>The optical radiation that occurs in a dynamic plasma pinch experiment depends in part on the amount and type of impurity atoms present in the system. By proper choice of these impurities, it should be possible to tailor the optical radiation from the pinch in intensity, spectrum, and duration such that the kinetic energy of motion (1OO to 1OOO joules) appears as radiation enengy in less than 10/sup -6/ sec. This should provide a good method of optical pumping for pulse optical maser operation. The optical radiation from the dynamic pinch is discussed and an experiment that is being set up to use this radiation as the pump source for an optical maser is described. (auth)

Adhesives. The effects obtained when diols and triols are used to cure Adiprene L-213 are discussed. Most of the formulations are very viscous and present difficulties in degassing operations. Ionium Project. Four plant samples having 1 ppm or more of Th/sup 2//sup 3//sup 0/ were analyzed for Th/sup 2//sup 3//sup 0/ in two different ways, one using HNO/sub 3/ digestion and the other using HClO/sub 4/ digestion. The difference between these two methods found for one sample is attributed to insolubility induced by calcining. Half Life of Radium-223. The decay of a purified Ra/sup 2//sup 2//sup 3/ sample was followed by alpha counting for 109 days; the results indicate that a long-lived impurity may be the cause of the nonconvergence of the probable error in the resolution time range. Purification of a composite sample containing Ac/sup 2// sup 2//sup 7/ to give a source of Ra/sup 2//sup 2//sup 3/ is described. Determination of Coincidence Correction. The coincidence correction was determined for a proportional alpha counter with Pb/sup 2//sup 1//sup 1/, and the best resolution times and half lives are given. Plutonium Alloy Research. The density of liquid Ce was measured from 825 to 1000 deg C with the vacuum pycnometer …

This report summarizes the IPD predicted vs. CPD measured values from twelve two-ton test batches. Two sets of comparisons are provided: CPD measured vs. IPD predicated based on techniques used since January 1, 1960 and CPD measured vs. IPD predicted based on techniques used prior to January 1, 1960.

This document provides the final report on the research project of chemical milling of counterbore recesses in the uranium wall of Zircaloy-2 clad uranium tubes. The scope of work, results and conclusions, and experimental procedures are provided.

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.

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 …

This document discusses a proposal that certain presently installed reactor process water outlet temperature data logging equipment in subject reactors to be replaced with new functionally simplified equipment of a more adequate design. The primary purpose of the proposed installation is to replace existing equipment which is obsolete and in three reactors is worn out to the point where the equipment is out of service frequently for periods of time up to 8 hours or more. The new equipment will provide reliable process tube temperature information for use in the functions of reactor control and product accountability. Based upon anticipated incremental production gains resulting from use of the new equipment, the amortization period for the project is calculated at 2.7 years.

Several safety parameters are calculated for a large, fast, Na-cooled, oxide-fueled reactor. The Doppler coefficient is studied as regards its effect on the reactivity during a power excursion, and its dependence on the neutron spectrum, on the concentrations of U/sup 238/, Pu /sup 239/ , a nd Pu/sup 240/, on spatial temperature and power distributions, and on temperature. Other parameters studied include the Na temperature coefficient of reactivity, and the reactivity insertion caused by total Na loss; reactivity coefficients caused by thermal expansion of fuel and steel cladding; the excess operating reactivity; and the reactivity caused by a fuel slump. The effects of a Be reflector on neutron lifetimes are determined. The safety aspects are balanced against economic considerations. (T.F.H.)

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This report, from the Chemical Processing Department at HAPO for February 1961, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations, facilities engineering; research; employee relations; and special separation processing and auxiliaries operation.

The transport of fission products from points of origin in unclad graphite matrix-type fuels to the reactor circulating system involves, as one of the steps. diffusion through the graphite matrix to the fuel element surface. As pointed out by Rosenthal. the fraction of a given fission product chain actaally reaching the fuel element surface will be small if the time for transport through the graphite is long compared to the half-lives of the volatile members. An important problem, therefore, is the determination of the effective transport rates of the various mobile elements and their daughter products of interest through various graphites suitable for use as fuel element compacts. as functions of temperature over the range of greatest immediate interest to reactor designers. The upper end of the range need not exceed about 1000 deg C. The transport of helium and arbon through various graphites has been the subject of considerable study by Watson. Evans, and other,. and a prelimilnary investigation of the high temperature transport of some ordinarily non-volatile elements has been carried out by Saunders. This work is briefly reviewed in relation to the final problem and the areas in which further information is needed most by reactor designers …

A mass flow measurement technique was devised which meets many requirements for mass flow measurement systems. In this technique. fluid is made to pass through a pivoted S-shaped tube. which tends to rotate due to the angular momentum of the fluid. A torque motor and torsion spring maintain zero displacement. the restoring force being a measure of the angular momentum, and density is measured by absorption of radiation passing through a window in the S tube. From the measured angular momentum and density, the mass flow can be computed. Two experimental mass gages using this technique have been constructed. (D.L.C.)

