slide presentation

Physics scoping studies of a 300-Mw(e) PuO/sub 2/-UO/sub 2/-fueled fast- breeder reactor are reported. Physics design parameters that effect fuel costs, full conservation, and reactor safety were evaluated for use in the selection of parameters for a reference design. The total breeding ratio varied from 1.1 to 1.5 in the range of parameters corsidered. Plutonium core loading ranged from 500 to 1500 kg. Doubling time was found to be reduced by high-density fuel and low steel content. A compromise figure on fuel-rod range of sizes (about 100 mils) yields a 5 operating reactivity and a small, negative sodium temperature coefficient. (J.R.D.)

The thermal expansions of commercial uranium dioxide specimens were measured up to the melting point. The linear expansion of dense, normal grain size UO/sub 2/ follows closely the equationi L = L/sub 0/(1 + 6.0 x 10/sup -6/t + 2.0 x 10/sup -9/t/sup 1.7 x 10/sup -12/t/sup 3/). An anomalous expansion was noted in the temperature range 1000 to 1500 deg C. Above 2500 deg C the rapid vaporization and crystal growth of UO/sub 2/ necessitate the application of heating techniques which provide rapid heating and quenching in order to obtain reliable data. The use of solar and arcmelting furnaces for this type of measurement is described. (auth)

disintegrating a corrosion scale from stainless steel systems without attacking the stainless steel. In this investigation particular sample of stainless steel corrosion scale appeared to be disintegrated most effectively in a solution of the trisodium salt of N-hydroxyethylethylenediaminetriacetic acid (Versenol) and ammonium acetate. In a subsequent test with this solution in a REED dynamic Ioop a considerable disintegration of scale from the loop was observed. in this test after a period of approximately 60 hours, the iron content in the solution was found to be 10 g per liter; thus indicating that the mixture of Versenol and ammonium acetate can be used to disintegrate partially the corrosion scale from a stainless steel system. (auth)

Weight increases during the oxidation of irradiated foils of pure copper were greater than for unirraaiated specimens. Enhanced reactivity appeared to be strongest in the thin-film region up to about 5 mu g/cm/sub 2/. Oxide film (Cu/ sub 2/O) thickness for both irradiated and unirradiated specimens was approximately 1200 A. Radiation did not affect the reduction of Cu/sub 2/O during the induction period (period in which the reduction proceeds very slowly or not at all). In later stages of the reduction process, a serious lack of reproducibility was observed. Radiation effects on films of Cu/sub 2/O formed by prior oxidation of the copper substrate decreased the kinetics of secondary oxidation. The secondary oxidation curve exhibited a large gap at the point of interrnption for irradiation. The development of an automatic recording microbalance of high sensitivity and a furnace for studies in reactor radiation fields is reported. Measurements were made of the electrode potentials of irradiated (5.5 x 10/sup 19/ neutrons cm/sup -2/) copper, aluminum, magnesium, and zirconium. Cell potentials were found to be dominated by the oxide films formed on the electrode surfaces. The results indicate that radiation does affect the local anode reaction potential. No significant difference between the …

Methods for analyzing chemical impurities in sodium metal samples are presented. Chemical analysis was used to determine impurities in calcium, carbon, chromium, iron, lithium, nickel, oxygen, potassium, and zirconium. Spectrographic analysis was used to determine other impurities. Sodium samples obtained from experimental apparatus were analyzed by these methods. (auth)

The general prospects of several radioisotopes are reviewed; the special properties of U/sup 232/ and Th/sup 228/ are poi nted out; and ionium (Th/sup 230/ ) and protactinium target materials are discussed from the sthndpoint of availability and chemical separations processes required for the preparation of U/ sup 232/ and Th/sup 228/. Outlines are given for potential schem es for the separation of U/sup 232/ and Th/sup 228/ from uranium milling pr ocess waste streams and from the irradiation products of Th/sup 230/--Th/sup 232/ mixtures. The high heat generating rates of these potent alpha emitters make them especially suitable for primary consideration as heat sources for small thermoelectric generators. The exceptionally high alpha activity suggests their use in special neutron sources as Ra-Be sources, and they may have sufficiently high neutron generating rates to be in contention with some of the smaller research reactors and experimental neutron producers. (B.O.G.)

