An experiment was designed utilizing a charge-exchange reaction to study the decay and interactions of the neutral K mesons produced. The experiment produced no events that could be interpreted as either the decay or interaction of neutral K mesons. The nature of the experiment and the possible explanations of this unexpected result are presented in this paper.

Figures have and can be generated that indicate a higher charge-discharge rate if required before 105-N will be comparable with existing reactors. Also, these figures show an apparent operating cost incentive to increase the charge-discharge rates proposed for 105-N. Although these figures may be true by themselves, other figures developed from the same information and stated on a basis that affords a true comparison, show that the proposed rates for 105-N are compatible with those in existing reactors. However, the accomplishments of existing reactors should be considered as a guide only and not as Criteria since the design basis has already been established for Project CAI-816. An average charge-discharge rate has been proposed for 105-N that is compatible with the two main ground rules of the Project. Namely, the capital cost limitation and the plant factor. This rate of 8 tubes/hr. is one that appears to be reasonable from the charge-discharge design aspects and there is a good possibility that it can be increased with operational experience.

From preface: This report covers Phase I of the authorized work, consisting of determination of the specific reactor concept and size for the most economic reactor, preparation of a reference design of the most promising concept and estimated capital and generating costs, evaluation of the variations in cost as a function of unit size, and evaluation of the reductions in ultimate costs for a power station having multiple units of the same type.

"This study is one of four initiated in 1958 by the Division of Reactor Development, U.S. Atomic Energy Commission, to assess the feasibility of power generation through the use of certain types of nuclear reactors. These studies are available from the Office of Technical Services, U.S. Department of Commerce, Washington 25, D.C."

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Presented at the 1958 Winter Meeting of the American Nuclear Society, December 9, 1958. Experimental and theoretical work on the selflimiting response of reactors to step and ramp insertions of reactivity is discussed along with the general characteristics of self-limiting power bursts. The static characteristics of the cores investigated are presented and the techniques of measurement are discussed. Data from step and ramp tests are presented and compared with the predictions of a theoretical formulation of reactor self- shutdown in terms of energy release. The discussion includes an evaluation of some postulated shutdown mechanisms in the light of experimental results. Some results of detailed calculations of shutdown effects due to specific mechanisms which are believed to contribute significantly to reactor self-shutdown are presented. (auth)

The national laboratory concept, laboratory objectives, the staff, research facilities. research activities, and administration are discussed in general terms and a financial statement is given. Fairly detailed accounts are given for the research programs in the fields of physics, accelerator development, instrumentation, applied mathematics, chemistry, nuclear engineering, biology, and medicine. (W.D.M.)

Given the moments of a function it is often desirable to obtain a numerical approximation of the function. A method for estimating the solutions of the Stieltjes and Hamburger moment problems is developed. Calculations indicate that the method is both feasible and accurate. (auth)

Relatively long time cycles are required for the standard nitric acid dissolution of enriched uranium using safe batch size techniques for criticality control. Upwards of forty eight hours are required for the total dissolution of solid of I and E enriched slugs in nitric acid. The dissolution capacity may be increased by: (1) using a dissolver critically safe by virtue of its geometry or, (2)increasing the rate of uranium dissolution. Dilute agua regia -- a mixture of hydrochloric and nitric acids -- is known to dissolve uranium much more rapidly than nitric acid. This reports present a flowsheet for the rapid dissolution of uranium slugs in dilute agua regia.

Report documenting the progress of the Army Gas-Cooled Reactor Systems Program to develop a mobile, low-power, nuclear power plant for Military field operation.

From introduction: In general, the objective of this study is to provide for the development of the conceptual design criteria for the most promising type of boiling water reactor and associated complete electrical generating plant consistent with a schedule calling for start of construction in July 1960.

From introduction: This section presents the principal results of the optimization study done to establish the economic basis for selection of the core and fuel design.

From introduction: The objectives of this study are to determine the estimated capital and energy costs of this typical conventional coal-fired installation, consistent with a schedule calling for completion of construction April 1964.

"Centrifugal-casting techniques were investigated as a method of producing hollow cylindrical extrusion billets of aluminum-35 w/o uranium. Among the variables evaluated were melt temperature, mold and pouring-spout configurations, mold speed, and method of pouring. With the equipment employed it was found that the best castings were produced stilizing a pouring temperature of 2400 F, a heavy-walled steel cylinder rotating between 700 to 900 rpm for the mold and bottom-pouring technique employing a retractable pouring spout. Sound, nonporous billets 26 in. long and 5 in. in diameter were produced with a yield after machining of over 75 per cent of the original charge. The major losses occurred in the pouring spout-and-cup assembly. This loss is relatively unaffected by the casting length; and, therefore, coatings of greater length than 26 in. should results in even greater recoveries.

Centrifugal-casting techniques were investigated as a method of producing hollow cylindrical extrusion billets of aluminum-35 wt.% uranium. Among the variables evaluated were melt temperature, mold and pouring-spout configurations, mold speed, and method of pouring. With the equipment employed it was found that the best castings were produced utilizing a pouring temperature of 2400 ction prod- , a heavy-walled steel cylinder rotating between 700 and 900 rpm for the mold and a bottom-pouring technique employing a retractable pouring spout. Sound, nonporous billets 26 in. long and 5 in. in diameter were produced with a yield after machining of over 75% of the original charge. The major losses occurred in the pouring spout-and-cup asserably. This loss is relatively unaffected by the casting length; and, therefore, castings of greater length than 26 in. should result in even greater recoveries. (auth)

Several decisions regarding basic project philosophy must be made in order to proceed with scope design and the preparation of equipment procurement specifcations. The purpose of this document is to present as much pertinent data as possible to allow the project representatives to become familiar with the problems involved. A meeting of Representatives is planned for the near future after receipt of project authorization to discuss the scope of this project and its relationship to CG-775. Emergency flow requirements of the K reactors for planned future power levels is approximately 32,000 gpm within 68 sec. A detailed study of the existing high-pressure cross-tie line reveals that a duplicate cross-tie line and five low lift pump operation would be required to provide this flow. The existing emergency generation capacity is not adequate to supply five low lift pumps and all other necessary emergency electrical loads. A possible solution to adequate emergency flows is to connect the proposed steam turbine pump directly to the risers and to consider the turbine pump as the last ditch system. If it is determined that this does not meet the criteria of separate systems, then an alternate solution must be found.

