Static overpressure and dynamic pressure versus time over surfaces processing different physical properties were measured on two tower shots, 6 and 12. On Shot 12, three surfaces were provided: the natural desert, a water surface consisting of a flooded area, and an asphalt surface. On Shot 6, desert and asphalt areas only were available. There were 123 channels of instrumentation installed for Shot 12, and 24 for Shot 6. From the data, a system of wave-form classification was devised for overpressure and dynamic-pressure- versus-time measurements. Incorporation of this system into data analysios indicates that it is possible for an ideal peak pressure to be identified with a nonideal wave form. Introducing both variables, wave form and peak pressure, into analyses reduces ambiguioties associoated with comparing results of different nuclear tests. The data show the effect of the nature of the surface upon airblast phenomena from a nuclear explosion. The effects of surface conditions upon shock phenomena are made more understandable by a review of temperature computatioons, using shock wave parameters in addition to an analysis based upon the arrioval time of the thermal pulse. A phenomenological discussion of precursor formation is presented, and comparisons are made using data from all …

Three new americium activities (Am{sup 238}?, Am{sup 243}, and Am{sup 244?}, the latter two formed by n,{gamma} reactions) are described and some additional information is given on previously reported americium isotopes.

Neutron density at the end of the slug is increased appreciably and therefore the heat production in the end of the slug is also increased. The end result is increased heat flow into the cap with correspondingly higher temperatures. The combined results and conclusions of documents CP-1580, CP-1989, and CP-2871 afford a means of calculating the aluminum end cap temperatures with consideration being give to the above mentioned {open_quotes}Wilkins effect,{close_quotes} i.e., that the heat production near the end of a slug with an aluminum end cap is greater than the heat production at the middle (axially) of a slug. The maximum temperature in the aluminum end cap is calculated to be about 278{degrees}F. for the {open_quotes}G{close_quotes} Pile. This temperature occurs at the center of the maximum performance process tube.

Pile temperature control is normally maintained by using four horizontal rods, two ``long`` rods extending to the far side of the pile and two ``short`` rods whose tips are on the near side. (1) This procedure prevents cycling of ``hot spots`` about the pile, but does not achieve symmetrical horizontal temperature distribution. Prior to pile operation the cadmium-containing cans were removed from the near ends of the two ``long`` control rods, numbers 5 and 11, at C pile, so that when these rods were in the ``full in`` position their poisoning effect outside the flattened region at the near side of the pile would be small. By using these ``half rods`` in conjunction with ``short`` rods it was expected that a nearly symmetrical horizontal temperature distribution could be maintained. A comparison of near-to- far temperature distribution at C-pile with other piles indicates that gains in production of 4% during early operation of a pile and 3% during later equilibrium operation can be realized by using the half-rods. On the basis of these observations ``half-rods`` are to be incorporated in the K-piles currently under construction. It is recommended that the ``long`` control rods in existing piles be replaced by rods altered …

The new KW reactor was placed in operation on January 4, 1955, and had been running at relatively low power levels for only 17 hours when it was shut down because of a process tube water leak which appeared to be associated with a slug rupture. After several days of unrewarding effort to remove the slugs and tube by customary methods, it developed that considerable melting of the tube and slugs had taken place. It was then evident that removal of the stuck mass and repairs to the damaged tube channel would require unusual measures that were certain to extend the reactor outage for several weeks. This report documents the work and findings of the Committee which investigated the KW reactor incident. Its content represents unanimous agreement among the three Committee members.

Studies in the 321 Building dissolver during December 1953 and January 1954, were successful in developing a laboratory-proved mercury-catalyzed dissolving flowsheet into a suitable plant procedure. However, this flowsheet was not adapted for Redox plant operation because of uncertainty about the possible presence of hydrogen above the lower explosive limit in the off-gases. Subsequent laboratory work has resulted in a better understanding of the hydrogen evolution, and has resulted in developing low hydrogen evolution flowsheets. When one of these flowsheets is selected for further work, it will be tested in the 321 Building dissolver with non-irradiated slugs to provide information for scaling-up the single-slug laboratory data to a plant-scale operation. It is the purpose of this memorandum to outline the factors considered to be pertinent to the 321 Building investigation, to be used as a guide in making preparations for the runs to be performed.