Progress is reported on the development of a cyclone which will remove particles larger than 8 microns. A method is proposed for a more efficient separation of particles by increasing the number of size separation filters in the sampling train. Preliminary tests with submicron polystyrene particles are being conducted. Numerous methods have been tried for counting the particles in a water droplet of the polystyrene aerosol. The criteria for a satisfactory method of counting particles are discussed. A proposed method to accomplish this is to use carbon-14 labeled polystyrene hydrosols. (For preceding period see ARF-3127-6.) (B.O.G.)

Early work on the antiproton, particularly that part which led to the first paper on the subject, is described. Conclusions that can be drawn purely from the existence of the antiproton are discussed. (W.D.M.)

Procedures for the dissolution of U-Mo alloy fuels to prepare feed solutions for low-acid (Redox) type solvent extraction processing are presented. U-Mo alloys can be dissolved in boiling ferric nitrate--ritric acid solutions to higher terminal urarium concentrations and lower terminal acidities without precipitation of uranyl molybdate than in nitric acid alone. Anion resin exchange studies indicate the presence of negatively charged iron-molybdenum complex ions in the solutions. The U-Mo alloys also dissolve more rapidly in ferric nitrate--nitnic acid solutions than in nitric acid alone; dissolution rate data are given. Curves delineating free acid, uranium, snd iron (III) concentrations within which solutions stable towards solids formation can be prepared from U-3 wt.% Mo and U-10 wt.% Mo alloys are presented. Stability during prolonged storage of uranium--molybdenum-ferric nitrate--nitric acid solutions is discussed. Data on the oxidation of plutonium in these solutions and on further neutralization of the solutions are presented. Fission product decontamination and product recovery obtained in solvent extraction studies simulating the Redox process are discussed. (auth)

Preliminary experiments were conducted to evaluate the use of the nuclear reactions Li/sup 6/ (n, alpha )H/sup 3/ and O/sup 16/(H/sup 3/,n) F/sup 18/ to determine the thickness of oxide films on metals. Sheets of thin paper and of aluminum, imbedded in powdered LiF, were irradiated with pile neutrons at a flux of 6 x 10/sup 11/ n/cm/sup 2//sec and counted with an end-window proportional counter. A saturation activity of 1.87 hr F/sup 18/ of 150 dis/min per microgram of oxygen was observed in the paper, but radioactivity due to impurities masked F/sup 18/ in the aluminum. It is concluded that a 1 A (0.01 mu gm/cm/sup 2/) oxide film thickness may be measured by a neutron irradiation at a flux of 10/sup 14/ n/cm/sup 2//sec but chemical separation of induced radioactivity from the bulk metal is essential. (auth)

The solid solubility limits of the miscibility gap in the U-Nb phase diagram were determined between 800 and 1000 deg C by analysis of diffusion couples using an electron microbeam probe technique. The feasibility of using xray-diffraction methods to measure residual stresses in U and Zr (Zircaloy-2) was investigated. The magnetic properties of hematite ( alpha -Fe/sub 2/O/sub 3/ ) were investigated. The crystal structures of Y/sub 5/Si/sub 3/ and Y/sub 5/Ge/ sub 3 are being investigated. The structural relationships in the pseudo binary system ZrFe/sub 2/ -ZrC r/sub 2/ were studied. (W.L.H.)

The operation of the freeze-seal type sodium pump requires a shaft freeze-seal capable of retaining sodium. A prototype two-coolant-region freeze seal for application on HNPF sodium pumps was designed and constructed. It was tested under environmental conditions to determine its operating characteristics and sodium retaining capabilities. (auth)

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 experimental program was initiated io determine the extent of fast neutron and gamma ray streaming through the SRL Mark II control and safety rods and to evaluate the adequacy of the shielding provided in these control and safety rods. The methods and procedures used to evaluate these problems are routine and proven for determining gamma-ray and fast neutron dosages using radiation sensitive films and gold foils. The final experimental results indicated that no excessive streaming of either gamma rays or fast neutrons is present above or around the SHE Mark II control and safety rods. The analytical attenuation methods used to calculate the fast neutron and gamma-ray streaming dose rates gave results that compared favorably with the experimental data. Even ihough the agreement was favorable, it cannot be concluded that these analyical methods would be equally valid for other annular geometries. Additional experimental work will be necessary in order to establish the validity for performing similar analysis, but the favorable agreement encourages the use of such methods until other methods are determined. (auth)

Revised HRT flowsheets are presented. These revisions cover such items as relocation of freezer units on the lines, corrections to the numbering of lines, valves or instruments, and the addition of a few lines in the service areas. The waste and vent system flowsheet was redrawn as two sheets. (C.J.G.)