A test was run to determine the capability of the nuclear instrumentation system to indicate reactor transient behavior and to test the linearity and alignment of all four channels. The response to a resctivity insertion was found to be linear in the range tested 5 x 10>s-/sup >> / 5 x 10>s-/ sup > /p Channel A showed the best aligmnent. (D. L.C.)

ABS>A test was made on the Radioactive Waste Disposal (RWD) system to determine the acceptability of procedures used in containing, processing, and disposing the wastes received from the plant during a steady-state operation. The RWD system was found to be adequate and to have 60,000 gallons available at all times in the Surge and Decay tanks for safety injection. (D.L.C.)

A flowsheet, based on laboratory-scale data, is presented for oxyhydrochlorination of 90% uranium --10% molyhdenum alloy with 15% HCl in air at 4OO deg C in 18 hr. Up to 90% of the molybdenum is volatilized during oxyhydrochlorination and another 3 to 6% is removed by a 2-hr treatment with pure hydrogen chloride at 400 deg C. Residual chloride is removed by a 4-hr treatment with moist air at 400 deg C, and the product uranium oxide is dissolved in 4M nitric acid to yield a stable solvent extraction feed solution of 1M uranium, 0.017M molybdenum, 175 ppm chloride, and 1.7M nitric acid. The stainless steel cladding of the original fuel would be removed mechanically and the core recanned in aluminum prior to transfer to the core processing facility. The aluminum can would be removed by hydrochlorination prior to core treatment. (auth)

The objective of this production test is to evaluate the behavior of large diameter tubular UO{sub 2} fuel elements during high temperature irradiation. Eighteen inch long tubular UO{sub 2} fuel elements 1.804 inch OD, 0.544 inch ID, in 0.060 inch Zircaloy-2 jackets will be irradiated in the KER loops either alone or in conjunction with other tests to exposures up to 3500 MWD/T of contained uranium.

A new model of the higher symmetry of elementary particles is introduced ln which the eight known baryons are treated as a supermultiplet, degenerate in the limit of unitary symmetry but split into isotopic spin multiplets by a symmetry-breaking term. The symmetry violation is sscribed phenomenologically to the mass differences. The baryons correspond to an eight-dimensional irreducible representation of the unitary group. The pion and K meson fit into a similar set of eight particles along with a predicted pseudoscalar meson X/sup o/ having I = 0. A ninth vector meson coupled to the baryon current can be accomodated natarally in the scheme. It is predicted that the eight baryons should all have the same spin and parity and that pseudoscalar and vector mesons should form octets with possible additional singlets. The mathematics of the unitary group is described by considering three fictitious leptons, nu , e/sup -/ , and mu /sup -/, which may throw light on the structure of weak interactions. (D. L.C.)

The electrical resistivities of polycrystalline samples of La, Pr, Nd, and Sm are reported in the temperature range 1.3 to 300 deg K. La exhibits a superconducting transition at 5.8 deg K. The curve for Pr has slope changes at 61 and 95 deg K. The Nd curve shows small jumps at 5 and 20 deg K. Sm shows slope changes at 14 and 106 deg K. (auth)

This is the monthly report for the Hanford Laboratories Operation, February 1961. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

Design, development, and testing efforts were continued on BORAX-V, EBR- I, EBR-II, EBWR, JUGGERNAUT ZPRIII, ZPR-VI, and ZPR-W. An evaluation program is outlined for Pebble Bed Reactor designs. Fast and thermal reactor safety studies were conducted. Experimental and theoretical studies in applied nuclear and reactor physics are dsscribed. Developments made in reactor components, fuels, and materials are discussed. Heat engineering studies were conducted on steam separation, and velocity and void distributions in two-phase systems. Fluidization and fluoride volatility separation, and chemical-metallurgical separation processes were studied. Advanced reactor concepts that were discusssd includsed. Basic Radiation Effects Beactor, Biogeonuclear Reactor, Fast Reactor Test Facility, compact high-power density fast reactors, AHFR hydraulic test loop, Packed Bed Reactor, and direct conversion. (For preceding period see ANL- 6328.) (B.O.G.)

Results are summarized for a survey of methods for predicting fuel element and coolant behavior in sodiumcooled fast reactors under abnormal operating conditions. Circumstances prior to, during, and after meltdown are considered, and the behavior of fuel and coolant during transients is discussed. Reactor excursions, theory of fuel element thermal transients, reactor coolant transients in single and two-phase flows, and specific topics such as two-phase flow patterns, two-phase pressure drop, and transient voids in boiling coolant reactors are also treated. Some applicable mathematical treatments are given along with an annotated bibliography. (D.L.C.)

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