摘要:

AbstractA series of experiments to study the amounts and forms of fission products released during simulated reactor accidents is described. These experiments consisted of melting miniature stainless-steel-clad UO2fuel elements in a helium atmosphere in the Oak Ridge Research Reactor and measuring the fission products released. Fission and gamma heat in the reactor raised the temperature of the miniature fuel element sufficiently high to melt the UO2without the use of external heat. In these experiments with UO2, nearly all of the iodine, tellurium, and cesium, and more than half of the strontium, zirconium, ruthenium, barium, and cerium were released from the fuel. Release of the latter group of fission products and uranium from a zone including the fuel and surrounding heat insulators was generally less than 3%. The minimum temperature of this zone during fuel melting was 1000 C. The retention of fission products within the high-temperature zone is considered to be significant since, during an actual reactor accident, temperatures corresponding to those of the high-temperature zone would probably occur only within the immediate area in which the fuel is overheated. Analysis of the observed distributions of deposited fission products yields information about their behavior and form. Ruthenium follows the stainless-steel cladding as it melts and vaporizes. Certain fission products are associated with millimicrometer-size particles of two size groups, one centered around 22 angstroms and the second around 30 angstroms in diameter. Comparisons of the fission-product-release values from in-pile and various out-of-pile experiments indicate that the in-pile releases are greater, probably because of more extreme temperatures.

ISSN:0029-5639    

DOI:10.13182/NSE64-1

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor

摘要:

AbstractA method described in a previous article for synthesizing three-dimensional flux distributions is here extended to reactor kinetics problems and to lifetime studies. The method is outlined and some numerical examples presented. The results of these show the method to be a practical way of solving time dependent reactor problems with a detailed spatial model.

ISSN:0029-5639    

DOI:10.13182/NSE64-A18315

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor

摘要:

AbstractIn the conceptual two-region molten-salt breeder reactor, fissionable U233will be recovered from the blanket as the decay product of Pa233. Since equilibrium concentrations of Pa233would result in appreciable parasitic neutron absorptions, the advantages of thermal breeding could be realized to a greater extent by removing both Pa233and U233from the blanket mixture. Methods for recovering these materials from molten-fluoride mixtures by precipitation as oxides are presented. Small-scale experiments clearly indicated that it is possible to remove protactinium from molten-fluoride solutions by a process that appears to be surface precipitation of protactinium on beryllium oxide or thorium oxide particles. Protactinium was removed from molten mixtures of LiF-BeF2-ThF4(67-18-15 mole %) by the addition of 1 to 2% by weight of solid beryllium oxide or thorium oxide. The removal efficiency was high when the initial concentration of protactinium was either in the range 1 to 2 ppb or 50 to 75 ppm. Uranium was successfully removed from solution in molten fluorides by use of a similar procedure. Approximately 2000 ppm uranium was precipitated from molten LiF-BeF2-ThF4(67-18-15 mole %) by the addition of 3% by weight of beryllium oxide. Comparable results were also obtained using thorium oxide as the precipitant.

ISSN:0029-5639    

DOI:10.13182/NSE64-A18316

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor

摘要:

AbstractThe General Atomic neutron-velocity selector has been used at the electron linear accelerator to study the inelastic scattering by liquid methane and liquid parahydrogen of monoenergetic neutrons with incident energies in the range 0.009 to 0.17 eV. The energy dependence of the total cross sections and the neutron spectra produced by specimens of these materials have also been measured. The inelastic scattering of slow neutrons (<0.010 eV) at 90°by liquid parahydrogen appears to be smaller than expected on the basis of the measured total cross section and the angular dependence calculated by Sarma. Perhaps this is related to the fact that the total cross section is larger than for freely rotating molecules, indicating the possible existence of some hindrance to molecular motion. The slowing-down power,σnE0/E, a quantitative measure of the neutron-moderating ability, is evaluated from the measured inelastic neutron-scattering data and compared for various neutron energies for the two liquids. A consideration of the various data leads to the conclusion (1) that solid methane is better than liquid parahydrogen for production of very“cold”neutrons (E0<0.007 eV), and (2) that parahydrogen is superior to liquid methane for production of cold neutrons withE0<0.005 eV.

