The problems encountered while developing loading methods for the ORNL water-cooled research reactors are discussed. The methods used to program the fuel loading for the LITR and ORR are described as well as the experimental methods that produced the data necessary for the solution of these problems. (auth)

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Research and development progress is reported on plastics, Ranger x-ray sensor, radioelements, isotope separation and purification, and reactor fuels and materials development. (M.C.G.)

ABS>Measures are outlined that are taken to control occupational radiation exposure incurred in the routine operation of a reactor and in the conduct of activities associated with its operation. Those measures taken to minimize the frequency and effects of minor radiation and contamination incidents are also discussed. (C.H.)

The appearance of a large mass 12 peak in pure helium afterglows prompted further investigations, from which it is believed that the maas 12 peak is carbon from a CO impurity, probably a result of outgassing of the walls of the discharge chamber, Mixtures of oxygen with helium and neon have been examined in an attempt to measure the O/sup +/ to O/sub 2/ charge transfer rate. Afterglows in hydrogen, water vapor, and nitrogen-- hydrogen mixtures have also been studied. (auth)

Methods are presented for determining the electric power requirements of a given communications mission in terms of mission and orbit parameters. Analyses were made of possible applications of available and projected space auxiliary power units in these satellites. The satellitc as a communication node is discussed. Example calculations are given. (M.C.G.)

The water plants and associated steam back-up facilities which supply coolant to Hanford`s eight production reactors are large and vital facilities. Their reliability is essential to the safety and continuity of reactor operation. Their output and efficiency are important to reactor productivity and to the unit cost of plutonium and other irradiation products. In the fiscal year just ended, the cost of operating the reactor water and steam plants was $16,000,000, which represented a major fraction (37%) of the total irradiation cost for the same period. Substantial improvements in water plant capacity and operating economy over an extended period are reflected in this operating cost figure. The purpose of this report is to highlight the capacity increases and cost reductions achieved during the five fiscal years 1958 through 1962.

Results of studies concerning the asymmetric/symmetric ratio for fission induced by neutrons of energies corresponding to various fission resonances are presented. These studies were carried out primarily with monoenergetic neutrons at 1.8, 2.3, and 4.3 ev. The fission ratios in the low-energy resonance region was found to change from resonance to resonance, and the ratios appear to fall into symmetry classes each of which has a characteristic peak-to-valley mass yield ratio. It appears that a spin dependence is involved in the mass distribution as predicted by the Bohr model. The exact nature of this dependence is not clearly understood. (J.R.D.)

Spectroscopic observations were made of the flow field behind a cylindrical hypersonic wave front resulting from an imploding electrical dischange into an initially low pressure quiescert gas. Sub-microsecond resolved spectra were obtained by use of a quartz prism monochromatorphotomultiplier detector system with photographic recording of an oscilloscope displayed output signal. Electrically produced plasma rings were established in detonabie as well as non-reactive gases. Strong indications were observed that exothermic reactions, as evidenced by H/sub 2/O emission, proceed in H/sub 2/ + 1/2 O/sub 2/ reactant mixtures within less than one microsecond under certain initial low pressure conditions. A reaction threshold at 1 mm Hg pressure was found for the particular reaction vessel used. Timeintegrated spectrometer photos support the conclusions reached with the transient spectra analysis equipment. (auth)

The diffusion rate of O in anion-deficient zirconia, ZrO/sub 1.994/, was determined by the interface migration of stoichiometric oxide and is represented by the equation D = 0.055 exp (--33,400 surface proces 3100/RT). A comparison was made with other processes that occur in the metal and the oxide. Excellent agreement was noted between activation energies of O diffusion in ZrO/sub 1.944/ and those for parabolic or cubic oxidation in both air and water. It appears that O diffusion in the oxide is rate-controlling during oxidation of the metal. The corrosion and oxidation behavior of Zr and some alloys are discussed in terms of the oxide defect structure and the electric conductivity behavior in the oxide. A speculative mechanism for corrosion transition to linear rates was suggested on the basis of preferential oxidation of a grain boundary metallic phase. The nature of the phase and of its formation and elimination are discussed. (auth)

