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

One of the uncertainties concerning the use of uranium-10 wt.% molybdenum in the ORNL Fast Burst Reactor is the thermal-cycling behavior of this alloy. Accordingly, an experimental program was undertaken to determine whether transformation or distortion of gamma-phase uranium--10 wt.% molybdenum can occur during simulated fast-burstreactor thermal cycling, and, should transformation occur, to establish the thermal cycling behavior of partially transformed uranium- - 10 wt.% molybdenum. Specimens were prepared by vacuum casting pins in coated- graphite molds, homogenizing the castings at 1650 deg F for 24 hr, and centerless grinding and cutting to 0.158-in. diameter by 1.5 in. long. The pins were sealed in evacuated Vycor tubes, heated rapidly both above and below the gamma- transformation temperature. and then cooled slowly to simulate reactor thermal cycling. Measurements of the time required to initiate transformation by conventional isothermal methods were employed to insure that the alloy material behaved as indicated in the literature. Observation of changes in appearance, dimensions, density, resistivity, and metallographic structure were used to obtain the desired information. It was found that thermal cycling did not cause growth or distortion of either gamma-phase or partially transformed uranium-- 10 wt.% molybdenum pins. Transformation of gamma phase to alpha + …

Valve stem freeze seals for high-temperature service in advanced sodium- cooled reactor systems were studied. An experimental model, suitable for use with a 6-in. size valve, operated satisfactorily under a variety of conditions. The freeze seal region was cooled by natural convection to ambient atmosphere; cooling by both circumferential and longitudinal finned sections was experimentally studied. The operating conditions included sodium bulk temperatures up to 1300 tained F, sodium pressures up to 75 psig, and ambient temperatures as high as 150 tained F. Anticonvection rings were pcsitioned in the sodium-filled annulus between stem and stemguide, and the effects of their presence was studied. Predictions of temperature profiles along the stem, using several different analytical methods, were compared with experimental results. (auth)

This document details activities of the Fuels Preparation Department during the month of June 1960. (FI)

Test quantities of thoria (aporoximately 3800 lb of thorium oxide and 3200 lb of mixed thorium-uranium oxide) were prepared during FY-1960 for members of the Reactor Experimental Engineering Division (REED). For preparation of thorium oxide, the calcination temperature was varied from 650 to 1800 deg C. The surface area of the fired oxide ranged from 0.5 to 25 m/sup 2//g and the mean particle size ranged from 1 to 8 microns. For preparation of mixed thorium-- uranium oxide, the calcination temperature was varied from 1050 to 1225 deg C. The mixed oxide contained either 0.5, 3.0, or 8.0 wt.% of uranium as requested with a mean particle size of 1.5 microns. The over-all losses of thorium and uranium were 19.4 and 18.3%, re spectively. Four 1400 lb batches of thorium nitrate were dissolved to prepare solutions containing 300 to 350 g/liter of thorium for solvent extraction studies in the Chemical Technolcgy Division and 2100 lb of dry, recovered waste thorium oxide was supplied to the ORNL Fuel Cycle Program. (auth)

The effects of ternary additions of up to 3 wt.% Sn or Zr to an Al-35 wt.% U extrusion alloy were evaluated on the basis of casting characteristics, UAl/sub 3/ retention, extrusion behavior, mechanical properties, and corrosion resistance. Both additions increased the fluidity of the alloy, and both promoted retention of UAl/sub 3/. The best fluidity was obtained by a 2 wt.% Sn addition, while Zr was the more effective stabilizer of UAl/sub 3/. The retention of UAl/sub 3/ decreased the extrusion pressure needed for fabrication and caused a corresponding decrease in tensile and creep-rupture properties. Reductions in strength were most noticeable at elevated temperatures. The 1000- hr stress-rupture strength of the binary alloy at 200 deg C (8300 psi) was approximately 25 and 11% higher, respectively, than the alloys containing 3 wt.% tin (6200 psi and 3 wt.% zirconium (7400 psi). The additions either slightly improved or had no effect upon the resistance of the Al-35 wt.% alloy in 150 deg C demineralized water. (auth)