The prompt neutron lifetime of the SRE was measured by both the oscillation and random noise techniques. Measurement by use of the oscillation technique gave a prompt neutron lifetime of (5 25 plus or minus 0 35) x 10/sup - 4/ sec for a calculated beta of 7 x 10/sup -3/. The measured noise response indicated a lifetime of (5.25 plus or minus 0.7) x 10/sup -4/ sec. Both measured values are in agreement with the calculated value of 5 x 10/sup -4/ sec. Four experiments utilizing the noise analysis technique were performed to determine the prompt neutron lifetime of the KEWB. All four experiments gave results which agreed within 3%, For an estimated beta of 8 x 10/sup -3/, the measured value obtained was (7.8 plus or minus 0.3) x 10/sup -5/ sec. This is in reasonable agreement with both the energy independert calculated value of 6.6 x 10/sup -5/ see and the value of 6.2 x 10/sup -5/ sec obtained from the experimental inhour equation The oscillation technique has been found to be better suited for lifetime determinations in reactors where the prompt neutron break frequency is less than 5 cps. Reactor noise analysis is more suitable for …

Tests were made to determine the head loss coefficient through the inlet plenum of the Pathfinder reactor and to determine the now distribution among the fuel element nozzles for various operating conditions--with all three pumps operating at the same flow rate and with any combination of only two pumps operating at the same flow rate. A quarter-scale wooden model was used for the tests. Air was used as the fluld. The loss coefficient was determined to be 1.8 plus or minus 0.3. The velocities of flow through the fuel element nozzles were determined to be within plus or minus 5 per cent of average flow when all pumps are operating and within plus or minus 10 per cent of average flow when only two pumps are operating. (auth)

Chemical control of the SM-2 was evaluated both as a partial substitute for burnable poison in the fuel element meat and as a means of improving plant performance. Based on a review of existing information, boric acid was chosen as the reference soluble poison. It was shown that 60% of the burnable B/sup 10/ in the fuel element matrix could be replaced by soluble B/sup 10/ in the coolant without impairing plant stability during load transients. The feasibility of improving power distribution and reducing the number of control rods by supplementing the burnable poison with chemical control was also demonstrated. A preliminary design of an injection and removal system was prepared for the SM-2. (auth)

>The materials problems associated with the present sodium graphite reactor system have generally been approached by using existing knowledge and data to meet the proposed operating conditions. This discussion reviews the general reactor concept and the specific materials used for the major reactor components: (1) shielding materials; (2) core materials; and (3) sodium cooling system materials. In each case, the materials problems and the materials used to minimize or eliminate these problems are described. Economical nuclear power is currently dependent on the flow of improved materials for high temperature use in high radiation fields. Rapid progress is being made in this respect. (auth)

A parameter study was made of some of the nuclear characteristics the HRT would have if the core and blanket were operated at different temperatures. The power density in the fuel solution at the inner surface of the core tank was found to be affected very little by the temperature distribution. However, the thermal flux at the core-tank wall increased when the blanket temperature was reduced (a consequence of the reduced critical concentration). (auth)

The properties of zirconium alloyed with aluminum tin, and molybdenum were investigated. Using reactorgrade zirconium sponge, 11 zirconium-base alloys were double arc-melted and cast into 6-in.-diam. ingots weighing 35 lb each. By such standard hot working procedures as extruding and rolling, the ingots were converted to 1/8-in.-thick strips. The extruded and rolled products were used for a variety of evaluation studies which included corrosion thermal conductivity, tensile, and creep tests. The alloys demonstrated short-time elevated temperature strength properties equal to or greater than type-304 stainless steel. Their corrosion resistance in sodium, at 1000 deg F, compares favorable with that of unalloyed zirconium. The creep resistance and the thermal conductivity were found to be less than those for type-304 stainless steel, but adequate for nuclear reactor application. (auth)

The ''absolute'' containment of ions in the DCX magnetic mirror field resulting from the cylindrical symmetry of the field is discussed. The regions of confine;, ment in space and momentum are plotted for 300-kev deuterons. (auth)

Two designs of sodium cold traps for the HNPF have been subjected to full scale tests, Performance features that were investigated include oxide removal efficiency, oxide capacity, pressure drop characteristics, economizer effectiveness, and temperature profiles, Results indicate that both designs should perform satisfactorily in the Hallam plant, (auth)