摘要:

AbstractA digital model for the application of analyzing the primary pressure transient of nuclear power plants is presented. No restrictive assumptions have been made, and a rigorous treatment of the conservation laws of energy and mass has been used. A comparison with experimental data from three different sources indicates good overall agreement. The model can be used as a design tool for pressurizer vessels and associated pressure control devices.

ISSN:0029-5639    

DOI:10.13182/NSE73-A19475

出版商:Taylor&Francis    

年代:1973

数据来源: Taylor

摘要:

AbstractNonlinear analysis has shown that when the buckling of a nuclear reactor with negative feedback is increased, the flux, under appropriate conditions, will proceed to a new asymptotically stable state. This contrasts with the linear theory which predicts a runaway. In this work, the method of“coordinate stretching”has been used to obtain the asymptotic solution of a nonlinear nuclear reactor under the combined effect of an initial positive disturbance and a negative feedback based on the Newton’s law of cooling.The minimum stability condition is derived by requiring that a bounded new equilibrium state exist. This condition sets an upper limit to the magnitude of the initial disturbance beyond which an equilibrium solution does not exist. Furthermore, the magnitude of the equilibrium flux is determined explicitly in terms of several relevant physical properties of the system: feedback coefficient, energy production rate, and rate of energy transfer to coolant.

ISSN:0029-5639    

DOI:10.13182/NSE73-A19476

出版商:Taylor&Francis    

年代:1973

数据来源: Taylor

摘要:

AbstractAs part of a continuing program to accurately measure neutron capture cross sections, we have obtained data for165Ho and197Au in the 167-eV to 0.6-MeV energy range. All capture reactions are measured relative to the fission rate of235U for neutron energies above 800 eV and to the10B(n,αo+α1) reaction below 13 keV. In addition, current best estimates ofσF(235U) andσn,α(10B) are used to convert these ratios to up-to-date capture cross sections. The 3.9-eV“black resonance”of165Ho is used to normalize all cross sections.

ISSN:0029-5639    

DOI:10.13182/NSE73-A19477

出版商:Taylor&Francis    

年代:1973

数据来源: Taylor

摘要:

AbstractResonance parameters for147Pm and148mPm have been determined and the resonances147Pm have been analyzed up to 317 eV. Only one resonance was observed for148mPm; its location at 0.17 eV is important for thermal reactor calculations Transmission measurements were initially made on four samples of147Pm2O3containing 1.6% of the decay product147Sm. Later measurements, made on a mixed sample of (148mPm +147Pm)2O3and on147Sm, provided resonance parameters for147Pm and an opportunity for a complete reevaluation of the previous147Pm results. Corrections for the samarium content could also be made. Both shape and area analyses were performed. The measured total cross section for147Pm at 0.025 eV was 198±8 b and a capture resonance integral of 2280±200 b was calculated. Using the parameters of the148mPm resonance at 0.17 eV, a corresponding total thermal-neutron cross section of 10 600 b was ob-tained. Integral measurements with cadmium-covered samples gave a value of (3.6±2.4)×103b for the resonance integral of148mPm, thereby setting an upper limit of 6000 b.

ISSN:0029-5639    

DOI:10.13182/NSE73-A19478

出版商:Taylor&Francis    

年代:1973

数据来源: Taylor

摘要:

AbstractThe thermalization of neutrons was experimentally measured for the purpose of testing Kushneriuk’s thermalization calculation in a cylindrical shell of moderator surrounding a fuel rod in a thermal reactor.The basis of this study was the measurement of activations for a set of five detectors—55Mn,197Au,115In,239Pu, and176Lu—inside the central fuel rod and the surrounding moderator at 20-cm hexagonal lattice pitch in the Atomic Energy of Canada Ltd.’s (AECL’s) heavy-water-moderated, natural uranium critical reactor, ZEEP. The neutron flux distribution, calculated by Kushneriuk’s method (first-order and exact solutions), was used in calculating detector activations. When compared, the measured detector activations and those calculated showed good agreement in the moderator, with the exception of points near and inside the fuel rod.The measured detector activations were also compared with the predicted values obtained from the HAMMER code. In this instance, the agreement was generally better than that obtained between measured and calculated values based on Kushneriuk’s method.

