An idealized model is postulated embodying the essential features of industrial caustic-chlorine cells with horizontal flowing-mercury cathodes. This model is examined in detail, and relations expressing the local anode potential, cathode potential, and ohmic potential drop in the electrolyte in terms of local current density and other parameters are established. These relations are combined to give a system of equations relating current density at any location along the cell to applied total potential and to operating conditions in the cell upstream of the point in question. Numerical solutions of these equations for several cases of cell operating conditions are carried out on a digital computing machine. The effects of changes in operating parameters upon average current density, individual electrode potentials, and current distribution are evaluated. (auth)

The Army Package Power Reactor (APPR-1) control system was analyzed to find its response to a startup accident. The system was found to be adequate, at 0.086%/sec rod withdrawal, for operation of at least twice the rated power. (C.J.G.)

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Correlation of a number of physical and chemical properties of 12 different UO/sub 2/ powders was continued. The UO/sub 2/ powders were studied by means of infrared absorption measurements, oxidation temperatures as determined by hot stage microscopy techniques, and B. E. T. surface area measurements. Additional pellets were prepared to study ceramic performance. (M.C.G.)

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The lifetime of a B/sub 4/C-in-tubes'' control rod may be limited by either nuclear worth depreciation or internal gas pressure buildup as helium is formed and released from the B/sub 4/C particles. Nuclear life is enhanced by thin tube walls for any given O.D. of tubing and by high B/sub 4/C density, but the time to failure from gas pressure buildup is increased by the opposite. The optimum B/sub 4/C density and tube wall thickness were calculated for maximum control rod life, using several release fractions of generated helium and a number of allowable cladding stress values. It was estimated that lifetimes of l0 or more years for B/sub 4/C-in-tubes control systems are achievable through the proper selection of tubing material and B/sub 4/C density values. The B/sub 4/C densities required for maximum lifetimes are in a range that can be obtained by vibratory compaction followed by swaging. (auth)

S>An inline densimeter was fabricated and tested in the laboratory. When operated in a range of 1.000 to 1.200 g/cc, an accuracy of plus or minus 0.2% in the mid-point of the range was easily achieved. The instrument also measures pulse amplitude and pulse frequency when used on a pulsating flow. A long life with high reliability is expected because of the simple all-welded construction and highly reliable electronic system. (auth)

To date a great deal of emphasis has been placed on flattening the side-to-side and top-to-bottom flux distribution with only minor effort to improve the front-to-rear distribution. Minor variations in the front-to-rear distribution have been achieved by horizontal control rod and Supplemental control positioning. It has-been reasonably well established that the rupture potential for one tube charge increases markedly with higher specific power and temperature; thus there is a great deal of incentive to flatten in the front-to-rear dimension. Although flattening in this dimension will caure increased neutron leakage out of the reactor, this is compensated by increased conversion efficiency resulting from a more uniform exposure distribution within the tube charge. The purpose of this document is to describe the basic analytical methods and the techniques, of flattening front-to-rear through the integrated use of enrichment and poison material in combination with natural uranium, and to point out the requirements to insure that total control criteria is satisfied in the event of a water loss with this loading. For the purpose of this survey report an old reactor, 32-piece charge length, and a symmetrical front-to-rear distribution were considered; however, the methods given can be extended quite easily to different length and …

Tables are presented for conversions from photon vacuum wavelength to photon energy and vice versa. The ranges covered are 500 to 12,000 angstroms and 1.000 to 26 electron volts. (D.L.C.)

