Showing papers in "Nuclear Science and Engineering in 1977"
TL;DR: In this paper, a class of acceleration schemes that resemble the conventional synthetic method in that they utilize the diffusion operator in the transport iteration schemes were investigated, and the authors investigated a set of acceleration methods that are similar to the ones described in this paper.
Abstract: We investigate a class of acceleration schemes that resemble the conventional synthetic method in that they utilize the diffusion operator in the transport iteration schemes. These schemes are not ...
288 citations
TL;DR: In this article, a coarse-mesh method for the solution of multidimensional neutron kinetics problems is presented that is based on the approximation of the desired solution by basis functions with local nonoverlapping supports corresponding to the volume elements of the spatial mesh.
Abstract: A coarse-mesh method for the solution of multidimensional neutron kinetics problems is presented that is based on the approximation of the desired solution by basis functions with local nonoverlapping supports corresponding to the volume elements of the spatial mesh. Integration of the approximating functions over their supports, and exploitation of continuity conditions for neutron flux and current, yields local seven-point difference operators with solution-dependent coupling coefficients. Due to the finite-difference (FD) structure of the resulting matrix equation, any technique developed for FD methods can be used for its solution. However, a novel (''almost implicit'') alternating direction explicit-implicit technique has been developed that is especially suited for coarse-mesh applications. Numerical examples that demonstrate the high efficiency of the method are presented. By using a spatial grid corresponding to the fuel element structure, it is possible to compute power distribution and its time history very accurately (at most, with a several percent error) at an economically tolerable expense.
137 citations
TL;DR: In this article, the heat transfer characteristics of a fluid-saturated porous media were investigated for the case of uniform internal heat generation with cooling from above, and analytical models of conduction and siamese conduction were presented.
Abstract: The heat transfer characteristics of a fluid-saturated porous media are investigated for the case of uniform internal heat generation with cooling from above. Analytical models of conduction and si...
101 citations
84 citations
TL;DR: In the analysis of fission product reactivity worths, measured in the fast reactor spectra of the STEK critical-experiments facility, extensive use is made of a statistical method of cross section adjustment, which applies adjustments to the evaluated cross sections, taking into account the existing correlations.
Abstract: In the analysis of fission product reactivity worths, measured in the fast reactor spectra of the STEK critical-experiments facility, extensive use is made of a statistical method of cross section adjustment. The principle is that adjustments are applied to the evaluated cross sections, as much as possible within their error limits and taking into account the existing correlations, in such a way that a better agreement between calculated and measured integral data is obtained. The method is briefly summarized in general terms, with some special applications needed for the STEK project. Then, a description is given of the practical realization for capture cross-section adjustment on the basis of reactivity worths of samples of fission product mixtures in different thicknesses measured in several fast reactor spectra. Details are given on the way the various contributions to the covariance matrix of group cross sections, including resonance self-shielding, are calculated for the fission product nuclides and for the mixtures. The paper outlines the methods used. Some limitations of the method and possible extensions in connection with standard nuclear data error files are discussed.
82 citations
TL;DR: Using projection operators, this article derived x-y geometry discrete ordinates-to-spherical harmonics (SN → PN-1.) fictitious sources defined in the literature as ray effect mitigating devices.
Abstract: Using projection operators, we rederive x-y geometry discrete ordinates-to-spherical harmonics (SN → PN-1.) fictitious sources defined in the literature as ray-effect mitigating devices. We define ...
70 citations
TL;DR: In this article, a one-dimensional (axial) model of the core is used to solve the two-group diffusion equations satisfied by the neutron noise, and the solution is composed of two terms that can be identified as the theoretical counterparts of the components found in experiments.
Abstract: In view of recent experimental work, the neutron noise in a boiling water reactor is believed to be separable into local and global components. It is the existence of the local component that makes possible the measurement of steam velocity by correlating the signals of axially placed in-core neutron detectors. A one-dimensional (axial) model of the core is used to solve the two-group diffusion equations satisfied by the neutron noise. The solution is shown to be composed of two terms that can be identified as the theoretical counterparts of the components found in experiments. The properties of the two terms are discussed in the special case of an axially propagating disturbance of the moderator density (steam content).
65 citations
TL;DR: In this paper, the authors measured the fission cross-section ratios of fission chambers from 0.1 to 30 MeV using ionization fission chamber and the time-of-flight data.
Abstract: We have measured the fission cross-section ratios 234U:235U, 236U:235U, and 238U:235U as a function of neutron energy from 0.1 to 30 MeV using ionization fission chambers and the time-of-flight tec...
61 citations
TL;DR: In this article, the authors discuss the nuclear structure and nuclear deformation in terms of nuclear structures and deformation of nuclear power plants, and propose a model of nuclear structure, which is based on nuclear structures.