The work carried out over a 2 1/2-yr period on the scavenging of radioactive particles which might be released by the reactor system of a nuclear- powered ship is summarized. Two types of dispersions were considered: aerosols and hydrosols. Radioactive aerosols were scavenged by heterogeneous coagulation with solid and liquid aerosols produced within the radioactive aerosol cloud. Liquid or highly hygroscopic particles, which can be classified as solid particles with liquld films on their surfaces, were found to be the most effective scavengers. A system of fine water spray and hydrolysis products of silicon tetrafluoride was found to be suitable for field application. Scavenging of radioactive cations, anions, and colloids of corrosion and fission products was studied in substitute ocean water, natural ocean water, and natural harbor water. A scavenging system composed of KMnO/sub 4/ and ferrous salts successfully removed most of the radioisotopes. Fe(OH)/sub 3/--MnO/sub 2/ hydrate adsorbed and absorbed radioactive species, thus transferring them from a liquid to a solid phase. Addition of Floc 111 to the system improved sedimentation. The KMnO/sub 4/-FeSO/sub 4/-Floc 111 system was found to bs suitable for field application. (auth)

An investigation was made to determine the stress field in the salt dome and the stress concentrations on the surface of the small (12-ft diameter) sphere. For the most part, the apparatus and techniques used in the investigation are new and have not been described in other reports. Therefore the theory, concepts, apparatus, and some of the lists made to determine the reproducibility of the apparatus are described briefly. (W.L.H.)

The thermocouple train is modified by the substitution of one eighteen-inch enriched and four twenty-four inch natural uranium tubular elements for the two thirty-six inch enriched tubular elements used on the original thermocouple train. The operating limits for the loading have been changed because of the change in charge power. This supplement also authorizes the addition of fuel elements upstream of the thermocouple train if the thermocouple element and heater elements do not provide enough heat to operate the loop at the desired temperature.

Experiments showed that 30% tributyl phosphate will not extract acid- deficient species of uranyl nitrate. Flooding throughputs for the Mark I stacked- clone contactor ranged from 600 cc/min organic at zero aqueous to 60 cc/min organic at 1950 cc aqueous. A large electronic vibrator of 5000-lb thrust was found somewhat inferior to pneumatic vibration for compacting oxide fuels into stainless-steel tubes. Tests were started on the use of fixed-bed CuO oxidizers for removing hydrogen contamination from helium gas streams. None of the variables studied within this period effected an increase in particle size in the denitration of thorium nitrate to produce ThO/sub 2/. The rate of uranyl sulfate loading on nitrate equilibrated Dowex 21K appears to be essentially independent of the loading solution sulfate concentration. The nitric acid concentrations corresponding to maximum UO/sub 2/-ThO/sub 2/ pellet dissolution rate were 15.5 M for Thorex solution and 13 M for the adjusted Darex solution. Two additional Semicontinuous Sulfex declad and Thorex core dissolutions of prototype Consolidated Edison fuel assemblies were made to complete the series of runs. The effective area of cylindrical UO/sub 2/ pellets dissolving in nitric acid was estimated from experimental rate measurements as a function of the fraction dissolved. …

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Since startup in January 1945 through December 1959, 4 {times} 10{sup 9} gallons of radioactive liquid wastes have been discharged to cribs and trenches at HAPO by the Chemical Processing Department facilities in the 200 Areas. These wastes contained approximately 2.5 {times} 10{sup 6} curies of beta emitters. The Scavenged Waste Recovery Program was completed in 1957 and Redox Plant process changes were made during the latter part of 1958. These changes resulted in significant reduction in the amount of radioactive materials that have been discharged to the ground in subsequent years, from a maximum of 8.2 {times} 10{sup 5} curies in 1955 to 9 {times} 10{sup 3} curies in 1959. Although these large amounts of radioactive materials have been discharged to the ground, periodic waste report inventories include no reduction due to radioactive decay. The estimated depletion by radioactive decay is the basis of this report.

A test was conducted to determine the radiation level of several scram shaft assemblies after their removal from the reactor vessel during Core 1 Seed 1 refueling. Data on radiation levels near the assemblies 75 days after shutdown are given. (J.R.D.)

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Wood, coke, and coconut-shell charcoals were evaluated for fluorine entrapment. The coconut-shell charcoal produced the smallest amount of solid and liquid reaction products. Efficient removal of fluorine was accomplished by the coconut-shell charcoal in a 5-in.-diameter reactor with a feed containing 25% fluorine at flow rates from 100 to 400 scfh and reactor-wall temperatures of 1200 to 1800 deg F. (C.J.G.)