An enriched I & E fuel element failed during irradiation under Production Test IP-109-AC in December, 1957. Radiometallurgical examination of this slug was requested by the IPD Process Analysis Operation. This report presents the results of the examination work.

Equipment required to provide production capacity of 525 tons per month of I & E fuel elements ``O,`` ``C,`` and ``K`` sizes in the 313 Building is specified in this document.

The report is divided into: Fissionable materials (2000 program): studies related to production reactors, studies related to separations plants; reactor development (4000 program): Studies related to plutonium recycle program.

In the development of fuel elements for the NPR, one potentially serious fuel element problem -- high temperature uranium swelling -- has not been experimentally investigated. A series of experiments has been proposed in which uranium fuel rod with different amounts of Zircaloy-2 cladding will be irradiated to high exposure at temperatures equivalent to those expected in an NPR. These experiments should show the importance of high temperature uranium swelling as a limiting factor in NPR fuel element behavior. To obtain sample rod temperature of 250 to 300{degree}C on the surface and 500 to 650{degree}C at the center, the rods will be irradiated in aluminum capsules loaded in standard reactor process tubes. The high temperatures will be obtained by restricting the heat flow from the uranium sample to the coolant. The purpose of this test is to determine the validity of the heat transfer calculations used in predicting the temperature drops within the capsule by irradiating one capsule at known flux conditions and measuring the temperature attained by the uranium rod sample. The data obtained from this test will be used in determining the irradiation conditions required for the full scale uranium swelling tests.

Purpose of this report is to determine optimum tank size from evaluating the quantities of principal construction materials with prevailing unit costs for various tank sizes. The materials were concrete in-place, reinforcing steel in-place, wood framework, 3/8 in. C steel plate liner in-place, earthwork excavation and backfill (engineering, overhead, piping, condenser, vapor manifold costs not included). Costs of optimum tank are distributed as follows: dome 25%, walls 28%, foundation 6%, floor 2%, steel liner 20%, earthwork 19%. For a given tank capacity, there is a definite optimum tank size; as the capacity increases, the diameter increases, and the height increases but at a lesser rate. Each diameter has an optimum height, which is that height at which unit cost of storage space is minimum for a given tank diameter. Optimum unit cost is $0.136/gallons for diameters 75--130 ft; for diameters<75 ft, the optimum unit increases. Tank forms 241-S, 241-SX, and 241-A were used in this study; storage cost of the analyzed tank was $0.121 compared to $0.136/gallon for 241-SX and 241-A, and $0.152/gallon for 241-S. Assumed unit costs for concrete and steel plate tank liner were 10% less than those of 241-SX and 241-A, causing the lower unit costs. Tanks …

Side-to-side flux distributions calculated for ``rods out`` operation are compared to the normal equilibrium distributions. The results are highly sensitive to enrichment and flattening unbalance. Near side tube power limits are reached at total pile powers as low as 39%, and as high as 85% of the normal equilibrium. power. Calculations indicate that, with the current loading patterns, B, D, DR, F, H, and KW, theoretically could make rods out scram recoveries without exceeding tube power limits, whereas C and KE could not. Longitudinal flux profiles corresponding to possible control rod and mixing piece configurations are compared to the ideal cosine distribution. Peak fluxes for a given column power vary from 86 to 111 % of a cosine maximum. The position of the maximum flux varies from 115 cm (5.1 slugs) downstream to 131 cm (5.8 slugs) upstream of the center line of the slug column. The effects of non-uniform (front-to-rear) enrichment and flattening charges are not considered.

The performance of I & E fuel elements under the original tests{sup 1} was such that they are now being charged at C Reactor on a production basis. Data obtained from the original test, however, were not sufficient to indicate long term corrosion effects of the use of these elements, nor to accurately define their operating characteristics. It is the intent of this test, therefore, to provide for long term corrosion monitoring in tubes charged with and without water-mixing pieces; and to obtain the required operating data to refine the operating characteristics of this fuel upon which power limits are based.

The present liquid third safety is considered undesirable for the piles at their present power levels. Any increase in power level and graphite temperature makes the need for an alternative third safety increasingly imperative. The use of the third safety is envisioned under conditions where the supply of cooling water to an operating pile is interrupted and concurrently the safety rods full to enter the pile because of slight shifting of the top shield or larger shifting of the winches, relative to the pile proper. Under these conditions, undesirable properties of a liquid third safety are present and are discussed in this memorandum.

