barriers to fission product leakage. The most advantageous location for ceramic coatings was found to be on the individual fuel particles, where the coating was subject to smaller forces and where a larger thickness-todiameter ratio could be used than if the coating were on the surface of the graphite sphere. Fuel elements were irradiated to burnups ranging up to about 6 at.% U/sup 235/. In all specimens containing a uniform dispersion of fuel, the graphite spheres were found to retain their structural properties after irradiation. Data are given on fuel particle coatings of A1/sub 2/O/sub 3/, pyrolytic carbon, and metals: surface coatings of siliconized silicon carbide, pyrolytic carbon, and metal carbides; properties of and the effects of irradiation on graphite spheres; the use of natural graphite in preparing a high-density matrix material; graphite fueling by thorium nitrate infiltration; subsurface metal and metal carbide coatings for graphite; and an in-pile loop program on the behavior …

The physical and mathematical reactor models which are used in Program ODD are discussed. In addition, the FORTRAN II source program listings, decimal data input sheets, and input and output for a sample case are given. Program ODD was designed to raake use of the Revised Nuclear Data System at ANPD which consists of twenty-five energy group cross-section data including high energy inelastic scattering matrices, resonance parameters for the resolved resonances, and thermalization scattering matrices for the near thermal energy region. The most unique aspect of the program is the mathematical technique employed for eliminating inner iterations and slow convergenc rates occasioned by the up- scattering'' in the thermalization region of the energy lattice. Direct inversion of the energy matrix coupling the thermal and last four epitherma groups provides simultaneous consistent solutions for thes groups within each power iteration. (auth)

A critical experiment program is carried out in a configuration similar to the BONUS reactor. The results give information concerning: the effects of different boilersuperheater geometries; the reactivity changes associated with superheater voiding or flooding; power regulation between the boiler and superheater regions; epithermal transmission probabilities for B-stainless steel and Cd control rods; the power flattening characteristics; and void simulation properties. The calculational methods used in the study predict the measured reactivity and power distribution to within the limits of experimental accuracy. (T.F.H.)

Drawings andd specifications for the PL reactor core are given. The requirements for the material procurement, fabrication, testing, and inspection of one P L-2 reactor core and spares are listed. (M.C.G.)

tems were cast and cured. Results of chemical tests on aa epoxy curlang exudate are included. Comparison of solvent effects on retention of radioelements by stainless steel was started and data are tabulated for Ac/sup 227/, Th/sup 227/, a nd Ra/sup 22//sub 3/. Work on protactinium was resumed after suspension of this project in 1960. Methods for preparation of small quantities of highly enriched U isotopes are being examined. Included in the survey are chemical exchange, electromagnetic separation, gaseous and liquid thermal diffusion, gas centrifugation, and photochemical techniques. Continued investigation of viscosities of La and Pr for use in Pu alcontinued along with studies of Pu bearing glass fibers. (J.R.D.)

Apparatus for the performanee testing of SNAP 10A thermoelectric elements was designed, constructed, and is now in operation. Elements may be tested for any desired length of tfme up to 1400 deg F and in a vacuum of 1 x 10/ sup -5/ of Hg. The equipment used for these tcsts may also be utilized for measuring Seebeck coefficient and resistance as a function of temperature. Element performance is derived from the data on voltages and temperatures. The performance variables which are reported in graphic form are as follows: loaded output voltage at any desired DELTA T; open circuit output voltage at any desired DELTA T; power output under optimum load conditions; current produced under matched load conditions; and internal resistance of the element. (auth)

A detailed analysis was made of the power distributions in the SM-2 experimental core at 68 deg F and the SM-2 reference core at 510 deg F. This analysis supersedes the power distribution calculations presented in APAE No. 69. The calculated distributions for the experimental core were normalized to measured data wherever possible in order to obtain corrections factors for application to the reference core. The over-all power distributions were calculated by synthesis of one-dimensional axial calculations with twodimensional radial calculations. The variation of the power distribution with fuel burnup is also presented. In order to improve the agreement between measured and calculated axial power distributions, flux-weighting the nuclear parameters in the rods-in and rods-out regions was investigated. (auth)

Specifications for the plant equipment for a P L-2 nuclear power plant are given. (M.C.G.)

