Details are given of a monitor for the continuous determination of deuterium in helium. Excess oxygen is added to the gas stream and the oxygen concentration determined before and after passage of the gas through a deuterium-oxygen recombination unit. Oxygen concentrations are measured with galvanic cells. The accuracy is better and +- 10 percent in the range 200 - 5000 v.p.m. deuterium. operating instructions for the Lucas Heights equipment are given.

This program was prepared as a pilot program for a larger computer and handles symmetrical reactors with core, side reflector and end reflectors, using 10 radial and 10 axial mesh regions. The output consists of the effective multiplication constant, the two flux distributions, and the fission source distribution. The extrapolated Liebmann process is used for the inner iterations.

A survey of the aqueous coordination chemistry of beryllium is given. The possible use of coordination chemistry in the separation of beryllium from fission products is discussed, outlining methods for separation processes.

The isotopes of lutecium were used to measure the neutron temperatures in two collimated beans of neutrons emerging from HIFAR>

Details are given of an apparatus used to dissolve irradiated ceramic, metallic, and carbide fuel specimens, to dilute the dissolver solutions accurate to a known volume, and to take aliquots with a specially adapted automatic burette. Procedures for its use are given.

Methods are described for the determination of traces of Be, Th, and U in concentrated sulfuric-phosphoric acid mixtures. When the Be concentration is sufficiently high, the chrome azurol S spectrophotometric method may be applied directly, and a small correction made for phosphate interference. At lower concentrations Be should be first separated by an acetylacetone extraction. Th must be separated from sulfate and phosphate before the thoronol spectrophotometric method can be used. This is achieved by precipitating Th as the fluoride, using Y carrier. U may be determined spectrophotometrically with arsonazo after separating Be, Th, suIfate, phosphate, and other impurities by anion-exchange from hydrochloric acid solution. In an alternative procedure, U is reduced to the tetravalent state and precipitated with Th as the fluoride, again using Y carrier. The determination is then completed by a-c polarography.

The dissolution of dense beryllia was studied in a variety of reagents. The dissolution rates were too slow to be of practical importance except those for hydrofluoric acid, sulfuric acid, and mixtures of sulfuric and phosphoric acids. The reaction with hydrofluoric acid was studied in more detail in an attempt to throw some light on the dissolution process. The initial dissolution rate appeared to be proportional to the square of the acid concentration between 0 and 20M. An apparent activation energy of 12 Kcal/mole BeO was obtained from the temperature coefficient of the dissolution.

The 25% Cr, 20% Ni type stainless steel has been proposed for use in the Australian High Temperature Gas Cooled Reactor in core structures, and in hot gas ducting. Thus a knowledge of the compatibility of this steel with high pressure carbon dioxide was required. Rates and mechanisms of corrosion were investigated for machined, vapour blasted, and etched pretreated samples of this steel, exposed to carbon dioxide up to 3,000 hours in the temperature range 650 degrees C at gas pressures from 3 p.s.i.g. to 280 p.s.i.g. Oxide film flaking as apparent at all temperatures investigated but was only severe for pre-ground samples at 710 degrees C and above, and for pre-vapour blasted samples at 760 degrees C and above. However, severe intergranular penetration was observed in pre-etched samples on exposure to carbon dioxide at 650 degrees C and above. Pressure of the gas appeared to have no systematic effect on the corrosion rate, at least in the temperature range investigated. The maximum useful temperature for which the steel could be used would be limited by the amount of oxide flaking permissible. In reactor gas circuits where a small amount of scale flaking could be tolerated, the steel is satisfactory …

A method for comparison and evaluation of thermoelastic stresses is given for a range of fuel element shapes based on parameters available from the initial study of a reactor system. The shapes studied, in descending order of stress level are circular rods, concentric tubes, flat plates, and a matrix of circular holes.

As part of a general study of the co-ordination chemistry of beryllium, the beryllium salicylate complexes have been investigated by ion exchange procedures. The evidence indicates that a neutral 1 : 1 and an anionic 1 : 2 chelate exist in solution under appropriate conditions, and their stability constants have been determined by ion-exchange methos. The values of the stability constants were found to be [beta]1 = 4.97 x 10 (12), and [beta]2 = 2.63 x 10(22).

Stable solutions of basic beryllium nitrate can be formed with beryllium concentrations up to 9M, and an NO3/Be mole ratio as low as 1:1. The efficiency of basic beryllium nitrate as an agent for salting out uranium into a tributylphosphate/kerosene solvent has been compared with that of other salts. It appears possible to separate uranium and thorium from beryllium and fission products using a modified Thorex process in which beryllium nitrate replaces aluminum nitrate.

The matrix of a dilute homogeneous H.T.G.C. reactor fuel employing metallic beryllium as a moderator can be selectively dissolved by a caustic soda solution containing salicylate ion. At least 99 percent of the uranium and thorium can be recovered as insoluble solids, but in the case of irradiated material the uranium loss might be higher. Some decontamination of the resulting beryllium solution from fission products and Pa233 can also be obtained. A tentative chemical flowsheet is proposed on the basis of the results obtained.