ISSN:0029-5639    

DOI:10.13182/NSE64-A18317

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor

摘要:

AbstractIrradiated UO2graphite fuel samples in which most of the fission products had recoiled into the graphite matrix were heated after irradiation, and then leached with nitric acid. The leach-ability of non-gaseous fission products was influenced by fission product concentration, by irradiation temperature, and largely by temperature and period of the heating. A possible rate-controlling mechanism for the fission product loss during the heating is discussed, and the results obtained are compared with some of the earlier work. The data may be interpreted as indicating that the fission products migrate through the graphite crystal to its surface according to a fast and a slow migration step. Escape from the graphite matrix by volatilization is apparently less rapid than the fast migration step; volatilization may be the rate-limiting mechanism in the loss process.

ISSN:0029-5639    

DOI:10.13182/NSE64-A18318

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor

摘要:

AbstractA control rod is described whose drop performance is essentially independent of the flow of reactor coolant. Theoretical studies covering wide ranges of parameter variations give a broad picture of the rod's potentialities.

ISSN:0029-5639    

DOI:10.13182/NSE64-A18319

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor

摘要:

AbstractA generalized method for estimating the energy release in hypothetical fast-reactor meltdown accidents is formulated. A modification of the Bethe-Tait method is derived from this more general formulation, and comparisons are made to an improved method, programmed for the IBM-7090 computer. Two basic assumptions are utilized: that the reactivity effects during disassembly can be calculated from perturbation theory, and that the decrease in density during disassembly can be ignored in the equations of hydrodynamics. It is shown that the threshold equation of state used in the Bethe-Tait method tends to cause an overestimate of the energy release for weak and moderate excursions, and that the saturated vapor pressure must be considered in those cases. The dependence of energy release upon prompt-neutron generation time, initial power level, rate of reactivity insertion, and Doppler effect is investigated.

ISSN:0029-5639    

DOI:10.13182/NSE64-A18320

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor

摘要:

AbstractAn experimental method of measuring parameters peculiar to a two-region nuclear reactor is developed requiring the measurement of the cross-power spectrum between the outputs of the two reactor regions when a random reactivity input is given to one of the regions. Using bandpass filters and an analog computer, the cross-power spectrum between the outputs of the two regions in the University of Florida Training Reactor has been measured. These data indicate that the propagation of a disturbance from one region of the reactor to the other region may adequately be described in terms of neutron-wave phenomena and that the method may be used to determine the multiplication factor of each region.

ISSN:0029-5639    

DOI:10.13182/NSE64-A18321

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor

摘要:

AbstractThis describes an experimental technique to determine the average velocity of the thermal-neutron spectrum as a function of time in a pulsed-neutron experiment. The measurement of the average velocity as a function of time is used to determine two parameters: the time necessary to establish an asymptotic spectrum, and the average velocity of the asymptotic spectrum. The variation in the asymptotic average velocity with material buckling is described by a“spectral-shift coefficient”which is related to the diffusion-cooling coefficient. It was found necessary to wait 2 milliseconds for the establishment of an equilibrium spectrum in graphite, and 0.6 milliseconds in heavy water, and that these values are insensitive to the geometric buckling. Values of the spectral-shift coefficient are given and compared with theoretical estimates.

ISSN:0029-5639    

DOI:10.13182/NSE64-A18322

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor

摘要:

AbstractA method of studying the burnup of high-cross-section materials in nuclear reactor control rods by neutron radiography is described. The technique has been applied to the examination of the burnup pattern of a CP-5 reactor control rod and has been found to provide a detailed picture of the burnup pattern, showing a very sharp transition region. The radiographic study has been made by a comparison method in which the neutron transmission of the irradiated cadmium control material has been compared to that of normal cadmium. In the regions in which the cadmium control material has been highly depleted in Cd113, the equivalent normal cadmium thickness comparison can be made to an estimated accuracy of 0.0006 inch.

ISSN:0029-5639    

DOI:10.13182/NSE64-A18323

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor