The successful completion of a program of experiments, including the Aircraft Reactor Experiment (ARE), has demonstrated the high probability of producing militarily useful aircraft nuclear power plants employing reflector-moderated circulating-fuel reactors. Consequently, and accelerated program culminating in operation of the Aircraft Reactor Test (ART) is under way. In order to adhere to the compressed schedule of the accelerated program, it is essential that the Atomic Energy Commission approve the 7500 Area in Oak Ridge as the test site by February15, 1955. This report summarizes the hazards associated with operating the contained 60-Mv reactor of the ART at the proposed Oak Ridge test site.

The method is based on caustic precipitation and centrifugation (which removes the Cs and small amounts of Sr, rare earths, Zr, Nb, and Ru). These are removed in the supernatant and run through a cation exchange column. This separates Zr-NB and Ru. The effluent is precipitated and the Zr-Nb is stored in an asphalt pit. The Ru then may be recovered from the precipitate. The precipitate from the original centrifugation is calcined, pressed and transported to a deep well.

In order to evaluate the behavior of an APPR type absorber rod, an irradiation test program has been established. Approximately 21 more samples are planned for testing under this request. The request proposes testing a full size APPR-1 type control rod in the MTR. The objective of the test is to better evaluate the neutron absorbing material proposed for the APPR-1 control rod.

The absorption in lead of Bremsstrahlung X radiation from a Sr^90 TiO3 pellet in the proximity of Hastelloy "C" was measured. The tenth value layer of the more energetic components of the X-ray continuum was determined to be 1.60 inches.

A method was developed for calculating shield induced gamma radiation escaping through openings in a biological shield. The method was applied to the HRT and the results indicated that the contribution to the dose from induced activity in the HRT shield was around 0.1 r/hr and was insignificant in comparison to to other mechanisms contributing to the escape of gamma rays through shield openings.

Resistance of carbon steel to corrosion in oxygenated high-temperature (250C) water was unexpectedly good at high oxygen concentration. Pertinent literature, critically examined, and toroid experiments indicted that at low oxygen concentration attack did increase with concentration, but as oxygen concentration was sufficiently increased, more protective films were formed on the metal. Some corrosion factors in the application of carbon steel to nuclear reactors systems are discussed.

Catalysts for use in recombining the gases produced by the radiolytic decomposition of water in thorium oxide slurries under neutron irradiation were investigated in out-of-pile tests using stoichiometric mixtures of hydrogen and oxygen. Most favorable results were obtained with a molybdenum oxide catalyst. Satisfactory rate also were attained with palladium and silver oxides. Copper, nickel, vanadium and chromium compounds were less effective.

A promising scheme for the chemical processing of a thorium breeder reactor of the two-region aqueous homogeneous type consists of the following operations: concentration of insoluble fission and corrosion products from the core system into a small volume of fuel solution, combining this slurry with irradiated thorium oxide slurry taken from the blanket, recovery of D2O by evaporation, dissolution of the thorium and uranium in HNO3, and, after a suitable cooling period, recovery of the uranium and thorium by solvent extraction for return to the reactor. The use of a hydroclone and underflow container arrangement for concentrating insoluble fission and corrosion products under simulated reactor conditions has been successfully demonstrated on dynamic loops. Solids concentration factors greater than 103 were demonstrated, and equilibrium solids concentration in the circulating solution less than 1 ppm was attained in these tests. Present data indicate that proper design and operation will minimize solids deposition in the reactor system and that the insoluble impurities can be effectively removed by the hydroclone. An alternate method of processing the slurry removed from the core system by the hydroclone consists of removing the room temperature insolubles by centrifugation, recovering the uranium from the supernatant by peroxide precipitation, thermal …

A capital cost breakdown of the Idaho Chemical Processing Plant, a directly maintained remotely operated plant for processing spent enriched uranium fuel assemblies from reactors, is presented. The capital investment in the plant, including design, construction, training, and preoperational costs, an estimate of the direct costs incurred by the Atomic Energy Commission, and a proportional part of the costs of Central Facilities, including the value of the land and improvements theorem when acquired by the Commision, was $31,105,899. The cost of design and construction was $25,212,231, of which $3,773,357 was expanded on design and inspection.

When it was discovered that two HRT flange bolts of a lot of 16 spares contained serious cracks, a program was launched to (1) determine the cause for the cracking, and (2) find methods for non-destructive testing the remainder of the 672 bolts shipment, a large portion of which had been installed in the HRT. Concurrently, inspection of 8 ferrules removed from an HRT flange revealed hairline cracking in 4 of them. Magnaglo, a magnetic particle inspection method using a fluorescent dye, proved to be the only definitive method for inspecting the bolts. The evidence gathered on the bolts pointed to quench cracking as the cause for the defects. Nothing abnormal was disclosed in regard to the bolt material. The alloy and heat treatment at present prescribed for the HRT bolts and ferrules are considered suitable. However, recommendations are made for plating with zinc, instead of formerly prescribed cadmium, to a thickness of 0.0002 inch, followed by a hydrogen relief treatment and a final bichromate chemical dip.

Short reports to the members of the Nuclear Cross Section Advisory Group from three groups: (1) High voltage group; (2) Fast chopper time-of flight spectrometer; and (3) Electronuclear research division.

Some refined calculations have been made, relative to a proposed delay prior to a dump, to determine the expected D2 concentration in the vent stream from the pressurizer gas bleeds during a dump of the Homogeneous Reactor Test (HRT). These calculations indicate that for vent valves have a Cv of 0.07 (venting time from 2000 psia to D2O saturation pressure of approximately 12 minutes), a delay period is not required since the D2 concentration is well below lower explosive limit. For vent valves having a Cv of 0.3 (venting time approximately 2.4 minutes), the calculation indicate that a delay before venting of approximately two minutes will be required. This is due entirely to the possibility of mass ebullition the D2. Since the pressure drops so quickly, the reactor solution becomes saturated with D2 before appreciable recombination can occur.