A compilation is presented of results obtained to date from a number of reactor physics calculations for the molten salt reactor experiment (MSRE). Included are one-dimensional multigroup and two-dimensional twogroup calculations of critical mass, flux, and power density distributions; gamma heating in the core can, reactor vessel, and core support grid; drain tank criticality; and an estimate of the beta , gamma , and delayed neutron dose rates due to fission products in the fuel contained in the pump bowl. For a cylindrical core 54 in. in diameter and 66 in. high, graphite-mcderated with 8 vol% fuel salt, the calculated critical loading is 0.76 mole% uranium (93.3% U/sup 235/), which is equivalent to a critical mass of 16 kg. At a reactor power of 10 mw, the peak power density in the core assuming a homogeneous mixure of fuel salt and graphite is 10 watts/cm/sup 3/, the average power density is 4 watts/cm/sup 3/. The computed peak thermal flux is 7.3 x 10/sup 13/ neutrons/cm/sup 2/ sec and the average is 2.5 x l0/sup 13/ neutrons/cm/sup 2/ sec. Gamma heating prcduces a power density of 0.2 watts/cm/sup 3/ in the core wall at the midplane and 0.4 watts/cm/sup 3/ in …

In determining radionuclide sorption by clay minerals, shortcomings in the filtration technique for solid separation and in the contact times selected for testing were noted. Filters were found to have a high affinity for cesium and strontium when these elements were present in tracerlevel concentrations. Sorption equilibrium was not established in 24 hr, and the contact time was extended to 7 days. The affinity of the clay minerals illite, kaolinite, montmorillonite, and vermiculite for selected radionuclides was established after these shortcomings in the testing procedure were corrected. Illite exhibited a very high affinity for Cs/sup 137/ (K/sub d/ =200,000 after 7-day contact); none of the clay minerals demonstrated exceptionally high affinity for strontium (K/ sub d/ = 4000 for kaolinite, which was the highest measured value). The behavior of cobalt and zirconium-niobium was complicated by the colloidal nature of the elements. The uptake of the above radionuclides by a composited sample of Clinch River sediment showed that cesium and strontium behaved in accordance with the mineralogic character of the sediment. The sediment sorbed more cobalt than was accountable by the mineral composition; organic matter interactions with the cobalt are thereby suspected. (auth)

S> Thicknesses of ordinary concrete required to shield the demineralizers and filters in the HFIR primary water system were computed for normal operating conditions and for abnormal conditions such as a meltdown of the fuel within the reactor. About 4 1/2 ft, 3 1/4 ft, and 4 1/4 ft of concrete are required to shield the cation exchange unit, the anion exchange unit, and the filter unit, respectively, to the most stringent of the following radiation levels: (a) 0.75 mr/hr for normal reactor operation or reactor operation with one defective fuel plate; (b) 1 r/hr immediately following the meltdown of 1% of the fuel; and (c) 1 r/hr 24 hours following a total fuel meltdown. Shielding thicknesses may be estimated for other tolerances from

An apparatus to study rare-gas fission-product release from nuclear fuel materials during postirradiation heating was developed. Xenon and krypton fission gases escaping from a small specimen during heating at constant temperature are measured using a continuous radioactivity monitor and charcoal adsorption traps. The rhodium-wound furnace is capable of operation at 1600 deg C. Helium carrier gas is purified by activated alumina, copper, and zirconium traps, and the oxygen and moisture contents of the gas are monitored continuously. The operating procedure and data are presented for a typical heating experiment in which fused uranium dioxide was studied. (auth)

Use of X-8001 Al alloy as cladding for Hanford reactors was initiated because of superior (laboratory) resistance to intergranular corrosion over that of C-64 alloy. However, since severe pitting attack was observed intermittently, an evaluation was carried out on X-8001 alloy fuel element cladding.

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An investigation was made to determine the structural behavior of selected components of the Oak Ridge Research Reactor for increased power level conditions. It was found that a reactor cooling water outlet temperature of 150 deg F will cause severe plastic strain cycling in the aluminum housings for the large test facilities. Increasing the reactor cooling water flow rate of 21,000 gpm will cause plastic deformations in certain reaons of the core box. These latter deformations can be tolerated, but the full implications asscciated with any change in pressure differential must be understood before adopting the above flow rate. (auth)

In an investigation of the detailed temperature structure of the seven- element cluster of cylinders surrounded by a sleeve which comprise the fuel assembly for the EGCR, the radiative interchange of heat between the rods and sleeve was evaluated. A procedure advocated by Hottel was used to determine the view factors for gray enclosures taking into account the multiplicity of reflections, absorptions, and secondary radiations. (auth)

This report for June 1960, from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance: Financial operations; facilities engineering; research; and employee relations.