Recoil ranges of C11 from the reaction C12(p,pn)C11 are presented for incident proton energies from 0.25 to 6.2 Gev. From these data it is concluded that a neutron evaporation mechanism cannot be the major mechanism. The result for incident energies of 3 and 6.2 Gev are consistent with a fast reaction consisting of a single inelastic nucleon-nucleon collision. Assuming this mechanism, an average kinetic energy of 19 Mev can be deduced for the struck neutron (before the collision) in the C12 nucleus.

7 7 6 4 : 9 8 9 6 5 : 9 aluminum nitrate waste was calcined in the close-coupled continuous evaporator-pot calciner unit. Waste oxides from TBP-25 waste were incorporated into glassy materials after the addition of phosphate and borate fluxing agents. Melts formed at 850 to 950 deg C were glassy solids when cooled and had densities from 2.41 to 2.47 g/ml. Waste volurne reduction factors were from 7.6 to 9.3. Low-Level Waste Treatment. A demonstration run was completed in the 60 liters/hr scavenging-ion exchange pilot plant on ORNL low-activity waste. Decontamination factors were satisfactory after 1500 resin bed voluumes of waste had been treated, i. e, Sr> 1000, TRE 100, Cs> 100, and sufficiently high for other activities that the waste activity was reduced to <10% of MPC. The physical properties of vermiculite, clinoptilolite, and rock phosphate were found to be suitable for second-stage treatment of process waste. Engineering, Econommc, and Hazards Evaluation. A cost study of the conversion of high-level solutions to solids by pot calcimation was completed. Aging had a negligible effect on costs for processing in a given vessel size. The lowest cost was 0.87 x 10/sup-2/ mill/kwh/sub 3/ for processing acidic Purex and …

Future operation of B-Plant for Phase I fission product processing involves may of the hazards associated with current chemical Processing Department activities in both the Purex and Redox Plants. The specific B-Plant hazards under Phase I operations are discussed in this report from the standpoint of comparable Purex and Redox hazards.

An investigation was conducted to determine which tubes of the 1A loop heat exchanger are leaking. Air pressure and probing tests are inconclusive and cannot be used to verify chemical sampling. (J.R.D.)

The chemical feasibility of the direct, continuous reduction of UF/sub 6/ to U with Na was shown in several tests. Up to 93.5% of the U content of UF/sub 6/ continuously reduced by Na in a reaction vessel was recovered as massive U metal of acceptable purity. A semicontinuous reactor for continuous reduction is described. (D.L.C.)

The flowsheet visualized from development work thus far for final plutonium recovery and purification will accept as feed a Purex partition stream without feed adjustment beyond the usual reoxidation. Extraction with trilaurylamine at approximately 0.3M appears suitable for 20 to 60 g Pu/liter product from 0.5 to 2 g Pu/liter feed. Scrubbing with either ((2 M or))2 M HNO/ sub 3/ is possible. Acetic acid is at present the first choice for stripping agent, with oil-soluble and aqueous-soluble organic reductants as alternates. (auth)

The performances of aqueous-homogeneous (AHBR), molten-salt (MSBR), liquid-bismuth (LBBR), gas-cooled graphite-moderated (GGBR), and deuterium- moderated gas-cooled (DGBR) breeder reactors were evaluated in respect to fuel yield, fuel cycle costs, and development status. A net electrical plant capability of 1000 Mwe was selected, and the fuel and fertile streams were processed continuously on-site. The maximum annual fuel yields were 1.5 mills/ kwhr. The minimum estimated fuel cycle costs were 0.9, 0.6, 1.0, 1.2, and 1.3 mills/kwhr at fuel yields of were 0.9, 0.9, 1.5, 1.5, and 1.3 mills/kwhr. Only the AHBR and the MSBR are capable of achieving fuel yields substantially in excess of 4%/yr, and therefore, in view of the uncertainties in nuclear data and efficiencies of processing methods, only these two can be listed with confidence as being able to satisfy the main criterion of the AEC longrange thorium breeder program, viz. a doubling time of 25 years or less. The development effort required to bring the various concepts to the stage where a prototype station could be designed was estimated to be least for the AHBR, somewhat more for the MSBR, and several times as much for the other systems. The AHBR was judged to rank first in …

The performances of aqueous-homogeneous (AHBR), molten-salt (MSBR), liquid-bismuth (LBBR), gas cooled graphite-moderated (GGBR), and deuterium-moderated gas-cooled (DGBR) breeder reactors were evaluated in respect to fuel yield, fuel cycle costs, and development status. A net electrical plant capability of 1000 Mwe was selected, and the fuel and fertile streams were processed continuously on-site.

The performances of aqueous-homogeneous (AHBR), molten-salt (MSBR), liquid-bismuth (LBBR), gas-cooled graphite-moderated (GCBR), and deuterium- moderated gascooled (DGBR) breeder reactors were evaluated in respect to fuel yield, fuel cycle costs, and development status. A net electrical plant capability of 1000 Mwe was selected with continuous processing of fuel and fertile streams. The maximum annual fuel yields were 16, 7, 4, 4, and 4.5%/yr, respectively at a fuel cycle cost of 1.5 mills/kwhr. The minimum estimated fuel cycle costs were 0.9, 0.6, 1.0, 1.2, and 1.3 mills/kwhr at fuel yields of 7, 1, 1, 2, and 3%/yr. At a fuel yield of 4%/yr, the costs were 0.9, 0.9, 1.5, 1.5, and 1.3 mills/kwhr. Only the AHBR and the MSBR are capable of achieving fuel yields substantially in excess of 4%/yr, and therefore only these two can be listed with confidence as being able to satisfy the mdin criterion of the AEC long-range thorium breeder program i.e., a doubling time of 25 years or less. The development effort required to bring the various concepts to the stage where a prototype station could be designed was estimated to be least for the AHBR, somewhat more for the MSBR, and several times as much for …

The performances of aqueous-homogeneous (AHBR), molten-salt (MSBR), liquid-bismuth (LBBR), gas-cooled graphite-moderated (GCBR), and deuterium-moderated gas-cooled (DGBR) breeder reactors were evaluated in respect to fuel yield, fuel cycle costs, and development status. A net electrical plant capability of 1000 Mwe was selected, and the fuel and fertile streams were processed continuously on-site.

