摘要:

AbstractExperimental measurements are reported for nucleate boiling of various saturated liquid polyphenyls which are of interest as nuclear reactor coolants. Heat flux is presented as a function of the difference between the heater surface temperature and saturation temperature, and correlations by various semitheoretical methods are discussed. A peak heat flux is reported for all but one of the liquids tested, and good agreement is found with previous work with similar fluids. Methods are suggested which may be used to estimate the nucleate boiling characteristics of these liquids during forced convection at elevated pressures and liquid subcooling, even though the present tests include only pool boiling studies. These methods may be used in feasibility analyses of boiling-polyphenyl cooled and moderated reactors.

ISSN:0029-5639    

DOI:10.13182/NSE59-A25610

出版商:Taylor&Francis    

年代:1959

数据来源: Taylor

摘要:

AbstractAn experimental and analytical study of the flux distribution of two-region core configurations has been made for the TRX facility. The purpose of this study was to obtain an estimate of the sizes of critical configurations that would yield the same values of the basic reactor parameters in the inner region as a critical core consisting entirely of the inner region material and geometry. Several two-region cores have been constructed and experimental measurements of thermal utilization, resonance escape probability, and fast fission effects have been performed. Slow and fast neutron activation distributions have also been obtained. Two inner regions were constructed utilizing 1.3 w/o enriched UO2fuel 0.384 in. in diameter and with a density of 10.53 gm/cm3. A third inner region utilized 1.3 w/o enriched uranium metal fuel with a diameter of 0.387 in. Light water served as the moderator and reflector in all cases. The experimental and theoretical results indicate that by utilizing two-region cores, measurements of microscopic parameters can be made for a wide variety of fuel sizes, fuel enrichments, and water-to-uranium volume ratios without the construction of full critical cores for each combination.

ISSN:0029-5639    

DOI:10.13182/NSE59-A25611

出版商:Taylor&Francis    

年代:1959

数据来源: Taylor

摘要:

AbstractAn efficient Monte Carlo method is described for computing the fundamental eigenvalue of the homogeneous transport equation. Importance sampling is used and, in effect, one neutron is followed through a large number of collisions.

ISSN:0029-5639    

DOI:10.13182/NSE59-1

出版商:Taylor&Francis    

年代:1959

数据来源: Taylor

摘要:

AbstractTemperature coefficients in large reactors can be obtained from the transient response of the flux to oscillations of control rods. A method is described with which the coefficients can be measured under full operating conditions and without special instrumentation or access to the pile. Thus, the technique is particularly useful in measuring the dependence of the coefficients upon hydraulic conditions, power level, and fuel exposure. The waveform of the perturbing oscillation of reactivity is trapezoidal so that the regular reactor control system can be used. In large reactors the flux shape changes during the portion of the cycle when the control rods are moving, but only the magnitude of the flux changes significantly while the control rods are stationary. The flux response during this latter portion of the cycle is analyzed for the temperature coefficients. The pile kinetics equations, coupled with equations for the temperatures of fuel, coolant, and moderator, are solved for the flux during the imposed oscillation. The temperature coefficients and their delay times are found by fitting computed fluxes to the observed flux.

ISSN:0029-5639    

DOI:10.13182/NSE59-A25613

出版商:Taylor&Francis    

年代:1959

数据来源: Taylor

摘要:

AbstractThe use of thermally black control sheets in a reactor is examined for three special cases: a finite slab reactor with a control sheet at its midplane, an infinite slab reactor containing an array of uniformly spaced control sheets, and a finite slab reactor with two control sheets placed symmetrically about the reactor centerline. The critical equation is obtained in each case and the physical significance of the solutions is examined by allowing the critical parameters to take limiting values. The conclusions reached are: (1) Forτ/L2«1, thermally black control sheets are effectively black to all neutrons and divide the reactor into independent parts provided the distanceWbetween control sheets, or core boundary and control sheet, is»τ3/2/L2. ForW≾τ3/2/L2, the control sheet is less effective. (2) Forτ/L2»1 andW»τ/L, thermally black control sheets effectively divide the reactor into independent parts. ForW≾τ/L, the control sheets are less effective. (3) Forτ/L2»1,W/L»1, andW2/τ∼1, a thermally black control sheet is relatively ineffective as compared with a sheet black to all neutrons. (4) The criteria for placing a given number of sheets most effectively in a reactor depend upon the worth of the sheets as determined from the conditions above. Thus, for sheets which are essentially black to all neutrons, the position of maximum effectiveness occurs when the reactor is cut into pieces of nearly equal size. However, for sheets of less worth, the positions of maximum effectiveness occur closer to the center of the reactor. In the limiting case, where the control effectiveness is very much smaller than the leakage from the reactor, the sheets should be placed about the reactor center, separated by about one diffusion length. It is pointed out that a very weak thermally black control element in a very large reactor may produce a large effect on the power distribution.

ISSN:0029-5639    

DOI:10.13182/NSE59-A25614

出版商:Taylor&Francis    

年代:1959

数据来源: Taylor

摘要:

AbstractAn analytical heat transfer solution is derived and evaluated for the general case of a turbulently flowing liquid metal which suddenly encounters a step-function boundary temperature in a channel system. Local Nusselt moduli, dimensionless mixed-mean fluid temperatures, and arithmetic-mean Nusselt moduli are given as functions of Reynolds and Prandtl moduli and a dimensionless axial-distance modulus. These solutions are compared with known solutions of more specific systems as well as with a set of experimental liquid-metal heat transfer data for a thermal entrance region. The results of this study can be used to define the temperature structure and heat transfer in solid-fuel element reactor cores and accelerator targets that are being cooled by liquid metals.

ISSN:0029-5639    

DOI:10.13182/NSE59-A25615

出版商:Taylor&Francis    

年代:1959

数据来源: Taylor

摘要:

AbstractTransfer functions of nuclear reactors and counterflow heat exchangers are derived from the partial differential equations with respect to time and spatial coordinates describing the transient behavior of nuclear power plants. These transfer functions can be approximated by lumped electrical networks and pure delays for analog computer studies. The procedure of approximation is illustrated by specific examples.

ISSN:0029-5639    

DOI:10.13182/NSE59-A25616

出版商:Taylor&Francis    

年代:1959

数据来源: Taylor

ISSN:0029-5639    

DOI:10.13182/NSE59-A25617

出版商:Taylor&Francis    

年代:1959

数据来源: Taylor