The use of organic cooling media for nuclear reactors operating at high power levels predicates the use of a coolant which will not react violently with metallic uranium in the event of a fuel element failure. This report describes the testing, and subsequent examination, of two pieces of irradiated uranium which were immersed in monoisopropylbiphenyl (MIPB) at high temperatures and pressures for periods of time up to twenty-five days. The uranium samples had different irradiation histories and cooling times. Similar experiments had been performed with unirradiated uranium by the Corrosion and Coatings Operation, and it was wished to determine whether irradiated uranium would react with MIPB in a different manner.

The purpose of this work has been to determine the production rate of tritium in fringe Li-Al alloy columns with the degree of precision necessary for economic analyses of such reactor loadings. These results are provided for use in such an analysis. This experiment indicates the production rate of tritium in the outermost fringe tubes to be T = 0.0216 M{sub E} = 0.175 M{sub t} where T = grams of tritium per full length (67 pieces) charge of Li-Al alloy material; M{sub E} = MWD/adjacent ton of E metal; M{sub t} = MWD/adjacent tube of E metal. The above values should apply for fringe loads utilizing greater or smaller quantities of E metal; that is, for isotope production loadings which are over or under-compensated from a reactivity standpoint. In the actual test load it was calculated that one gram of tritium and 13.5 grams of Pu were made for each 21.3 grams of U-235 burned up. During the same time interval the displaced uranium loading would have generated 24.3 grams of Pu and burned up 29.9 grams of U-235. The factor which seems to limit the accuracy with which these data can be interpreted is the ratio of the …

This monthly document details activities of the Hanford Operations Office during the month of November 1959. (FI)

This report gives results of a study of the effect of changes in axial power distribution on rupture potential. Possible interrelationships between this effect and the effects of other reactor variables were investigated.

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This report consists of an outline for a program for testing the air-supplied face masks currently being used as protection against P-10 atmospheres at HAPO. The test procedures which follow are intended to determine circumstances under which the mask ceases to offer protection to the wearer. To keep the scope of the test program to a minimum and also to facilitate interpretation of the findings in the light of P-10 atmospheres, the test conditions chosen were those which would be most unfavorable for the mask from a protection standpoint.

Preliminary observations on tritium production fro extracted Z slugs in the H-10 load indicated yields which were lower than those calculated by approximately twenty percent. The calculated conversion efficiency for loading is 0.835. Results of a measurement of the conversion efficiency for three separate tubes are reported and found to be lower than 0.835 by approximately twenty percent, or very nearly the same as the discrepancy in tritium production. The values measured are 0.67, 0.61, and 0.60 for the three tubes respectively. Details of 25 burnup calculations are also presented with a measured depletion factor of 0.584 grams of 25 destroyed. Reasons for the large discrepancy in conversion efficiency are not known at the present time. However, some possibilities are discussed.

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Research on polonium chemistry and physics is being continued on a reduced scale. Those problems which have direct bearing on the preparation of the metal and those which will lead to a better understanding of any phenomenon encountered are being continued. Data from the calorimetric measurements of polonium on Calorimeters 39 and 46 were treated by the method of least squares to give a grand mean half life of polonium equal to 138.3993 {+-} 0.0093 days (1-13). New apparatus has been constructed for the measurement of the vapor pressure of polonium and its compounds in the low and the high temperature ranges. Technical difficulties have delayed the actual collection of data (1-22). The constant current magnet to be used for the measurement of the Hall effect of polonium has been tested and found satisfactory. A device utilizing nuclear magnetic resonance absorption for determining field strength has been constructed on a preliminary basis (1-24). A total of 208 lines in the ultraviolet region between 3200 A and 1900 A have been measured with an error of less than {+-}0.02 A (1-28). Studies on the X-ray and spark spectra of polonium and its compounds have been conducted. A preliminary study by X-ray …

The SNAP Experimental Reactor (SER) is a small power reactor that is to be built and operated in the original SNAP-II critical facility. The reactor is described, and the hazards previously evaluated for the SNAP II critical faciity are reviewed.