摘要:

AbstractThis paper describes the transient behavior of TRIGA, a light-water-cooled reactor using fuel-moderator elements composed of uranium and zirconium hydride. The large, prompt negative temperature coefficient—an inherent characteristic of these fuel-moderator elements—limits reactor power transients primarily by means of fuel-element temperature rise rather than by void formation in the core. Step reactivity insertions of up to 1.6% resulted in peak powers of 250 Mw with no detectable boiling of the core water or expulsion of water from the core.

ISSN:0029-5639    

DOI:10.13182/NSE59-A28840

出版商:Taylor&Francis    

年代:1959

数据来源: Taylor

摘要:

AbstractThe Milne problem, with absorption, of neutron transport theory is solved by expanding the angular distribution in Tchebycheff polynomials, rather than the more conventional Legendre polynomials. It is shown than the Tchebycheff approximation of orderN, theTNapproximation, gives results for the extrapolated end point which are closer to the exact results over most of the range of absorption values considered than the correspondingPNapproximation.

ISSN:0029-5639    

DOI:10.13182/NSE59-A28841

出版商:Taylor&Francis    

年代:1959

数据来源: Taylor

摘要:

AbstractAn iterative method for solving problems in neutron transport theory is described. The method uses the result of a low-order spherical harmonies calculation as the initial function and attempts to improve this with one iteration. The results for the one-velocity Boltzmann equation with plane symmetry show that good accuracy is obtained for many practical cases.

ISSN:0029-5639    

DOI:10.13182/NSE59-A28842

出版商:Taylor&Francis    

年代:1959

数据来源: Taylor

摘要:

AbstractThis system uses a large organic scintillator as a moderator for a burst of fast neutrons, many of which are subsequently captured by the hydrogen in the scintillator. The pulses produced by the 2.2-Mev captureγ-rays are observed by four photomultiplier tubes whose anodes are paralleled. The output pulses are amplified and counted by a 10-Mc scaler. The scaler is gated to count for 300μsec after the pulse, during which interval background is very small. Statistically significant information on total neutron output may be obtained for as few as 103neutrons per pulse, with practically no upper limit. Relative calibration of the system is simple, and absolute calibrations are stable and reproducible.

ISSN:0029-5639    

DOI:10.13182/NSE59-A28843

出版商:Taylor&Francis    

年代:1959

数据来源: Taylor

摘要:

AbstractIn its elevated-temperature critical-assembly experiments, the University of California Radiation Laboratory, Livermore, will utilize stainless-steel-clad fuel elements. The fuel element consists of 0.001- or 0.002-in. thick enriched-uranium foil packaged in a welded 0.002-in. thick type-347 stainless-steel envelope. The design requirements for the fuel elements are stated, as are the considerations which led to the design selected. Beta heat treatment of the uranium was found necessary to provide compatible thermal-expansion characteristics for the two materials in the fuel element.

ISSN:0029-5639    

DOI:10.13182/NSE59-A28844

出版商:Taylor&Francis    

年代:1959

数据来源: Taylor

摘要:

AbstractIn reactors having D2O or Be moderators, the photoneutrons produced by gamma rays from long-lived fission products give rise to transient effects which could introduce appreciable errors into various types of reactivity measurements. This paper presents digital computer calculations of these effects in D2O for criticality measurements, subcritical multiplication measurements, rod drops, and rising period measurements. It is found that in some cases appreciable errors are possible even after one hour waiting periods. Since the Be photoneutron data cannot be resolved into groups, calculations for Be could not be done. However, the nature of the effects that could be expected is discussed.

ISSN:0029-5639    

DOI:10.13182/NSE59-A28845

出版商:Taylor&Francis    

年代:1959

数据来源: Taylor

摘要:

AbstractA Monte Carlo calculation of thermal utilization is described with emphasis on the statistical procedures employed. The reactor model has slightly enriched uranium fuel rods in a hexagonal lattice. The moderator is ordinary water but is treated by the code as a hydrogen gas. General principles are given for using Monte Carlo techniques and it is shown that time reducing methods are as effective as the better known variance reducing methods in decreasing the cost. Good agreement with experimental results was obtained, indicating that such thermal utilization calculations are feasible and are not sensitive to the model of the water molecule.

ISSN:0029-5639    

DOI:10.13182/NSE59-A28846

出版商:Taylor&Francis    

年代:1959

数据来源: Taylor

摘要:

AbstractWhen flux measurements are made in reactors or in piles, large ion chambers are commonly used. The current output of these chambers is read in terms of flux. The chambers depress the flux, however, and a correction should be applied to get the value of the unperturbed flux. The flux perturbation was measured in a large graphite diffusing medium, the Argonne National Laboratory Standard Pile, and found to be between 5% and 25% when measured on the outer surface of typical ion chambers. At about 10 in. from the end of the chamber the perturbation was no longer observed. The flux was measured with a small fission counter which, of itself, did not depress the flux appreciably.To measure the flux depression inside an ion chamber, the latter was simulated by stacking boron-coated aluminum plates above and below the small fission counter used previously. The measurement of the flux depression was found to be in good agreement with that which can be estimated from a calculation in which an exponential absorption is assumed.From these experiments it is concluded that the value of the flux measured with a large boron coated ion chamber gives an estimation of the flux within 20% to 50% of the unperturbed value depending on the amount of boron in the chamber, while the estimation of the flux is within 5% to 15% when measured with a large U235-coated fission counter. It should be noted that, although these results apply in a graphite diffusing medium, they do not necessarily apply in an absorbing medium such as the heavy concrete which usually surrounds the instrument holes in reactors.

ISSN:0029-5639    

DOI:10.13182/NSE59-A28847

出版商:Taylor&Francis    

年代:1959

数据来源: Taylor

摘要:

AbstractThe distributions of neutron flux and void in an orificed core of a boiling water reactor are calculated analytically by use of the relation between void generation and power. The analytical method is described, and it is shown that comprehensive surveys of the nuclear and heat performance of a boiling water reactor are possible with little effort. As an example, the technique of making the void map is shown. An alternate method for this problem is also described.

ISSN:0029-5639    

DOI:10.13182/NSE59-A28848

出版商:Taylor&Francis    

年代:1959

数据来源: Taylor

摘要:

AbstractA method is demonstrated whereby the velocity of flow during a coastdown transient in a loop may be analytically determined without the use of pump characteristic curves. The method is approximate, and it appears that the error involved is of the order of magnitude that may be expected in hydraulic calculations. A fair agreement is noted when the results of the analytical method are compared with those obtained by the use of the characteristic curves of both a single-suction and a double-suction centrifugal pump. A further comparison of the analytical results with collected experimental data for flow coastdown also shows a fairly good agreement.

ISSN:0029-5639    

DOI:10.13182/NSE59-A28849

出版商:Taylor&Francis    

年代:1959

数据来源: Taylor