The work described in this report was done to determine the path of decomposition of plutonium (IV) oxalate and to determine the factors affecting the reactivity of the oxide with the hydrogen fluoride.

An experiment designed to evaluate the in-pile performance of the U-Mg fuel material when irradiated to high burnups has been completed. Twelve specimens of the fuel material which contained uranium particles that packed 50 volume per cent, (91.5 weight per cent), uranium in a magnesium matrix were canned in Zircaloy cans and irradiated in the Materials Testing Reactor to 0.1 (1000 MWD/T), 0.3 (5000 MWD/T), 1.0 (10000 MWD/T) and 2.0 20000 MWD/T) per cent burnup of the total uranium atoms; more exactly, 1 MWD/T = 1.16 x 10⁻⁴ per cent burnup of the total uranium atoms. Irradiation of the twelve capsules began on August 1, 1954. The burnup figures used in this report are calculated values assuming a conversion ratio for the capsules of 1.0. Because of the lack of confirmed experimental burnup data for exposures of this magnitude, there is a possible error in the calculated values of about 20 per cent at 2.0 per cent burnup. However, recent results based on chemical analysis for cesium indicate that the calculated values of burnup agree quite closely for the higher exposures. Burnup estimates based on the results of the chemical analysis will be published when they become available. Six of …

The attractiveness of a helium cooled, heavy water moderated, natural uranium central station power plant has been investigated. A fuel element has been devised which allows the D20 to be kept at a low pressure while the exit gas temperature is high. A preliminary cost analysis indicates that, using currently available materials, competitive nuclear power in foreign countries is possible.

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It is the purpose of the Fuel Element Technical Manual to Provide a single document describing the fabrication processes used in the manufacture of the fuel element as well as the technical bases for these processes. The manual will be instrumental in the indoctrination of personnel new to the field and will provide a single data reference for all personnel involved in the design or manufacture of the fuel element. The material contained in this manual was assembled by members of the Engineering Department and the Manufacturing Department at the Hanford Atomic Products Operation between the dates October, 1955 and June, 1956. Arrangement of the manual. The manual is divided into six parts: Part I--introduction; Part II--technical bases; Part III--process; Part IV--plant and equipment; Part V--process control and improvement; and VI--safety.

The operating conditions to which the special Mark VIII quatrefoils will be exposed during the proposed piloting program have been compared with the conditions required to cause burnout, using an established method of calculating these conditions. The results of this comparison permit the following conclusions to be drawn: (1) With normal flow of coolant through the special elements the heat flux to be encountered in the R-8 cycle (1400 MW) will be 70% or that required to cause burnout (30% margin from burnout). (2) With a reduction of coolant flow to 82% of normal through one tube of a special element, burnout of that fuel column is possible in the R-8 cycle. (3) In the R-6 cycle (1280 MW), the margin from burnout in the special Mark VIII quatrefoils is 42% with full flow and 20% with the above reduced coolant flow. A similar comparison of operating conditions predicted for the L-3 cycle (full Mark VIII charge) shows that, even at the highest power level (1250 MW), the margin from burnout is greater than 55% with normal flow and 40% with reduced flow.

``Safe-failure`` is an important criterion for power reactor fuel elements designed to contain core materials which react rapidly with high-temperature coolant. Out-of-pile experiments indicate that, without a ``safe-failure`` fuel element, expensive and possibly catastrophic reactor shutdowns may occur if the fuel element jacket fails during irradiation in high temperature water. That jacket failures occasionally will occur, regardless of the material or fabrication process involved, seems to be generally agreed. Fuel elements involving cluster arrangements, oxide cores, and wafer assemblies with ``bulkheads`` for compartmentalization have been proposed as being relatively safe in event of jacket failure. A fourth possibility is a type of long cartridge fuel element.

A report based on a study about the factors which affect the amount of chemical reaction between water and Zircaloy 2 at high temperatures. Also, experimental programs for the measurement of radiation emissivity, chemical reaction rates, and diffusion rates have been completed.

When acid leach liquors obtained from U ores containing Mo are contacted with anion exchange resin, both U and Mo are adsorbed. The adsorbed Mo is not removed in the norraal U elution process, and when the U-barren resin is recycled, the resin sites already occupied by Mo are unavailable for U adsorption. In succeeding cycles, more and more sites become occupied by Mo, and the resin capacity for U is seriously lowered. This rapid accunmulation of an adsorbed species is known as resin poisoning.'' The many disadvantages of such a situation are self-evident. (auth)

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S>The Zircex Process for the recovery of zirconiuun from zirconium clad fuel elements is presented. Various types of apparatus were investigated for desublimation and collection of the solid zirconiuun tetrachloride. Of primary importance is the determination of particle size and distribution of the ZrCl/sub 4/ condensed for, in general, the larger and more uniform the particle size, the easier the collection. (W.L.H.)

Feasibility studies of a high flux, solid fuel research reactor are presented. By means of a parameter study, a reactor consisting of a cylindricai fuel annulus submerged in heavy water is described. The thermal neutron flux peaks in the heavy water adjacent to the annulus and is a maximum in the region surrounded by the fuel annulus.. while the minimum thermal flux occurs in the annulus. The fast flux has the opposite shape. Calculations indicate that practical peaking factors, ratios of maximum thermal flux in the heavy water to average thermal flux in the fuel annulus, as high as eight can be obtained. A typical reactor in which the maximum thermal and fast fluxes are greater than 10/ sup 15/n/cm/sup 2/-sec is also described. (auth)

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This report is a study of the feasibility of using a fused salt fuel reactor in a central station electric generating plant.

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The first group of Pu--Al MTR fuel elements fubricated were rejected for excessive dimensions and blistering. The fuel element constitution and fubrication was reviewed, and a new Pu--Al ratio and revised rolling techniques were suggested. (D.E.B.)

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