ISSN:0029-5639    

DOI:10.13182/NSE73-A19479

出版商:Taylor&Francis    

年代:1973

数据来源: Taylor

摘要:

AbstractCylindrical proportional counters were used to measure the scalar neutron flux in the core and reflector of the235U-fueled fast subcritical reactors, STSF-7 and -9. These data have been compared with time-of-flight angular spectrum measurements in the same reactors, and with transport theory calculations of these assemblies utilizing ENDF/B-III cross sections.The agreement of the two measurement techniques at high energy indicates that the emission time correction procedures used for the time-of-flight studies on subcritical (keff=≍0.92) assemblies with a 220-m flight path are reliable. At intermediate energies, the two types of measurements and the calculations were in good agreement for the STSF-9 oxide-fueled core, but significantly different for the metal-fueled STSF-7 core in the region of fine structure associated with the pronounced aluminum resonances below 300 keV. The time-of-flight data exhibited some energy mismatch here, and the 4πcounter was subjected to more aluminum“shielding”than allowed for in the calculations. The shielding effect was strongest in the metal-fueled STSF-7 core, because of the absence of the oxygen moderator.At low energy, where the Doppler effects of the thermal coefficient of reactivity are concentrated, the calculations underpredict the flux. The time-of-flight data, which are the most reliable here, show the greatest disagreement with theory.

ISSN:0029-5639    

DOI:10.13182/NSE73-A19480

出版商:Taylor&Francis    

年代:1973

数据来源: Taylor

摘要:

AbstractThe effective delayed-neutron fraction,βeff, has been measured in three cores of the fast zero power reactor FRO. The variance-to-mean method, in which the statistical fluctuations of the neutron density in the reactor are studied, was used. The experimental results were compared with theoretical values ofβeffcalculated with perturbation theory. The difference was at most 10%, which should be com-pared to the accuracy in the experimental values of≍3%.

ISSN:0029-5639    

DOI:10.13182/NSE73-A19481

出版商:Taylor&Francis    

年代:1973

数据来源: Taylor

摘要:

AbstractCriticality studies were made of three-dimensional arrays of uranium-metal cylinders enriched to 93.2 wt% in235U. Four weight groups of units, ranging from 10.4 to 26.2 kg of uranium in five geometries, were employed to determine the critical surface separation between units as a function of the number in an array. The influence on criticality of hydrogenous neutron reflecting and moderating materials, unit shape, array shape, and of other controlled perturbations to some assemblies was examined. Monte Carlo calculations were performed of the experimental assemblies to confirm the neutron multiplication factors and to interpret the behavior of several subcritical assemblies. The Hansen-Roach neutron cross-section sets reproduce the results of the experiments, in a majority of cases, to an accuracy of≍1% inkeff.

ISSN:0029-5639    

DOI:10.13182/NSE73-A19482

出版商:Taylor&Francis    

年代:1973

数据来源: Taylor

摘要:

AbstractThe reactivity worths of synthetic plutonium-uranium clustered fuel elements in a heavy-water-moderated assembly have been experimentally determined using the reactor oscillation method.Several test-fuel compositions have been investigated, representing varying degrees of fucl burnup and burnup distributions; two uranium samples with different U enrichment have been used as standard.The technique selected was aimed to establish“clean”experimental conditions, in order to effectively simplify the analysis of the results. Basically, the technique involved oscillating, according to a square-wave pattern, a 6-m-long fuel element containing a 50-cm-high test section with the fuel composition to be investigated the corresponding neutron density modulation was interpreted in terms of a Fourier analysis.The results of the experiment form a consistent set of data that can be used as test values for refined reactor burnup calculation codes. The overall experimental error, typically±0.015 pcm (1 pcm = 10-5Δkeff/keff). is considered remarkably low in view of the massive experimental setup required.A method for the theoretical analysis of the measured reactivity worths is presented. A multigroup perturbation transport calculation in one dimension (S4approximation) has been developed to account for the radial environmental conditions. The axial effects have been evaluated with a two-dimensional transport calculation. The group cross-section data used in the analysis were basically taken from the GAM-II and GATHER-II libraries.Using the same basic one-dimensional code with an appropriately adjusted input parameter, infinite lattice multiplication factors have also been calculated from he experimental reactivity results. These results are compared to the values of k∞obtained from null-reactivity measurements of identical clusters which were performed in association with the Comitato Nazionale Energia Nucleare, Italy, in RB-1 Reactor in Bologna. The agreement between the two sets of results is satisfactory.

ISSN:0029-5639    

DOI:10.13182/NSE73-A19483

出版商:Taylor&Francis    

年代:1973

数据来源: Taylor

摘要:

AbstractFor the specification of the cross sections of the fissile isotopes in the neu-tron energy region of unresolved resonances, the single-level formalism is often used, while an analysis of the cross sections in the resolved region indicates that a multilevel formula may be more appropriate. In this paper, we compare the statistical properties of the cross sections generated using the single-level formalism with those obtained by a multilevel formulation. The multilevel parameters were chosen to give the same average cross sections as the single-level formalism. The comparison indicates that there are small, but significant, differences between the statistical properties of the cross sections obtained with the multilevel formalism and those obtained with the single-level formula. The differences are probably too small, particularly when Doppler broadening is considered, to affect reactor calculations.

ISSN:0029-5639    

DOI:10.13182/NSE73-A19484

出版商:Taylor&Francis    

年代:1973

数据来源: Taylor