Abstract: (1977). Nuclear Structure, Vol. II: Nuclear Deformations. Nuclear Science and Engineering: Vol. 62, No. 4, pp. 771-771.
56 citations
TL;DR: Delayed neutron emission spectra from thermal-neutron fission of 233U, 235U, 239Pu, and 241Pu were analyzed in this article...
Abstract: Delayed neutron emission spectra from thermal-neutron fission of 233U, 235U, 239Pu, and 241Pu, from fast-neutron fission of 232Th, 235U, 238U, and 239Pu and from high-energy neutron (14.7-MeV) fiss...
53 citations
TL;DR: In this article, multivariate autoregressive (AR) procedures are introduced as diagnostic tools to extract dynamic,characteristics for detection of malfunctions of a boiling water reactor (BWR) power plant.
Abstract: Multivariate autoregressive (AR) procedures are introduced as diagnostic tools to extract dynamic,characteristics for detection of malfunctions of a boiling water reactor (BWR) power plant. The pro...
TL;DR: In this article, a simple one-parameter model is presented for calculating the distribution of independent yield strength between ground and isomeric states of primary fission products formed by neutron-induced fission of actinide nuclei.
Abstract: A simple one-parameter model is presented for calculating the distribution of independent yield strength between ground and isomeric states of primary fission products formed by neutron-induced fission of actinide nuclei Yield branching ratios are calculated as a function of neutron energy (thermal, fast, and 14-MeV) for 33 cases that span 144 fission product nuclei having isomeric states with known spins
TL;DR: Optimized iteration methods for the solution of large-scale fast reactor finite difference diffusion theory calculations are presented, and the performance of a computer code employing these methods is compared with that of several existing production diffusion theory codes for a range of typical problems.
Abstract: Optimized iteration methods for the solution of large-scale fast reactor finite difference diffusion theory calculations are presented, along with their theoretical basis. The computational and dat...
TL;DR: In this article, the authors measured beryllium neutron-production cross sections using the time-of-flight technique at incident neutron energies of 5.9, 10.1, and 14.2 MeV, and at laboratory angles of 25, 27.5, 30, 35, 45,...
Abstract: We measured beryllium neutron-production cross sections using the time-of-flight technique at incident neutron energies of 5.9, 10.1, and 14.2 MeV, and at laboratory angles of 25, 27.5, 30, 35, 45,...
TL;DR: The concept of channel theory is used to locate spatial regions that are important in contributing to a shielding response, and sample problems are given to exhibit and verify properties predicted by the mathematical equations.
Abstract: The concept of channel theory is used to locate spatial regions that are important in contributing to a shielding response. The method is analogous to the channel-theory method developed for ascertaining important energy channels in cross-section analysis. The mathematical basis for the theory is shown to be the generalized reciprocity relation, and sample problems are given to exhibit and verify properties predicted by the mathematical equations. A practical example is cited from the shielding analysis of the Fast Flux Test Facility performed at Oak Ridge National Laboratory, in which a perspective plot of channel-theory results was found useful in locating streaming paths around the reactor cavity shield.
TL;DR: In this article, a set of approximate solutions for the isotropic two-dimensional neutron transport problem has been developed using the interface current formalism, which has been applied to regular lattices of rectangular cells containing a fuel pin, cladding, and water, or homogenized structural material.
Abstract: A set of approximate solutions for the isotropic two-dimensional neutron transport problem has been developed using the interface current formalism. The method has been applied to regular lattices of rectangular cells containing a fuel pin, cladding, and water, or homogenized structural material. The cells are divided into zones that are homogeneous. A zone-wise flux expansion is used to formulate a direct collision probability problem within a cell. The coupling of the cells is effected by making extra assumptions on the currents entering and leaving the interfaces. Two codes have been written: The first uses a cylindrical cell model and one or three terms for the flux expansion, and the second uses a two-dimensional flux representation and does a truly two-dimensional calculation inside each cell. In both codes, one or three terms can be used to make a space-independent expansion of the angular fluxes entering and leaving each side of the cell. The accuracies and computing times achieved with the different approximations are illustrated by numerical studies on two benchmark problems.
TL;DR: In this paper, a major effort was made to obtain an accurate experimental determination of the energy response and efficiency function of the neutron detector over the entire neutron energy range of interest, and the best description was obtained with the distribution N/sub 1/(E) exp(-1.02E)sinh(2.32E)/sup /sup 1///sub 2//.
