摘要:

AbstractA numerical algorithm with parallelism was used to solve the two-group, multidimensional neutron diffusion equations on computers characterized by shared memory, vector pipeline, and multi-CPU architecture features. Specifically, solutions were obtained on the Cray X/MP-48, the IBM-3090 with vector facilities, and the FPS-164. The material-centered mesh finite difference method approximation and outer-inner iteration method were employed. Parallelism was introduced in the inner iterations using the cyclic line successive overrelaxation iterative method and solving in parallel across lines. The outer iterations were completed using the Chebyshev semi-iterative method that allows parallelism to be introduced in both space and energy groups. For the three-dimensional model, power, soluble boron, and transient fission product feedbacks were included. Concentrating on the pressurized water reactor (PWR), the thermal-hydraulic calculation of moderator density assumed single-phase flow and a closed flow channel, allowing parallelism to be introduced in the solution across the radial plane.Using a pinwise detail, quarter-core model of a typical PWR in cycle 1, for the two-dimensional model without feedback the measured million floating point operations per second (MFLOPS)/vector speedups were 83/11.7, 18/2.2, and 2.4/5.6 on the Cray, IBM, and FPS without multitasking, respectively. Lower performance was observed with a coarser mesh, i.e., shorter vector length, due to vector pipeline start-up. For an 18×18×30 (x-y-z) three-dimensional model with feedback of the same core, MFLOPS/vector speedups of∼61/6.7 and an execution time of 0.8 CPU seconds on the Cray without multitasking were measured. Finally, using two CPUs and the vector pipelines of the Cray, a multitasking efficiency of 81% was noted for the three-dimensional model.

ISSN:0029-5639    

DOI:10.13182/NSE89-A23609

出版商:Taylor&Francis    

年代:1989

数据来源: Taylor

摘要:

AbstractA new ray-tracing method for the calculation of collision probabilities within arbitrary three-dimensional geometries has been developed. This method is used to discretize the neutron transport equation for heterogeneous rectangular cells containing zones of mixed cylindrical and rectangular geometry. For multicell applications, the interface current (IC) method provides the coupling between cells. The solution to the IC equations over multicell domains consisting of rectangular three-dimensional cells is improved by using an alternate direction implicit iteration scheme with variational acceleration. Results include comparisons of this technique with SHETAN for simple geometries and the analysis of a three-dimensional extension of a two-dimensional 15×15 pressurized water reactor benchmark problem.

ISSN:0029-5639    

DOI:10.13182/NSE101-217

出版商:Taylor&Francis    

年代:1989

数据来源: Taylor

摘要:

AbstractA change in the atomic number density of an element may induce a shift in the neutron spectrum, resulting in a change in all the group constants in a multigroup representation. This is referred to as the spectral shift effect. The arbitrariness inherent in the concept of reactivity is investigated by taking the spectral shift effect into account. To this end, the reactor period of a transient resulting from a spectral shift is investigated, using first-order perturbation theory. It is then shown that the result leads to a new choice for the shape function in the general formulation of the reactor dynamical parameters such as reactivity. Using a new scheme, numerical calculations are made for RBMK-1000 and light water reactors (LWRs). It is found that for LWRs the void coefficient is always negative, while for RBMK-1000 it tends to be positive as the burnup proceeds.

ISSN:0029-5639    

DOI:10.13182/NSE89-A23611

出版商:Taylor&Francis    

年代:1989

数据来源: Taylor

摘要:

AbstractA nonlinear reactor model is developed taking into account several feedback effects, such as moderator and fuel temperatures, xenon absorption, and soluble boron concentration, through energy balance relations in the core. The resulting equation belongs to a class of nonlinear boundary value problems, and it is shown through bifurcation theory that there may exist multiple steady-state solutions for a range of parameters that correspond to various design and operating conditions. Solutions are obtained numerically for ranges of the parameters by the arc-length continuation method in combination with Newton’s method. Stability analysis is also applied to each solution to investigate whether the solution is stable or not. When the stable and unstable regions of the steady-state solutions are plotted for a wide range of the parameters, we can choose a range of the reactor design and operating conditions such that the reactor does not encounter unstable situations.

