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Covariances of fission-integral measurements at the NBS /sup 252/Cf and ISNF facilities and at the ORNL-PCA facility

Description: Integral measurements together with accompanying uncertainty estimates have been used for the past fifteen years in cross section adjustments. As the field of cross section adjustment came of age, the crude uncertainty estimates were replaced - only in principle, initially - by a quantitative cross section uncertainty covariance description and by uncertainty correlations of integral experiments. There is current interest in the fission reaction rate ratio measurements in the NBS standard neutron fields by people involved in fast reactor cross sections. Also those in the LWR pressure vessel surveillance dosimetry program are interested in these measurements and in similar measurements performed in the Oak Ridge Pool Critical Assembly (PCA). A careful re-examination of uncertainty analysis is presented.
Date: January 1, 1979
Creator: Wagschal, J.J.; Maerker, R.E. & Gilliam, D.M.
Partner: UNT Libraries Government Documents Department

Development and verification of multicycle depletion perturbation theory

Description: Recently, Williams has developed a coupled neutron/nuclide depletion perturbation theory (DPT) applicable to multidimensional and multigroup reactor analysis problems. This theoretical framework has been verified using the newly developed DEPTH module within the context of the VENTURE modular code system. The accuracy and usefulness of this alternate calculational method for burnup analyses has been demonstrated for a variety of final-time response functionals. However, these examples were restricted to single-cycle depletion analyses due to the theoretical assumption that the nuclide density field was continuous in time. Clearly, in multicycle problems, the nuclide concentrations must vary discontinuously with time to model refueling and shuffling operations or discrete control rod movements. Thus, the purpose of this work is to generalize the original DPT framework to include nuclide discontinuities and to verify that this generalization can be employed in realistic multicycle applications.
Date: January 1, 1980
Creator: White, J.R. & Burns, T.J.
Partner: UNT Libraries Government Documents Department

Nuclear data adjustment methodology utilizing resonance parameter sensitivities and uncertainties

Description: This work presents the development and demonstration of a Nuclear Data Adjustment Method that allows inclusion of both energy and spatial self-shielding into the adjustment procedure. The resulting adjustments are for the basic parameters (i.e., resonance parameters) in the resonance regions and for the group cross sections elsewhere. The majority of this development effort concerns the production of resonance parameter sensitivity information which allows the linkage between the responses of interest and the basic parameters. The resonance parameter sensitivity methodology developed herein usually provides accurate results when compared to direct recalculations using existing and well-known cross section processing codes. However, it has been shown in several cases that self-shielded cross sections can be very non-linear functions of the basic parameters. For this reason caution must be used in any study which assumes that a linear relationship exists between a given self-shielded group cross section and its corresponding basic data parameters.
Date: January 1, 1984
Creator: Broadhead, B.L.
Partner: UNT Libraries Government Documents Department

CSRL-V ENDF/B-V library and thermal reactor and criticality safety benchmarks

Description: CSRL-V, an ENDF/B-V 227-group neutron cross-section library which has recently been expanded to include Bondarenko factor data for unresolved resonance processing, was used to calculate performance parameters for a series of thermal reactor and criticality safety benchmarks. Among the thermal benchmarks calculated were the Babcock and Wilcox lattice critical experiments B and W-XIII and B and W-XX. These two slightly-enriched (2.46%) UO/sub 2/, water-moderated, tight-pitch lattice experiments were chosen because (a) they have similar U/sup 238/ resonance shielding characteristics as power reactor cores, and (b) they provide benchmark results representative of high-leakage and low-leakage lattices, respectively. Among the criticality safety benchmarks calculated were homogeneous, highly-enriched (93.2%) uranyl fluoride spheres with hydrogen-to-uranium ratios varying from 76 to 972.
Date: January 1, 1982
Creator: Ford, W.E. III; Diggs, B.R.; Knight, J.R.; Greene, N.M.; Petrie, L.M. & Williams, M.L.
Partner: UNT Libraries Government Documents Department