Effective energy cutoffs have been calculated on an IBM7090 computer for Cd, Gd, Sm, and B filters as functions of filter geometry, the ratio of Maxwellian to epithermal flux (assumed to be 1/E), the lower energy limit of the 1/E flux, the energy corresponding to the Maxwellian most probable (modal) velocity and filter thickness. The geometrical configurations were spherical (which on the assumptions madc is equivalent to a beam flux case), cylindrical and slab. By the use of two or three different filters (Cd and Gd and perhaps Sm) it should be possible to detect resonances in the thermal to cutoff energy regions, in addition to measuring resonance integrals and thermial cross sections of unknown nuclides. (auth)

The postulate that the average number of lattice displacements is directly proportional to the available energy is carried one step further; it is assumed that damage to steel (particularly in regard to brittle fracture) is proportional to the number of lattice vacancies that occur. The model, although crude, permits estimation of the relative damage resulting from differences in neutron spectra. The results can be used as a rough method of correcting damage data for the effect of the neutron-energy spectrum. Radiation damage calculations for steel, relative to those for a fission spectrum, were made for neutron spectra that result from fission neutrons penetrating water or graphite. The results were plotted as a function of effective distance from the fission source. From this plot it is possible to make a conservative estimate of the correction factor to apply to damage data obtained with different neutron spectra. (auth)

In May, 1961, it waa determined no renewed nuclear reaction in the SL-1 pressure vessel was possible as long as water was excluded. It was therefore possible to proceed with the final phase of the SL-1 recovery. This work consisted of moving the pressure vessel and core to the Hot Shop at the north end of the National Reactor Testing Station, dissecting and analyzing the reactor and its components, cutting up and burying the reactor building, and decontaminating the rest of the SL-1 area. These things were accomplished by the General Electric Company between May 1961 and July 1962. It was determined that the central control rod was bound in its shroud at a position corresponding to 20- inch withdrawal. Analysis of the pertinent data showed that the amount of reactivity associated with this rod position, inserted at a rate compatihie with manual withdrawal of the rod, can explain the significant evidence which was coliected. No other means of withdrawing the rod was found to be in accordance with the evidence. It was found that the relatively low yield (130 Mw-sec) nuclear excursion produced a water hammer with pressures up to 10,000 psi, which, in turn, caused the pressure vessel …

The temperature coefficient of the Reactivity Measurement Facility was found to be 49 plus or minus 1 mu k/ deg C (1 mu k = 10/sup -6/ DELTA k/k) in the range 15.4 to l7.8 deg C. The change in the net reactivity of a standard sample was --0.48 plus or minus 0.02, --0.66 plus or minus 0.03, and --0.78 plus or minus 0.02 mu k/ deg C in three measuring positions. These low values generally make temperature corrections insignificant. The above results are compared with previous determined values. This information developed in the RMF should be generally applicable to flux-trap-type reactors such as the Advanced Reactivity Measurement Facility (ARMF) and ARMF-ll, now under construction. RMF was dismantled in April 1962. (auth)

The relative cross section for the production of protons from water by electron bombardment was measured by a modulated crossed beam technique over an energy range from threshold to 500 ev. The appearance potential for protons was measured by comparison with the ionization potential of hydrogen and found to be 19.4 ev. This result is in agreement with earlier results. The overall curve is smooth with the exception of a slight shoulder at about 45 ev. This shoulder may indicate the presence of more than one process for the formation of protons. (auth)

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Studies indicated that the use of soluble poisons as a primary criticality control offers economic and other advantages in that it permits the factors of vessel size and shape and solution concentrations to be dictated by considerations other than those of criticality. It is believed that soluble poison criticality control can be made as reliable as other methods of coaditional control if the application is preceded by adequate development work and is monitored by multiple. independent safeguards. The studies included multigroup machine calculations of the required content of poisons in solutions of fissile and fertile material, a compilation of data on the detection, stability, decontamination, and costs of soluble poisons, and an assessment of the possible effects of a nuclear excursion. (auth)

A seven- to twenty-fold volume reduction can be obtained from fluidized bed calcination of aqueous aluminum nitrate wastes, depending on the operating conditions employed and their effect on the intra-panticle porosity and absolute density of the calcined alumina. Among the calcining variables, only the bed temperature and the fuel aluminum concentration had a significant effect on the intra-particle porosity of alumina generated during studies conducted primarily in a two-foot-square fluidized bed calciner. A quantitative correlation of the effect of these variables is presented. Alumina with an intra-particle porosity as low as five per cent can be generated by employing a suitable combination of low bed temperature and dilute aluminum feed concentration. Feed sodium concentration and product alpha alumina content were found to have minor effect on intra-particle porosity. Results also show that an inverse relationship exists between the nitrate content of the calcine and the calcination temperature. (auth)