Abstract: The prompt fission neutron spectrum emitted by a sample of /sup 235/U irradiated with 0.53-MeV neutrons has been measured in the 0.6- to 15-MeV energy range by using time-of-flight (TOF) techniques. In the present work, a major effort was made to obtain an accurate experimental determination of the energy response and efficiency function of the neutron detector over the entire neutron energy range of interest. For this purpose, the TOF spectrometer was calibrated with respect to energy in the 0.5- to 21-MeV range by observing neutron groups from various nuclear reactions. The energy dependence of the neutron detector efficiency was determined by observing the angular distributions of the H(n,n)H process in the 1- to 15-MeV energy range. The overlapping 0.6- to 3-MeV energy range was covered by the T(p,n)/sup 3/He reaction. The result of the fission neutron spectrum measurements has been used to find a suitable distribution function describing the data in the entire energy interval. The best description was obtained with the distribution N/sub 1/(E) exp(-1.02E)sinh(2.32E)/sup /sup 1///sub 2//. 6 figures, 4 tables, 39 references.
TL;DR: In this paper, the Galerkin formulation of the finite element method is applied in space and angle to the equivalent integral law, or weak form, of the first-order neutron transport equation.
Abstract: The Galerkin formulation of the finite element method is applied in space and angle to the equivalent integral law, or weak form, of the first-order neutron transport equation. The existence of a unique solution to the resultant system of algebraic equations is demonstrated using the positivity of the transport operator. Numerical results are given for the one-dimensional plane geometry application, including comparisons with the one-dimensional discrete ordinates code ANISN. A problem with strong heterogeneities is considered, and the use of discontinuous angular and spatial finite elements is shown to result in a marked improvement in the results. The success of the discontinuous elements is examined and it is seen that the discontinuous angular elements effectively match the analytical discontinuity in the angular flux at ..mu.. = 0 for plane geometry. Also, the use of discontinuous spatial elements is shown to result in treating continuity of the angular flux at an interface as a natural interface condition in the direction of neutron travel.
TL;DR: In this article, two methods for solving the transient group diffusion equations for reactors composed of large homogeneous nodal regions were developed, in which nodal coupling constants are in effect computed by an analytical method; in the second, a polynomial expansion of the flux is used.
Abstract: Two methods are developed for solving the transient group diffusion equations for reactors composed of large homogeneous nodal regions. In the first scheme, nodal coupling constants are in effect computed by an analytical method; in the second, a polynomial expansion of the flux is used. Both methods yield accurate static and dynamic results in very short computing times.
TL;DR: In this paper, the accuracy and efficiency of a class of asymmetric weighted residual methods, as applied to neutron diffusion equations, is presented. And it is shown that for normal reactor conditions, sufficiently accurate results can already be obtained with a third-order polynomial without mixed-derivative terms.
Abstract: A systematic study of the accuracy and efficiency of a class of asymmetric weighted residual methods, as applied to neutron diffusion equations, is presented. Polynomials up to the sixth order are considered, with and without mixed spatial derivative terms. It turns out that the sixth-order polynomial with mixed derivative terms is most efficient; yet, for normal reactor conditions, sufficiently accurate results can already be obtained with a third-order polynomial without mixed-derivative terms.
TL;DR: In this article, two methods of uncertainty anaylsis, called the response surface method and the crude Monte Carlo method, were compared for the probability density function of the peak cladding temperature as computed by a simplified nuclear code that was subjected to seven uncertainty parameters.
Abstract: A demonstration of two methods of uncertainty analysis was carried out to assess their utility for future use in treating computer models of nuclear power systems. The two methods of uncertainty anaylsis, called the response surface method and the crude Monte Carlo method, produced comparable results for the probability density function of the peak cladding temperature as computed by a simplified nuclear code that was subjected to seven uncertainty parameters. From these density functions, the upper cumulative tail probabilities were obtained and were shown to be measures of parameter margin. The response surface method provides sensitivity coefficients and also an inexpensive framework for evaluating the effects of the various assumptions inherent in the method. The crude Monte Carlo method provides no sensitivity coefficients and requires a complete rerun if a single uncertainty input density should be changed. The response surface method is recommended for use, where economically feasible, since the advantages of the method far outweigh the disadvantages.
TL;DR: In this paper, three classes of true analogs of the one-dimensional singleGauss and double-Gauss are considered for two-dimensional x-y problems with rectangular spatial mesh subdivisions.