Data are presented that show a marked species difference in the degree of damage and in the course of events seen in the bone marrow and peripheral blood after a given dose of radiation. The data taken together indicate strongly, but do not prove, that the picture seen results primarily from damage to the stem cell population. This damage to the stem cells is manifested later in the more mature differentiating cells of the marrow, and in the peripheral blood. The picture seen results both from death of stem cells, and also injury leading to reduced proliferative capacity, with later death of the injured stem cell and all of its progeny. Thus the picture seen is a result of both quantitative and qualitative changes resulting directly from the exposure of the stem cells, and species differences relative to sensitivity and time course of events appear to depend to a large degree on differential sensitivity of the bone marrow stem cell populations to death and injury. (auth)

Research was continued during the report period on the use of isotope neutron sources for producing short-lived radioisotopes. Experiments with a newly constructed betacounting cell are reported in which a 50-curie Be--Po neutron source was used. Study of the radiation chemistry of polymers was continued concerning the effects of polymer structure on free-radical formation. Free-radical formation in several additional polymers was studied. Preliminary work is also reported in an investigation of internal irradiation effects on the chemical activity of catalysts. (J.R.D.)

Under the conditions of PT-IP-247-A, four columns of self-supported and four columns of rib-supported I&E 1.47% enriched fuel elements were irradiated to determine their relative performance under severe operating conditions. Four of the self-supported and two of the rib-supported control elements were received for nation. The two rib-supported control pieces, which were classified as ``near failures`` had received average exposures of 353 and 359 MWD/T at average specific power levels of 91 and 108 KW/ft before they were discharged because of other ruptured pieces in the tubes. The nominal specified canned fuel dimensions of the rib supported elements was 1.445 in. OD, .310 in bore, and 7.640 in. long. The first two self-supported elements were selected for examination on the basis of high weight losses sustained during irradiation, and the second two were selected to determine the effect of specific power levels on the AlSi bonding. The average specific power levels of the four self-supported elements varied from 79 to 110 KW/ft and the exposure varied from 845 MWD/T to 949 MWD/T. The nominal canned dimensions of the self-supported elements, which were made oversize to attain high annular coolant temperatures with respect to the interior coolant channel, were 1.460 in. …

On September 18, 1967, KE Reactor experienced,a failure in a K5E fuel element in tube 2863, exposure 634 Mwd/t. On September 24, 1967, two additional failures occurred. All of these elements, clad with X8001 alloy were from the same canning lot. All fuel elements from this canning lot were termed rupture-prone and were discharged. The elements from five columns including column 2863, but neither columns 2865 nor 3256, were sent to 105-C Basin Fuel Examination Facility for further examination. An incipient failure was found and sent to Radiometallurgy in an effort to determine the cause of failure. During the course of examination, the spire was found to contain beehive-shaped growths approximately 1/8-inch high and 1/8-inch 9 in diameter at the base. The incipient failure contained four of these growths protruding from the spire wall. Closer examination showed that the spire contained numerous other ``beehives`` in less advanced stages of growth. The ruptured element from 2863 also contained a ``beehive``. The second section of this report covers examination of several ``beehives`` and presents information concerning the possible cause or causes of growth.

A technique has been developed for slow cycle strain fatigue testing using specimens of thin-wall tubing of the type under consideration for use as super heat fuel cladding. Data on type 304 stainless steel have been obtained under radiation and in the absence of radiation. The strain cycle fatigue life of this alloy at 1200 to 1300 deg F is decreased three fold by the presence of a neutron flux of 1.5 x 10/sup 14/ > 1 Mev. Out-of-reactor data have been obtained on Inconel at 1300 deg F. (auth)

Reactor Fuel Elements--temperature peaking in SNAP due to surrounding rods; systems for nuclear auxiliary power (SNAP)--reactor fuel temperataure peaking due to surrounding fuel rods; temeprature--calculations of peaking of, in SNAP fuel due to sourround fuel rods.

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A program was initiated on the spectrophotometric study of aqueous solution chemistry. The goal is operation at temperatures up to at least 330 deg C and at pressures up to 200 atm, and to near the critical point if this appears to be feasible. A spectrometer capable of operation under these extreme conditions is being designed. (W.L.H.)

This report provides the status of production test control tube loadings in reactor process tubes containing significant amounts of SS materials.

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