From major objective: The over-all aim of the study has been to obtain a thorough geologic picture of the Chattanooga shale in the area under consideration, with emphasis upon stratigraphic relationships and associated special problems from which, in turn, implications may be obtained as to the origin and age of the formation and possibly the source of the uranium. The study has not been aimed primarily at exploration for uranium and its economic aspects.

Production of Pu from separations plants and output of unfabricated Pu exceeded commitments. Purex plant set a new record high for U processed. Production and shipments of UO{sub 3} met schedules. Purex solvent extraction battery performed below normal, probably because of poor solvent quality. NaOH addition to Redox coating removal waste is being reduced. A 3fold improvement in Recuplex product Al impurity was achieved by means of a specific gravity difference > 0.15 between dilute aqueous feed and extractant. Sintered, high-silica crucibles are being tested in RMA production line in Finished Products Operation. Scope design of a fission product shipping cask was completed; powder temperature should be below 440 F for 1 MCi cerium-144 + impurities. Feasibility of using one outside Purex canyon entrance (stairwell opening) for relief damper opening was tested and found to be insufficient. A drawing of the 6-inch continuous centrifuge being evaluated as a vacuum drum filter on RMA button line was reviewed. Casks were designed for the NPR project. (DLC)

A concentration cell electrode was installed in a fritted glass surface and was used to measure the surface KCl concentration while water was being transpired through the surface into a mixed tank of 0.1 M KCl. The results from the first Fluorox run made with crude UF/sub 4/ showed that 85.3% of the theoretical amount of UF/sub 6/ was accounted for, with 17.9% being collected in cold traps and the remainder being consumed in various side reactions. Installation was completed of apparatus to study the electrolytic removal of nickel on an engineering scale from concentrated HRT fuel solution. An extremely low yield stress of 0.02 lb/sq ft was measured for a thoria-urania slurry containing 2.5 wt.% Al/sub 2/O/sub 3/ and spherical particles of 1.7 micron mean diameter. Preliminary data on the rate of uranium sorption on Dowex 21K from sulfate solutions were correlated with a simple spherical diffusion model. Laboratory scale studies to determine the effect of reflux time, HNO/sub 3/ concentration, and air sparge on chloride removal from APPR-type Darex dissolver product showed that an air sparge is definitely beneficial in chloride removal. Preliminary cyclic cleaning tests showed that boiling 25% caustic solution was adequate for cleaning type F …

The stability of unclad zirconium hydride was investigated as a reactor moderator in carbon dioxide coolant. Two ranges of hydride were tested, an alpha- plus deltaphase hydride (ZrH/sub 0.9 plus or minus 0.2/) and a delta-phase hydride (ZrH/sub 1.5/). The delta-phase hydride exhibited excellent dimensional stability, but suffered as great as 20% hydrogen loss at 1200 deg F, together with severe corrosion. The alphaplus delta-phase hydride swelled and warped in a temperature gradient but did not lose much hydrogen and did not corrode severely even at 1200 deg F. The results indicate that zirconium hydride could not be used as unclad moderator in a carbon dioxide atmosphere in the temperature range desired for the Maritime Gas Cooled Reactor. (auth)

From abstract: "The toxicological properties of beryllium and compounds of beryllium are briefly reviewed, together with the historical development of the recommendations for maximum permissible beryllium air concentrations. The application of the enclosure technique presently in use at this laboratory for the control of beryllium hazards is described... Monitoring procedures used to evaluate the performance of enclosures are discussed."

The toxicological properties of beryllium and compounds of beryllium are briefly reviewed, together with the historical developmert of the recommendations for maximum permissible beryllium air concentrations. The application of the enclosure technique for the control of beryllium hazards is described. Emphasis is placed on the design objectives of partial and total enclosures and the related function of auxiliary components. Monitoring procedures used to evaluate the performance of enclosures are discussed. (auth)

An evaluation of the temperature excursion and its possible consequences arising from loss of coolant from the Plutonium Recycle Test Reactor (PRTR) was made for four different postulated ruptures in the primary heavy water coolant system. As a basis for the evaluation, a series of computations was made. These were based on incremental heat and mass balances for sections of Zircaloy-clad UO/ sub 2/and Pu- Al fuel elements. Solutions to each problem defined by the postulated break size and its location were defined by finitedifference approximatioms performed by an IBM 653 machine digital computer. The four postulated ruptures were: (1) a complete parting of the 14-in.-diameter outlet pipe near the upper ring header so that coolant would be lost from both broken ends; (2) a rapture equivalent to a 14-in.-diameter hole in the primaryloop piping adjacent to the upper ring header; (1) a complete parting of a 1 3/4-in. upper jumper; and (4) a complete parting of a 1 3/4-in. bottom jumper. The Pu-Al elements represent the most critical component; melting of these elements would begin about 219 seconds after the rapture occurred if emergency backup light water coolant were not available to the system. It was found that the …

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