Abstract: The main features of this paper are the utilization of inherent two-dimensional symmetries and the development of accurate angular quadrature coordinates and weights especially suited for the net and/or partial currents and all the net and/or partial moments of the neutron flux up to a given order. Three classes of true analogs of the one-dimensional single-Gauss and double-Gauss are considered for two-dimensional x-y problems with rectangular spatial mesh subdivisions. The first is single-range quadrature, most suitable for the asymptotic regions where the vector flux of neutrons can be well approximated by polynomials in ..cap omega../sub x/ and ..cap omega../sub y/ defined over the entire unit sphere of angular directions ..cap omega... This quadrature can be used whenever distances between material interfaces are large with respect to the neutron mean-free-path (mfp). The second is double-range quadrature, most suitable at material interfaces where the unit sphere can be split into two hemispheres, one in each material region, and the vector flux can be well approximated by two possibly distinct polynomials in ..cap omega../sub x/ and ..cap omega../sub y/, one in each hemisphere. This quadrature can be used whenever material interfaces and currents are important along either the x or the y directionmore » but not both. The third is quadruple-range quadrature, most suitable at corners where the unit sphere can be split into four quadrants and the vector flux can be well approximated by four possibly distinct polynomials in ..cap omega../sub x/ and ..cap omega../sub y/, one in each quadrant. This quadrature explicitly allows for discontinuities at corners and is appropriate for highly heterogeneous problems where distances between material corners are small with respect to the mfp. For simplicity, only product formulas are considered, where the angular integrals are split into separate integrals over polar and azimuthal directions.« less
TL;DR: In this article, a strategy for flow-boiling analysis for the equal velocity and equal temperature model is presented, with particular emphasis on the role of benchmark solutio in the analysis.
Abstract: A strategy for flow-boiling analysis development is illustrated through application to the equal-velocity and equal-temperature model Particular emphasis is placed on the role of benchmark solutio
TL;DR: In this paper, the multiple collision technique was applied to the monoenergetic time-dependent neutron transport equation for pulsed plane source emission in an infinite medium to obtain the flux due to a pulsed point source in the same medium.
Abstract: The multiple collision technique as applied to the monoenergetic time-dependent neutron transport equation for pulsed plane source emission in an infinite medium is used to obtain the flux due to a pulsed point source in the same medium This result is then integrated to determine the flux due to the corresponding pulsed line source problem The semi-infinite albedo problem is also shown to be solvable using the multiple collision approach A generalization to include delayed neutrons follows directly from the multiple collision treatment, as does an equivalence between a monoenergetic time-dependent problem and a particular stationary slowing down problem in infinite geometry Results are tabulated and comparisons are made to provide benchmark solutions to the fundamental time-dependent transport problems considered and thus bridge the gap between theory and practice
TL;DR: The 238U(n,f) cross section has been measured from 3 eV to ≃ 100 keV with the Rensselaer Intense Neutron Spectrometer as mentioned in this paper.
Abstract: The 238U(n,f) cross section has been measured from 3 eV to ≃ 100 keV with the Rensselaer Intense Neutron Spectrometer, a 75-ton lead slowing down spectrometer at the Gaerttner Laboratory at Renssel...
TL;DR: The application of generalized perturbation theory, both time dependent and static, to burnup problems is discussed in this paper, where simple analytical forms, applicable for many cases of interest, are derived.
Abstract: The application of generalized perturbation theory, both time dependent and static, to burnup problems is discussed. Simple analytical forms, applicable for many cases of interest, are derived for ...
TL;DR: In this article, the subject of width fluctuation correction to average compound-nucleus cross sections is reviewed, with special emphasis on neutron capture and scattering cross sections, and a review of recent stati...
Abstract: In this Note the subject of width fluctuation correction to average compound-nucleus cross sections is reviewed, with special emphasis on neutron capture and scattering cross sections. Recent stati...
TL;DR: In this paper, the authors used a heavily shielded Nal detector in conjunction with the white neutron spectrum from the Oak Ridge Electron Linear Accelerator (OELA) to calculate the gamma-ray energy distributions.
Abstract: Cross sections for the production of gamma rays with energies of 0.3 < E/sub ..gamma../ < 10.5 MeV have been measured as a function of neutron energy over the range 0.1 < E/sub n/ < 20.0 MeV. Results were obtained for 22 elements that are commonly encountered in the calculation of radiation effects. The measurements were made using a heavily shielded Nal detector in conjunction with the white neutron spectrum from the Oak Ridge Electron Linear Accelerator. Incident neutron energies were determined by time-of-flight over a 47-m flight path, while gamma-ray energy distributions were obtained from pulse-height unfolding techniques. Elemental differential cross sections are presented for Li, C, N, F, Mg, Al, Si, Ca, V, Cr, Fe, Ni, Cu, Zn, Nb, Mo, Ag, Sn, Ta, W, Au, and Pb.
TL;DR: In this paper, the worth of a calibrated 252Cf spontaneous fission source, together with its worth in terms of delayed neutron fractions in fast reactor spectra, was investigated.
Abstract: Integral measurements of delayed neutron fractions in fast reactor spectra by two different techniques were carried out. The worth of a calibrated 252Cf spontaneous fission source, together with ab...