A test was performed at the Shippingport PWR, whose purpose was to determine the amount of heat the fuel shipping container cooling system was capable of removing while in a state of equilibrium. Studies were also conducted to determine if after four hours of operation, the increase of the internal wall temperature was less than 30 deg F, and finally, to determine the effects of loss of cooling water from the fuel shipping container. (auth)

A program was written to perform polynomial least squares fits on the IBM 7090 computer. Weighting factors may be included with the data if desired. The program includes a subroutine which does the fit and furnishes the calling program with the coefficients, standard errors in the coefficients, staandard error of fit, and information necessary to compute a complete error analysis; and a calling program which reads input, calls the subroutine, and writes requested output. The subroutine may be used separately in any Fortran or FAP program where a least squares fit is needed. (auth)

Program written to perform polynomial least squares fits on the the IBM 7090 computer.

The radiation survey of the Blanket Exit Water Instrumentation assemblies Core I, Seed 1 was performed on January 15 and 16, 1960, approximately 100 days after reactor shutdown at 5806 EFPH. The observed radiation levels varied with height from 150 to 500 mr/hr; the highest activity was found at the bottom of both support tubes of each of the four assemblies monitored, which, in operation, would be located 5 in. above the top of the blanket fuel assembly. The two inner port Blanket Exit Water Instrumentation units and the two Auxiliary Seed Exit Water Instrumentation assemblies were not monitored. (D.L.C.)

C Reactor was shut down on a scheduled basis at 8:30 a.m. March 6, 1961 for the purpose of overboring 17 process channels. this report will cover that outage and discuss problems encountered in completing the tasks involved in overboring.

Annual report submitted by the Boy Scouts of America to Congress describing highlights from 1960, activities, growth programs, public relations, resources, organizational leadership, and other information about scouting programs.

Helium-permeability and porosity were determired at room temperature for specimens from a typical EGCR moderator-grade graphite block. Permeability, at a mean pressure of 2 atm, ranged from 26 to 200 (av. 86.5) millidarcys. Permeability data indicated that turbalent flow was never obtained with helium in these tests and that helium permeating the moderator graphite at EGCR operating conditions (taken to be: 600 deg C; DELTA P, 10 lb/in./sup 2/ per inch of graphite; mean P, 400 lb/in./sup 2/) was in the viscous flow region. Daroy's law and the reported constants are applicable for flow computations involving moderator graphite under these conditions. Porosity ranged from 20.6 to 29.4% (av. 23.8%), and there was no correlation between porosity and pemaesbility variations. The large variations encountered were believed to reflect the nonuniformity of the specimens, since duplicate determinations showed excellent agreement. Permeabilfty did not change appreciably with direction of flow and did not vary consistently with respect to the extrusion or any other axis. Preparation of the specimens did not appear to introduce appreciable surface effects. (auth)

The objective of this production test was to evaluate the effects of Zr-2 jacket uniformity, thickness, and temperature on non-uniform clad straining. Test details are provided.

A detailed account of the reactor physics studies for the final conceptual design of the Advanced Test Reactor is presented. The diffusion theory methods used for calculations of flux distributions and reactivity effects are described and compared with measurements and with higher order approximations to transport theory. These comparisons show diffusion theory to be adequate for the ATR conceptual design. Two-dimensional flux distributions for a number of shim control conditions and experimental loadings were determined by PDQ-3 and TRANSAC-PDQ. The worths and effects on flux distributions of chemical and of blade type mechanical shim controls were compared. The effects of heavy water and of beryllium reflectors on reactivity and flux pattern were calculated. The time-dependent behavior of the reactor was investigated by use of TURBO and CANDLE. The changes in shim control poison and test and core flux distributions with fuel burnup were calculated and the full-power cycle time estimated. An investigation was made of the xenon transient after a fullpower shutdown and recovery. Results of one- and twodimensional fuel depletion studies are compared. The results of a number of time independent one-dimensional calculations and parametric studies are presented. Some comparisons were made of the results for one-dimensional and two-dimensional …

Descriptions are given of various procedures used in determining leaks in the tubes of the 1B heat exchanger. Air pressurization tests determined leakage and leak rate of nine tubes. The leak-location-detector-probe method was found promising for locating defects along the length of the tube. Results of the probalog, dye-penetrant, and ultrasonic tests proved inconclusive in determining leak locations. (B.O.G.)

Report discussing discrete energy levels observed within certain crystals which are due to perturbations of energy levels of the free ion by an electrostatic field arising from the crystal lattice. The analytic procedures for determining the field from the charge configuration are given, and the resulting fields are classified according to their symmetry. After a general survey of group-theoretical ideas, the applicable groups are analyzed in detail, and characters appropriate for both integral and half-integral angular momenta of the free ion are tabulated. Text includes tabulations, equations, and matrices using Wigner and Racah coefficients.

This report is about the congressional priviledge, because of which people could be immune to liabilities for Slander and Libel.

The purpose of this study is to determine if a hazardous condition is created by accidental closing of the coolant loop isolation valves�while the reactor is operating.

None