Solid angle and surface density as criticality parameters

Description: Two methods often used to establish nuclear criticality safety limits for operations with fissile materials are the surface density and solid angle techniques. The two methods are used as parameters to express experimental and validated calculations of critical configurations. It is demonstrated that each method can represent critical arrangements of subcritical units and that there can be established a one-to-one correspondence between them. The analyses further show that the effect on an array neutron multiplication factor of perturbations to the array can be reliably estimated and that each form of fissile material and unit shape has a specific representation.
Date: October 1, 1980
Creator: Thomas, J.T.
Partner: UNT Libraries Government Documents Department

The Los Alamos Critical Experiments Facility Program

Description: Critical assemblies of precisely known materials and reproducible and easily calculated geometries have been constructed at the Los Alamos National Laboratory since the 1940s. Initially, these assemblies were built to provide information necessary for the nuclear weapons development effort. Subsequently, intensive studies of the assemblies themselves were undertaken to provide a better understanding of the physics of the fission process and other nuclear reactions in the nuclear materials from which these machine were constructed and in other materials irradiated in these assemblies. Some of these assemblies (notably Jezebel, Flattop, Big Ten, and Godiva) have been used as benchmark assemblies to compare the results of experimental measurements and computations of certain nuclear reaction parameters. These comparisons are used to validate both the input nuclear data and the computational methods. In addition to these normally fueled benchmark assemblies, other assembly machines are fueled periodically to provide specific and detailed results for parameter sensitivity studies for a large number of applications. Some of these machines and their applications are described.
Date: January 1, 1987
Creator: Dowdy, E.J.
Partner: UNT Libraries Government Documents Department

Magnitude of bias in Monte Carlo eigenvalue calculations

Description: Most Monte Carlo eigenvalue calculations are based on power iteration methods, like those used in analytical algorithms. But if N/sub H/, the number of histories in each generation is fixed, then such Monte Carlo calculations will be biased. Various arguments lead to the conclusion that eigenvalue and shape biases are both proportional to 1/N/sub H/, but little more is known about their magnitudes. Numerical experiments on simple matrices suggest that the biases are small, but information more relevant to real reactor calculations is very sparse. In fact to determine the bias in real reactor calculations is quite expensive. It seems worthwhile, therefore, to try to understand the Monte Carlo biases in systems more realistic than arbitrary matrices, but simpler than real reactors. For this reason biases in simple one-group model problems have been computed.
Date: January 1, 1983
Creator: Bowsher, H.; Gelbard, E.M.; Gemmel, P. & Pack, G.
Partner: UNT Libraries Government Documents Department

Implementation, verification, and application of multicycle depletion perturbation theory

Description: Several application-oriented features of generalized depletion perturbation theory (DPT) are analyzed from the viewpoint of the reactor designer. The detailed theory is first reduced to some new terminology necessary for an adequate understanding of DPT. Using this terminology, the main features and computational accuracy of this new technique are illustrated through representative DPT calculations utilizing a CDS-type heterogeneous reactor model. Several examples are presented that indicate the potential of DPT methods as an alternate computational tool for certain types of reactor physics analyses.
Date: January 1, 1980
Creator: White, J.R.; Burns, T.J. & Williams, M.L.
Partner: UNT Libraries Government Documents Department

Event Handler: a fast programmable, CAMAC-coupled data acquisition interface

Description: The purpose of this paper is to describe the architecture and performance of the Event Handler, a fast, programmable data acquisition interface which is linked to and through CAMAC. The special features of this interface make it a powerful tool in implementing data acquisition systems for experiments in nuclear physics.
Date: January 1, 1978
Creator: Hensley, D.C.
Partner: UNT Libraries Government Documents Department

Finite difference solution of the time dependent neutron group diffusion equations