ISSN:0029-5639    

DOI:10.13182/NSE89-A23612

出版商:Taylor&Francis    

年代:1989

数据来源: Taylor

摘要:

AbstractA theoretical treatment is developed for the deposition and transport of an aerosol in a multicompartment system in which there exists a pressure-induced gas flow. Based on a solution of the relevant gas equations, the aerosol equation is first formulated and then solved numerically by both discretization and moments techniques. The former method is more accurate, but the complex nature of the problem means that the computing time required can be prohibitive, especially when the number of compartments is large. The moments technique, based on a gamma or lognormal distribution, requires substantially less computing time, and to estimate its accuracy, a validation comparison has been made with the discretization method. The technique was then applied to two multicompartment accident situations. Results show that the moments method based on the gamma distribution is significantly more accurate than the lognormal-based one and is also in close agreement with the results from the AEROSIM code.

ISSN:0029-5639    

DOI:10.13182/NSE89-A23613

出版商:Taylor&Francis    

年代:1989

数据来源: Taylor

摘要:

AbstractThe dynamic behavior of aerosols is of considerable interest in nuclear reactor source term studies. Because of the intractable nature of the dynamics, this behavior is studied through solutions of a spatially homogenized equation. There has been a need, however, to understand the effect of the homogenization process on the calculated aerosol distributions. To provide insight into the nature of the approximation and the accuracy of the results calculated with the homogenized (averaged) equations, some typical aerosol distribution problems are solved both with the spatially dependent and the homogenized versions of the aerosol dynamic equations. Comparisons of results show that while in some instances homogenization can be quite useful, there are realistic circumstances where it can lead to substantial deviations from accurate results as obtained by the equation that allows for spatial dependence of aerosol distribution.

ISSN:0029-5639    

DOI:10.13182/NSE101-269

出版商:Taylor&Francis    

年代:1989

数据来源: Taylor

摘要:

AbstractThe gas-dynamic response of argon to fission-fragment energy deposition is simulated, for the first time explicitly including the coupling between the gas density, which is spatially and temporally varying, and the power density. In simulations of three experiments with different initial fill pressures of argon, good agreement was found between calculated and observed pressure rises, after the experimental pressure rise data from one case were used as a calibration. However, in each case, the calculated thermal energy deposition corresponding to the experimental pressure data was about half the fission-fragment kinetic energy release into the gas predicted by neutron and fission-fragment transport calculations. Also, the experimental pressure data exhibited a decay not seen in the simulations, which did not incorporate an energy-loss mechanism.

ISSN:0029-5639    

DOI:10.13182/NSE89-A23615

出版商:Taylor&Francis    

年代:1989

数据来源: Taylor

摘要:

AbstractIt is a common practice in ENDF/B to represent neutron cross sections in the unresolved resonance region by specifying the average values and distribution laws of resonance parameters. This formalism allows the calculation of resonance self-shielding and of the variation of resonance selfshielding with temperature, two important reactor parameters. For many applications it is necessary to estimate the uncertainties in these model average resonance parameters. A possible approach to derive such uncertainties is described, using as an example the ENDF/B-V representation of238U in the range from 4 to 45 keV.

ISSN:0029-5639    

DOI:10.13182/NSE89-A23616

出版商:Taylor&Francis    

年代:1989

数据来源: Taylor

摘要:

AbstractThe thermal neutron induced fission cross section of243Am has been measured using a very intense, very clean thermal neutron beam at the Institut Laue-Langevin, Grenoble. Great care has been taken to ensure the purity of the sample material. As a result of these measurements, a value of (74±4) mb has been obtained for the243Am(nth, f) cross section.

ISSN:0029-5639    

DOI:10.13182/NSE89-A23617

出版商:Taylor&Francis    

年代:1989

数据来源: Taylor

摘要:

AbstractUsing the NJOY nuclear data processing system, three multigroup MATXS-formatted nuclear data libraries were generated based on the European data files JEF-1 and EFF-1. After processing with TRAMIX, TRANSX, or TRANSX-CTR, these libraries can be read into most transport and diffusion codes.For the neutron analysis of gas-cooled or water-moderated thermal reactor systems (including high converter pressurized water reactors), a 70-group WIMS-BOXER structured library was generated. A general-purpose fine-group library in 308 groups is provided for thermal as well as for fast reactor systems. A coupled 175 neutron/42 photon-group library in VITAMIN-J structure was created for the analysis of shielding problems and fusion blanket design. The three MATXS files can be requested from the Nuclear Energy Agency Data Bank.

ISSN:0029-5639    

DOI:10.13182/NSE89-A23618

出版商:Taylor&Francis    

年代:1989

数据来源: Taylor