None

The activation energy, activation volume, snd frequency factor were evaluated for yielding (delay time for yielding, upper yield stress, lower yield stress, and Luders band propagation) and flow (friction stress, flow stress, and dislocation mobility) for various irons and steels from data in the literature. It was found that the values of these flow parameters and their stress dependence were the same, within experimental error, for both yielding and flow, and for all the materials considered. This suggests that either the same dislocation mechanism is controlling in every case, or that one or more mechanisms possees approximately the same values for these parameters. The dislocation mechanism for which there was closest agreement between theoretical calculations snd experimental data was overcoming the Peierls stress. On the basis of the available experimental data and the present analysis, it is suggested that the upper and lower yield stresses in iron and steel may represent the sudden generation of a large number of dislocations by the double cross-slip mechanism of Koehler and Orowan, rather than the breaking away from a Cottrell atmosphere. (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).

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.

None

The plant design for PL-2, a 1000-kw net electric direct cycle boiling water nuclear power plant, is presented. The design includes all buildings, foundations, and structures required for the installation of the plant in a snow tunnel. (M.C.G.)

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)

None

Plastic Research. The tensile strength of Dacron-filled diallyl phthalate was determined to average 4377 psi. Composition and stress-strain data are tabulated for ten adhesive films. Analytical studies of an adhesive exudate are reported. Radioelements. Results of analysis of ioniumbearing raffinates and residues for Th/sup 232/ and of aged>s Ra/sup 223/ for Ac/sup 2/2/sup >/s7>s are given. Progress on Pa recovery from raffinates and residues and separation from Nb is reported. Isotope Separation and Purification. Proposed work on gas centrifugal and photochemical separation of uranium isotopes is discussed. Progress on xenon and helium isotopes separation and purification is outlined. Reactor Fuels and Materials Development. The density of liquid La at 945 to 1000 deg C was determined. The performances of an oscillating Cup viscometer with La and Bi and of a high-temperature calorimeter with Po/sup 210/ are described. A study of the compatibility of Haynes 25 alloy with Pu at 9O0 deg C indicated that very little penetration took place after the first hour at 900 deg C. The efficiency of escape of alpha particles from a glass fiber containing 10 wt% pu oxide was determined to be approximately 72%. Eight glass compositions were evaluated for their ability to dissolve 15 …

Uncontrolled releases of large amounts of iodine occurred during the early stages of RaLa operation at the ldaho Chemical Processing Plant. A ten- fold reduction in the iodine content of the off-gas was achieved by process modifications, primarily the addition of mercury salts to the acidic process solutions. An additional ten-fold reduction was obtained by installing an activated charcoal adsorption unit in series with the original iodine removal scrubber. The iodine content of particulate entrainment limited the over-all iodine removal efficiency of the revised RaLa off-gas iodine removal system. (auth)

Experiments with unirradiated Consolidated Edison reactor ThO/sub 2/-- U0/sub 2/ fuel pellets indicated that uranium losses to Sulfex and Darex decladding solutions were proportional to the U0/sub 2/ content of the pellets. For example, after 7 hr, losses to boiling initial Darex solution (5 M HNO/sub 3/- -2 M HCl) were 0.45 and 0.65% from pellets containing 3 and 9% U0/sub 2/, respectively. The initial rate of dissolution of these pellets in 200% excess boiling 13 M HNO/sub 3/--0.04 M NaF-0.1 M Al(NO/sub 3/)/sub 3/ was essentially independent of the U0/sub 2/ content. Rates were 2.1, 3.0, and 2.4 mg min/sup -1/ cm/sup -2/ for pellets contdining 3, 6, and 9% U0/sub 2/, respectively. The presence in the dissolvent of the soluble neutron poisons H/sub 3/B0/sub 3/ and Cd(NO/sub 3/)/sub 2/ in concentrations up to 0.1 M and 0.075 M, respectively, had little effect on the rate of dissolution of sintered UO/sub 2/-- ThO/sub 2/ fuel pellets. (auth)

This report summarizes experimental work completed on Subcontract AT-147.