One of the early candidate fuel elements for N Reactor use was the coextruded, Zircaloy-2-clad seven-rod cluster. As part of the program of evaluating the seven-rod cluster geometry, three-rod cluster fuel elements, two and three feet long, were irradiated. These long cluster fuel elements were irradiated to determine the distortion (or sag) which might occur at the center of the unsupported length during irradiation. Two three-rod clusters made up of 0.630 inch diameter rods, containing natural uranium cores were irradiated in KER Loop 3. The rods of one cluster were three feet long; the rods of the other were two feet long. The three-feet long rods were supported at their ends and at their midlengths, the two-feet long rods only at their ends. During the irradiation, the maximum core temperature was 435 C. The fuel elements were discharged from the loop after they had reached an exposure of 1400 MWD/T. Following the discharge, the fuel elements were visually examined in the KE view pit. No sag was observed in any of the rods. The test demonstrated that two- and three-feet long rod cluster fuel elements can be irradiated without appreciable sag occurring in the rods.

One of the early candidate fuel elements for the N Reactor was the seven-rod cluster fuel element. An objective of the program to determine the suitability of the seven-rod cluster fuel element for N Reactor use was to evaluate the irradiation performance of coextruded, Zircaloy-2-clad, seven-rod cluster fuel elements over a range of exposures from low exposures to high exposures. This report describes the irradiation testing of an enriched seven-rod cluster fuel element which was irradiated to 520 MWD/r.

The General Electric Company, as prime contractor to the AEC at Hanford, is proposing to modify the lattice characteristics of the 8 3/8-inch lattice reactors for the purposes of improving the conversion ratio of these reactors. The proposed overbore modification of the reactors would remove the existing aluminum process tubes, enlarge the diameters of the graphite channels by about one-half inch, insert smooth-bore Zircaloy-2 process tubes and refuel the reactor with larger size, self-supported fuel elements. The overbore fuel will remain the internally-and-externally-cooled cylindrical type, but the weight per foot will be about twice that of the present fuel element. The removal of the inlet and outlet piping connections which would be required in the overboring process will permit the replacement of the existing fittings with ones of improved design. Furthermore, new orifices and venturis which are compatible with the hydraulic characteristics of the overbore tube and fuel geometry and the pumping system will be installed. No basic changes are proposed in the pumping system though the reactor flaw rate may be increased 5--10 percent by changes in hydraulic characteristics depending on the water plant flow capacity.

The effect of modifying the dislocation structure by room-temperature prestraining on the subsequent yielding behavior at 77 deg K is being studied. Results on tantalum single crystals indicated that a considerable decrease in the early yield stress for a given strain is effected by prestraining at room temperature. Twinning was observed in the tantalum at 77 deg K and normal rates of strain. It is shown that the potential barrler to dislocation motion in crystalline solids can be measured in relatively pure bcc transition metals such as Nb, Ta, Mo, and W. Work is being carried out to extend the method of differential calorimetry to measurements of stored-energy-release spectra in deformed bcc metals. Measurements for copper are discussed. (M.C.G.)

Run 25 was the final period of power operational of the HRT. The reactor was operated for periods of 62, 8, 52, and 80 hours at 5 Mw with no outward indication of fuel and core and blanket average temperatures of 270 and 230 deg C, respectively. The uranium concentration in the was 1.7 to 2.0 g U/kg D/sub 2/O. Longer periods of operation were prevented by mechanical difficulties, notably with the fuel feed pump. While the reactor was subcritical after the last of the above runs, the upper patch in the core tank wall became dislodged, allowing greater core-to-blanket mixing. The resultant blanket uranium concentration was 2.9 g U/kg D/sub 2/O. The reactor was subsequently operated at April 28, 1961. The experiment was operated at high temperature for a total of 10,866 hours. The system was critical for a total of 8,841 hours and produced 16,295 Mwhours of power. The fuel, heavy water, and some corrosion specimens were recovered, and the reactor was stored in an assembled state. (auth)