摘要:

AbstractA nonlinear optimization method based on first-order generalized perturbation theory (GPT) and mathematical programming has been extended to three dimensions in the code OPTEX and applied to a realistic problem in the physics design of Canada deuterium uranium (CANDU) reactors. The choice of three-dimensional linear GPT for computing the cost coefficients is justified, and the optimization approach is discussed in reference to methods used for light water reactor fuel manage-ment. The design problem consists of simultaneously adjusting the fueling rate distribution and the grading of the adjuster rods in the core, while satisfying limits on the maximum bundle and channel powers at full power equilibrium refueling. Passage to three dimensions is a requirement for a real-istic modeling of equilibrium refueling in CANDU. It has a significant effect on the system equations, which become nonlinear with the inclusion of the axial dimension. The nature of the constraints is also affected: Separate limits on channel and bundle powers must now be accounted for. These problems are addressed, and a practical optimization scheme is proposed that can handle realistic CANDU core and fuel management design problems.

ISSN:0029-5639    

DOI:10.13182/NSE92-A23919

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractA reconstruction method is developed for recovering pin burnup characteristics from fuel cycle calculations performed in hexagonal-z geometry using the nodal diffusion option of the DIF3D/REBUS-3 code system. Intranodal distributions of group fluxes, nuclide densities, power density, burnup, and fluence are efficiently computed using polynomial shapes constrained to satisfy nodal information. The accuracy of the method is tested by performing several fast reactor numerical benchmark calculations and by comparing predicted local burnups with values measured for experimental assemblies in the Experimental Breeder Reactor II. The results indicate that the reconstruction methods are quite accurate yielding maximum errors in power and nuclide densities that are<2% for driver assemblies and typically<5% for blanket assemblies.

ISSN:0029-5639    

DOI:10.13182/NSE92-A23920

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractA new nodal method is developed for the solution of one-group discrete ordinates (SN) problems with linearly anisotropic scattering in x,y-geometry. In this method, the“spectral Green’s function”(SGF) scheme, originally developed for solving SNproblems in slab geometry with no spatial truncation error, is generalized to solve the one-dimensional transverse-integrated SNnodal equations with the“constant”approximation for the transverse leakage terms. The resulting“SGF-constant nodal”(SGF-CN) method is more accurate than conventional coarse-mesh methods for deep penetration problems because it treats the scattering source terms implicitly and exactly; the only approximation involves the transverse leakage terms. In conventional SNnodal methods, the transverse leakage terms and scattering source are both approximated. We solve the SGF-CN equations using the one-node block inversion iterative scheme, which uses the best available estimates for the node-entering fluxes to evaluate the node-exiting fluxes in the directions that constitute the incoming fluxes for the adjacent nodes as the equations are swept across the system. Finally, we give numerical results that illustrate the accuracy of the SGF-CN method for coarse-mesh calculations.

ISSN:0029-5639    

DOI:10.13182/NSE92-A23921

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractA discrete ordinates response matrix method is formulated with anisotropic scattering for the solution of neutron transport problems on massively parallel computers. The response matrix formulation eliminates iteration on the scattering source. The nodal matrices that result from the diamond-differenced equations are utilized in a factored form that minimizes memory requirements and significantly reduces the number of arithmetic operations required per node. The red-black solution algorithm utilizes massive parallelism by assigning each spatial node to one or more processors. The algorithm is accelerated by a synthetic method in which the low-order diffusion equations are also solved by massively parallel red-black iterations. The method is implemented on a 16K Connection Machine-2, and S8and S16solutions are obtained for fixed-source benchmark problems in x-y geometry.