Description: In this thesis two unrelated topics of reactor physics are examined: the prompt jump approximation and alternating direction checkerboard methods. In the prompt jump approximation it is assumed that the prompt and delayed neutrons in a nuclear reactor may be described mathematically as being instantaneously in equilibrium with each other. This approximation is applied to the spatially dependent neutron diffusion theory reactor kinetics model. Alternating direction checkerboard methods are a family of finite difference alternating direction methods which may be used to solve the multigroup, multidimension, time-dependent neutron diffusion equations. The reactor mesh grid is not swept line by line or point by point as in implicit or explicit alternating direction methods; instead, the reactor mesh grid may be thought of as a checkerboard in which all the ''red squares'' and '' black squares'' are treated successively. Two members of this family of methods, the ADC and NSADC methods, are at least as good as other alternating direction methods. It has been found that the accuracy of implicit and explicit alternating direction methods can be greatly improved by the application of an exponential transformation. This transformation is incompatible with checkerboard methods. Therefore, a new formulation of the exponential transformation has been developed which is compatible with checkerboard methods and at least as good as the former transformation for other alternating direction methods. (auth)
Date: August 1, 1975
Creator: Hendricks, J.S. & Henry, A.F.
Partner: UNT Libraries Government Documents Department

Acceleration techniques for response matrix methods

Description: Application of the power iteration method to the multigroup response matrix equations reduces them to a series of one-group problems. Applying acceleration techniques to each of these monoenergetic problems results in substantial reductions in computational effort. The use of point over-relaxation methods in the solution of these equations is described. Over-relaxation is also applied to the outer-iteration eigenvalue and source estimations with considerable success. 10 references. (auth)
Date: January 1, 1975
Creator: Sicilian, J.M. & Pryor, R.J.
Partner: UNT Libraries Government Documents Department

Application of the FORSS sensitivity code system to fast reactor analysis

Description: The FORSS Sensitivity Analysis Code System is described in terms of its objectives and present capabilities. An example is made of a problem specified by the Processing Methods Testing Subcommittee of the Code Evaluation Working Group, i.e., the determination of integral parameters, sensitivities to cross- section data, methods and data uncertainties, and required cross-section accuracies for an infinite media of ZPR 6/7 core composition. (auth)
Date: October 22, 1975
Creator: Weisbin, C.R.; Oblow, E.M. & Mynatt, F.R.
Partner: UNT Libraries Government Documents Department

KENO IV: an improved Monte Carlo criticality program

Description: KENO IV is a multigroup Monte Carlo criticality program written for the IBM 360 computers. It executes rapidly and is flexibly dimensioned so the allowed size of a problem (i.e., the number of energy groups, number of geometry cards, etc., are arbitrary) is limited only by the total data storage required. The input data, with the exception of cross sections, fission spectra and albedos, may be entered in free form. The geometry input is quite simple to prepare and complicated three-dimensional systems can often be described with a minimum of effort. The results calculated by KENO IV include k-effective, lifetime and generation time, energy-dependent leakages and absorptions, energy- and region-dependent fluxes and region-dependent fission densities. Criticality searches can be made on unit dimensions or on the number of units in an array. A summary of the theory utilized by KENO IV, a section describing the logical program flow, a compilation of the error messages printed by the code and a comprehensive data guide for preparing input to the code are presented. 14 references. (auth)
Date: November 1, 1975
Creator: Petrie, L.M. & Cross, N.F.
Partner: UNT Libraries Government Documents Department

Multivariate statistical pattern recognition system for reactor noise analysis

Description: A multivariate statistical pattern recognition system for reactor noise analysis was developed. The basis of the system is a transformation for decoupling correlated variables and algorithms for inferring probability density functions. The system is adaptable to a variety of statistical properties of the data, and it has learning, tracking, and updating capabilities. System design emphasizes control of the false-alarm rate. The ability of the system to learn normal patterns of reactor behavior and to recognize deviations from these patterns was evaluated by experiments at the ORNL High-Flux Isotope Reactor (HFIR). Power perturbations of less than 0.1 percent of the mean value in selected frequency ranges were detected by the system. 19 references. (auth)
Date: January 1, 1975
Creator: Gonzalez, R.C.; Howington, L.C.; Sides, W.H. Jr. & Kryter, R.C.
Partner: UNT Libraries Government Documents Department

Final report [on solving the multigroup diffusion equations]