The feasibility of determining particle sizes in the submicron range by employing an activation analysis - centrifugation method has been demonstrated. It is believed that this method is now applicable to the analysis of thorium oxide for submicron particles. The same techniques are, in most instance, usable in determining particle sized in other sample materials.

A carrier containing 138.99 liters of solution, uranium concentration 202.04 g/liter with an isotopic concentration of 97.3% U-233, was prepared for shipment. The total uranium was 28,062 +/- 60 g (95% confidence level) and the U-233, 27,305 +/- 66 g (95% confidence level).

At the request of the Alloy Preparation Group, the Nondestructive Test Development Group investigated the feasibility of determining the clad thickness on Mark X MTR Fuel Plates. As the use of induced eddy-currents was considered to be the most promising approach, a prototype instrument and probe coil utilizing this principle was developed to measure clad thickness. The results of the investigation conducted with this instrument indicate that the clad thickness of this type of fuel plats can be measured to withing +- 0.001 in.

The enthalpies and heat capacities of seventeen fluoride mixtures in the liquid state have been determined using Bunsen Ice Calorimeters and copper block calorimeters. The fluoride mixtures were composed of the fluorides of two or more of the following metals: lithium, sodium, potassium, beryllium, zirconium, and uranium. The enthalpies and heat capacities of most of these mixtures were studied in the solid state also. Estimates of the heat of fusion have been made. General empirical equations have been developed which represent the enthalpies and heat capacities of the fluoride mixtures in the liquid and in the solid state.

An estimate has been made of the gamma radiation levels at access holes in the HRT Shielding when the plugs have been removed to service or maintain the reactor. In every case the radiation level at the holes was greater than the maximum permissible exposure rate of 0.3 roentgens per week. The radiation through the holes can be attenuated to some extent by flooding the reactor cell up to the flange to be disconnected. However, shielding would still be required and it is more practical to provide a small additional shield thickness to compensate for the moderate attenuation that could be gained from flooding.

Few data exist for the thermal conductivity and viscosity of gases at very high pressure. The possibility of using gases for heat transfer media at pressures up to 100 atmospheres and above raised the problem of estimating variations in the conductivity and viscosity at high pressure. Generalized plots are presented which are based on the work of Enskog, Eucken, and Hirschfelder et al. Some pertinent data from Hirschfelder et al and from Hilsenrath et al are presented.

A six-inch diameter "O" ring from the HRT Mockup was examined metallographically after 2452 hours exposure to uranyl sulfate solution at 300C. surface defects, except for cold work defects were found only in areas exposed to uranyl sulfates. Defects found were pits, transgranular cracks, surface cracks, and grain attacks.

The results of an examination of "O" ring gaskets remove from the HRT following a discovery that the water in the leak detector system contained 1000 ppm chloride are presented.

This technical report is the second of a series which concerns the separation of ions by solvent extraction with trialkyl phosphine oxides (TOPO). This investigation has consisted in the extraction of various ions from acidic solutions with extractants that are representative of these ore specific phosphine oxides - triphenylphosphine oxide (THPO) and tris-2-ethylhexylphosihine oxides (TEHPO). In general it is observed that: (a) the order of increasing capacity of extraction is THPO > TOPO >TEHPO. (b) No ion is extracted by THPO or TEHPO that is not extracted by TOPO under certain conditions. (c) The effect of hydrogen ion concentration is greater in TEHPO systems than it is in the other two, which indicated greater selectivity of extraction with TEHPO.

An inconel loop circulating fluoride fuel (62 1/2 make [unintelligible] NaF, 12 1/2 make [unintelligible] ZrF4, 25 make [unintelligible] UF4, 92 [unintelligible] enriched) was operated at 1485°F with a temperature difference of about 35°F in the Low Intensity Test Reactor for 645 hr. For 475 hr of this time the reactor was at full power, and fission power generation in the loop was 2.7 kw, with a max length power density of 0.4 kw/cc. The total volume of fuel was 1290 cc (5.o kg [unintelligible] and the the flow through the irradiated section was 8.6 fps (Reynolds number 5500). The loop has been disassembled and has been examined by chemical and metallographic analyses. Ne acceleration of corrosion of decomposition of fuel by irradiation was noted, although deposition of fission-product ruthenium was absorbed. Ne mass transfer of Inconel was formed, and the corrosive [unintelligible] was general and relatively light. The average corrosive generation, in the usual form of subsurface yields, was 0.5 [unintelligible], the maximum penetration was 2 to 3 miles.

Twenty-five hundred psi ring-type flanges, ring gaskets, bolts, and special connectors were tested for adaptability to the aqueous homogenous reactor. High pressure line closures were studied to obtain empirical data pertinent to the selection or design of a connector capable of withstanding sustained thermal cycling and high pressures encountered in the aqueous homogenous reactor. Specialized stress-strain measurement techniques yielded information concerning flange deformation, ring type gaskets, bolts, and special connectors. The results indicated that no totally acceptable connector is currently available. Most promising of the combination of components tested during this period was a 2500 psi ring type flange with an accurately machined octagonal gasket and Grade B-7 bolts.

The operation and design of a high pressure recombination loop for the recombination of H2, D2, and O2 produced by the radiolytic decomposition of water which is used a solvent for fuel in the homogeneous reactors are presented.