Concentric tubular reactor fuel element geometries to give equal coolant outlet temperatures are presented. Oscillations from tube to tube in thickness and temperatures generally occur but it is possible to eliminate them by choice of the centre element. This may be a fuel rod or a non-heat—producing rod with or without a surrounding annulus of fuel. The geometries and temperatures are dependent on the voidage and on a non-dimensional parameter equivalent to a Biot number based on the channel equivalent diameter.

The concept of a dispersion fuel element is discussed with particular reference to irradiation damage. The application of this concept to the A.A.E.C. H.T.G.C. reactor system is outlined and the limitations imposed by irradiation damage considerations are discussed. The maximum desirable heavy metal - beryllium ratio (i.e. U+Th:Be) for the various systems under consideration should be about 1:55 for the system (U,Th)Be13 in Be, 1:13 for the system (U,Th)O2 in Be, and 1:8 for the system (U,Th)O2 in BeO. The disadvantages of keeping uranium and thorium in separate particles are discussed and it is suggested that to minimize irradiation damage effects, the fuel particles should consist of solid solutions of the uranium and thorium compounds.

The high flux research reactor HIFAR contains ten tons of heavy water which acts as moderator and primary coolant. Over an eighteen months period regular microbiological examinations have been carried out on samples of heavy water taken from various parts of the circuit. The heavy water circuit provides an interesting opportunity for the study of microorganisms because of the high isotopic purity (greater than 99.6 per cent.), and the high chemical purity of the heavy water in the reactor. Furthermore, during its passage through the reactor core the water and suspended bacteria are subjected to intense irradiation, the neutron flux being approximately 10 14 neutrons cm-2 sec-1. The presence of bacteria in the heavy water circuit has been demonstrated and experimental results and methods used are discussed. Some evidence is presented to show that the ion—exchange resin bed contributes nutrients to support bacterial growth.

Breakaway corrosion of Be in moist CO2 can be avoided if the Be is fabricated using preoxidized powder. The powder is preoxidized by heating in dry O/sub 2/. Preoxidation of Be powder was measured as a function of temperature and time of heating in O/sub 2/. The subsequent reactions of the preoxidized powder in moist CO/sub 2/ at 700 deg C were studied and the effect of increasing amounts of added oxide was measured. A model is proposed to explain the inhibition of corrosion by added oxide. (auth)

Measurements were made of the capacity of pre-treated Victorian brown coal for removing trace quantities of Sr2+ and Cs+ at pH 8.0 and 9.5 in the presence of various concentrations of Ca2+. At 20 p.p.m. Ca2+ breakthrough was immediate.

Some preliminary measurements have been made of the rate of dissolution of beryllium when refluxed with ammonium fluoride solution. The rate of dissolution exhibits pseudo first order dependence on the "free" fluoride concentration defined as the concentration of total fluoride less that assumed to be present as BeF4. The solubility of uranium metal in ammonium fluoride - beryllium fluoride solutions increases with free fluoride concentration, where the uranium is present in solutions as U(1V). Thorium metal is attacked only slightly under similar conditions. Beryllium may be selectively leached from UBe13 by ammonium fluoride solutions if the fine A-/Be mole ratio is greater than 4:1. The solubility of uranium under these conditions suggests that this is not a satisfactory solvent for mixtures of these alloys with Be when the U/Be ratio is small, but that it could be used successfully as a decladding agent for an oxide fuel clad in beryllium meta.

Comparison of the adsorption characteristics of the desiccants silica gel, alumina, and molecular sieves has shown that molecular sieves have by far the greatest capacity of the desiccants at the low partial pressures considered. Equilibrium data in the form of isotherms were established over the range of variables expected in the coolant circuit of a proposed Australian H.T.G.C. reactor. The mass transfer from the gas phase to molecular sieves is such that no correlation could be attempted for the adsorption zone height; the height proved to be too small.

The indirect addressing feature of the I.B.M.1620 computer has been used to overcome the difficulty of cross—referencing separately assembled subroutines within a computer programme. A relocatable assembler has been devised which permits the separate assembly and testing of such subroutines. The concept of a "next subroutine" has been introduced, and its applications to interpretive systems such as Fortran shown.

A description is given of a method suitable for the automatic solution of certain optimum design problems on a digital computer for cases where the number of constraints imposed on the design is not greater than the number of design variables. The problem is transformed to one requiring the minimization or maximization of an unconstrained function, for which a gradient method is used.

A comparison of methods of capital cost estimation used for nuclear fuel reprocessing plants shows that, because of the special nature and complexity of such plants, cost estimation methods for conventional chemical plants involving the use of cost factors are not applicable and will give low estimates. Cost factors which are available from other countries where reprocessing plants are installed should be used with caution since those factors apply only for the particular design philosophy used and pertain to industrial conditions which are different in this county. Capital cost estimation methods involving direct take-offs from detailed design drawings are necessary to obtain reliable estimates. The methods of estimating operating costs for nuclear reprocessing and conventional chemical plants are similar.

A method is given for constructing the complete neutron spectrum for a well-moderated thermal reactor such a HIFAR, from the total effective flux, the temperature of the Maxwellian, the epithermal spectral index and the total integrated fission flux. A sample calculation is also included.

The reactor transfer function of HIFAR has been measured at low power and compared with the calculated response determine the prompt neutron lifetime. Lifetime values in the range 600 to 800 microseconds, depending on the position of absorbers withing the core and reflector , were obtained with an estimated error of less than 100 microseconds.