Description: Progress achieved in the development of variational methods for solving the multigroup neutron diffusion equations is described. An appraisal is made of the extent to which improved variational methods could advantageously replace difference methods currently used. (DG)
Date: January 1, 1975
Creator: Birkhoff, G.
Partner: UNT Libraries Government Documents Department

Relationship of observed flow patterns to gas core reactor criticality

Description: The gas core reactor requires the establishment of stable and unique flow patterns. A recent series of room temperature flow tests have studied the hydrodynamics, particularly involving gases of differing densities. In an actual operating gas core reactor, the central gas of vaporized uranium will have a much higher density than the surrounding coolant. Testing was done in two different sized chambers (18 inch and 36 inch diameter) to study hydrodynamic scaling. Air was employed as the ''coolant'' gas. Air, argon, and freon, smoked for identification, was used to simulate the fuel. A variety of injectors at various locations in the cavity were employed. (auth)
Date: January 1, 1975
Creator: Macbeth, P.J.; Kunze, J.F. & Rogers, V.C.
Partner: UNT Libraries Government Documents Department

Linear filtering applied to Monte Carlo criticality calculations

Description: A significant improvement in the acceleration of the convergence of the eigenvalue computed by Monte Carlo techniques has been developed by applying linear filtering theory to Monte Carlo calculations for multiplying systems. A Kalman filter was applied to a KENO Monte Carlo calculation of an experimental critical system consisting of eight interacting units of fissile material. A comparison of the filter estimate and the Monte Carlo realization was made. The Kalman filter converged in five iterations to 0.9977. After 95 iterations, the average k-eff from the Monte Carlo calculation was 0.9981. This demonstrates that the Kalman filter has the potential of reducing the calculational effort of multiplying systems. Other examples and results are discussed. (auth)
Date: January 1, 1975
Creator: Morrison, G.W.; Pike, D.H. & Petrie, L.M.
Partner: UNT Libraries Government Documents Department

Determination of diffusion parameters using response matrix theory

Description: A method is presented which provides the accuracy of the response matrix method, but without requiring the development of a response matrix reactor code. The method is used to determine diffusion parameters which, when used in conventional reactor diffusion codes, provide the same results as a response matrix reactor code. (auth)
Date: November 1, 1975
Creator: Pryor, R.J. & Sicilian, J.M.
Partner: UNT Libraries Government Documents Department

Application of depletion perturbation theory to fuel cycle burnup analysis

Description: Over the past several years static perturbation theory methods have been increasingly used for reactor analysis in lieu of more detailed and costly direct computations. Recently, perturbation methods incorporating time dependence have also received attention, and several authors have demonstrated their applicability to fuel burnup analysis. The objective of the work described here is to demonstrate that a time-dependent perturbation method can be easily and accurately applied to realistic depletion problems.
Date: January 1, 1979
Creator: White, J.R.
Partner: UNT Libraries Government Documents Department

Critical heat flux tests with high pressure water in an internally heated annulus with alternating axial heat flux distribution

Description: Critical heat flux experiments were performed with an alternating heat flux profile in an internally heated annulus. The heated length was 84 inches with a square wave alternating heat flux profile over the last 12 inches having a maximum-to-average heat flux ratio of 1.76. Test data were obtained at pressures from 800 to 2000 psia, mass velocities from 0.25 x 10/sup 6/ to 2.8 x 10/sup 6/ lb/hr-ft/sup 2/ and inlet temperatures ranging from 400 to 600/sup 0/F. Two different electrically heated test sections were employed both with 72 inch uniform and 12 inch alternating heat flux sections. The second test section had a 0.44 inch hot patch with a peak-to-average heat flux ratio of 2.7 superimposed on the alternating flux profile at the exit end. Critical heat flux results with the alternating heat flux profile and with the superimposed hot patch were shown to be equivalent to those obtained in previous tests with a uniform heat flux profile except for several data points at low mass velocity and high enthalpy for which there is an apparent experimental bias in the uniform heat flux results.
Date: September 1, 1979
Creator: Beus, S.G. & Humphreys, D.A.
Partner: UNT Libraries Government Documents Department