ISSN:0029-5639    

DOI:10.13182/NSE92-A23922

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractThe one-group, steady-state neutron diffusion equation in two-dimensional Cartesian geometry is solved using the nodal integral method. The discrete variable equations comprise loosely coupled sets of equations representing the nodal balance of neutrons, as well as neutron current continuity along rows or columns of computational cells. An iterative algorithm that is more suitable for solving large problems concurrently is derived based on the decomposition of the spatial domain and is accelerated using successive overrelaxation. This algorithm is very well suited for parallel computers, especially since the spatial domain decomposition occurs naturally, so that the number of iterations required for convergence does not depend on the number of processors participating in the calculation. Implementation of our algorithm on the Intel iPSC/2 hypercube and Sequent Balance 8000 parallel computers is presented, and measured speedup and efficiency for test problems are reported. The results suggest that the efficiency of the hypercube quickly deteriorates when many processors are used, while the Sequent Balance retains very high efficiency for a comparable number of participating processors. This leads to the conjecture that message-passing parallel computers are not as well suited for this algorithm as shared-memory machines.

ISSN:0029-5639    

DOI:10.13182/NSE92-A23923

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractState feedback control provides many advantages, such as stabilization and improved transient response. However, when state feedback control is considered for spatial control of a nuclear reactor, it requires complete knowledge of the distributions of the system state variables. Also, if the reactor is in a transient, flux mapping systems that are based on steady-state conditions are not appropriate for an accurate representation of the operating state of the reactor.A method is described for reconstructing the measurable and unmeasurable state variables in a nuclear reactor from output measurement data, which can be used to generate input for a feedback control system or serve as a core observer (estimator) in a reactor transient. The method is based on a Luenberger-type observer theory that is extended to infinite-dimensional distributed parameter systems.The method was applied to a simple reactor model in one spatial dimension and one energy group with xenon dynamics that exhibited spatial oscillations. The resulting observer was tested by using model-based data for measurement output. The results show that the spatial distributions of iodine, xenon, and neutron flux are estimated very well by the observer using information from a finite number of sensors.

ISSN:0029-5639    

DOI:10.13182/NSE92-A23924

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractReleases of radionuclides and the production of aerosols during the molten core/concrete interaction (MCCI) phase of degraded core accidents in light water reactors are termed“ex-vessel releases.”The VANESA and METOXA codes were respectively developed by the U.S. Nuclear Regulatory Commission and the Industrial Degraded Core Rulemaking (IDCOR) program to quantify ex-vessel releases. Comparison of calculations by VANESA and METOXA (under identical initial and boundary conditions) show that except for niobium and strontium species, the predicted ex-vessel radionuclide release rates are within an order of magnitude of each other. In an actual application of these two codes to the source term quantification of severe accidents, the initial and boundary conditions for the calculations could be significantly different, as demonstrated in an analysis of an anticipated transient without scram accident sequence in a boiling water reactor. For the same amount of debris, the MCCI thermal-hydraulic results provided for METOXA from a DECOMP calculation tend to drive more radioactive material from the debris pool than those provided for VANESA from a CORCON/MOD2 calculation. The MAAP code, however, predicts that less mass is involved in the MCCI.

ISSN:0029-5639    

DOI:10.13182/NSE92-A23925

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractThe time-dependent spectrum of neutrons in the water-moderated Oak Ridge Electron Linear Accelerator (ORELA) target is calculated using a modified version of the MORSE multigroup Monte Carlo code with an analytic hydrogen scattering model. Distributions of effective neutron distance traversed in the target are estimated with a time- and energy-dependent algorithm from the leakage normal to the target face. The 10-eV to 20-MeV energy range is adequately represented in the MORSE code by the 174-group VITAMIN-E cross-section library with a P5expansion. An approximate representation of the ORELA positron source facility, recently installed near the target, is included in the calculations to determine any perturbations the positron source might create in the computed neutron distributions from the target. A series of coupled Monte Carlo calculations is performed from the target to the positron source and back to the target using a next-event estimation surface source for each step. The principal effect of the positron source is an increase in the distance for the lower energy neutron spectra, producing no real change in the distributions where the ORELA source is utilized for experiments. Different configurations for the target are investigated to simulate the placement of a shadow bar in the neutron beam. These beam configurations include neutrons escaping from (a) the central tantalum plates only, (b) the entire target with the tantalum plates blocked out, and (c) only a small area from the water. Comparisons of the current data with previous calculations having a less detailed model of the tantalum plates are satisfactory.

ISSN:0029-5639    

DOI:10.13182/NSE92-A23926

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor