摘要:

AbstractAn upwind numerical method for a six equation two-fluid model is presented based on a linearized Riemann solver. The construction of this approximate Riemann solver uses an extension of Roe’s method that has been successfully used to solve gas dynamics equations in aerodynamics problems. As far as the two-fluid model is hyperbolic, this numerical method seems very efficient for the numerical solution of two-phase flow problems. In practice, this new numerical method has proved to be stable on nonstaggered grids and capable of generating accurate nonoscillating solutions for two-phase flow calculations. The scheme was applied both to shock tube problems and to standard tests for two-fluid codes.

ISSN:0029-5639    

DOI:10.13182/NSE96-A24180

出版商:Taylor&Francis    

年代:1996

数据来源: Taylor

摘要:

AbstractA Godunov scheme is proposed for the simulation of impact problems and detonations where nonstationary fronts and interfaces are tracked as boundaries of subregions that move in time. In each subregion and at each time step, a new grid is created by the use of boundary-fitted coordinates. The numerical method is based on a finite-volume approach in the space-time domain, and the fluxes are calculated using the solution of Riemann problems. Numerical results are shown for several impact and detonation problems, showing the efficiency of this approach.

ISSN:0029-5639    

DOI:10.13182/NSE96-A24181

出版商:Taylor&Francis    

年代:1996

数据来源: Taylor

摘要:

AbstractThe Institute of Nuclear Energy Research (INER) integral system test (IIST) facility is a reduced-height, reduced-pressure test facility constructed at INER that is used to simulate the thermal hydraulics of the Maanshan nuclear power plant (NPP). A small-scaled facility is not capable of simulating all the physical phenomena of an NPP because the behavior of an NPP during accidents is very complicated. Proper scaling then plays an important role in the design of a test facility to ensure the usefulness and applicability of experimental data obtained from a small-scaled facility. However, distortions caused by necessary compromises in the design and construction of a small-scaled test facility exist. The analysis here evaluates whether the inherent distortions in the IIST facility will distort the thermal-hydraulic behaviors of a natural-circulation experiment and influence the usefulness and applicability of the experimental data. Based on the current calculations, the IIST experimental results are found to be partially distorted. Appropriate consideration of and correction for these distortion effects are needed before the results of the IIST natural-circulation experiments can be used to reliably investigate the Maanshan NPP behavior expected by way of an appropriate scale-up procedure.

ISSN:0029-5639    

DOI:10.13182/NSE96-A24182

出版商:Taylor&Francis    

年代:1996

数据来源: Taylor

摘要:

AbstractIn solving few-group neutron kinetic equations in multidimensions, one must select time step sizes as a function of time such that the temporal truncation error introduced by the discrete time derivative approximation is limited to ensure the desired fidelity. When using the Euler backward finite difference to approximate the first derivative of the flux—a popular approximation because it ensures numerical stability—the truncation error is know to be O(Δt2) and proportional to the second derivative. By employment of the double-time-step-size technique, modified to reduce the frequency that double-time-step-size solutions are required, an estimate of the second derivative can be obtained, leading to an efficient computational algorithm for determining the near-optimum time-step-size sequence to ensure the desired fidelity.

ISSN:0029-5639    

DOI:10.13182/NSE96-A24183

出版商:Taylor&Francis    

年代:1996

数据来源: Taylor

摘要:

AbstractImprovements of gamma-ray transport calculations in Sncodes aim at taking into account the bound-electron effect of Compton scattering (incoherent), coherent scattering (Rayleigh), and secondary sources of bremsstrahlung and fluorescence. A computation scheme was developed to take into account these phenomena by modifying the angular and energy transfer matrices, and no modification in the transport code has been made. The incoherent and coherent scatterings as well as the fluorescence sources can be strictly treated by the transfer matrix change. For bremsstrahlung sources, this is possible if we can neglect the charged particles path as they pass through the matter (electrons and positrons) and is applicable for the energy range of interest for us (below 10 MeV). These improvements have been reported on the kernel attenuation codes by the calculation of new buildup factors. The gamma-ray buildup factors have been carried out for 25 natural elements up to 30 mean free paths in the energy range between 15 keV and 10 MeV.

ISSN:0029-5639    

DOI:10.13182/NSE96-A24184

出版商:Taylor&Francis    

年代:1996

数据来源: Taylor

摘要:

AbstractAlthough resonance neutron captures for238U in water-moderated lattices are known to occur near moderator-fuel interfaces, the sharply attenuated spatial captures here have not been calculated by multigroup transport or Monte Carlo methods. Advances in computer speed and capacity have restored interest in applying Monte Carlo methods to evaluate spatial resonance captures in fueled lattices. Recently published studies have placed complete reliance on the ostensible precision of the Monte Carlo approach without auxiliary confirmation that resonance processes were followed adequately or that the Monte Carlo method was applied appropriately. Other methods of analysis that have evolved from early resonance integral theory have provided a basis for an alternative approach to determine radial resonance captures in fuel rods. A generalized method has been formulated and confirmed by comparison with published experiments of high spatial resolution for radial resonance captures in metallic uranium rods. The same analytical method has been applied to uranium-oxide fuels. The generalized method defined a spatial effective resonance cross section that is a continuous function of distance from the moderator-fuel interface and enables direct calculation of precise radial resonance capture distributions in fuel rods. This generalized method is used as a reference for comparison with two recent independent studies that have employed different Monte Carlo codes and cross-section libraries. The Monte Carlo studies have been found to undercount reference radial resonance captures in the moderator-fuel interface region. The steep radial capture gradients within 0.50 mm of the interface account for the majority of resonance captures and take place where Monte Carlo spatial resolution is poor and the effects of resonance peaks on neutron flux are large. Inconsistencies in the Monte Carlo application or in howpointwise cross-section libraries are sampled may exist. It is shown that refined Monte Carlo solutions with improved spatial resolution would not asymptotically approach the reference spatial capture distributions. It is suspected that the resolved resonance peak and off peak cross sections may not be represented or accounted for appropriately in the Monte Carlo calculations and should be reviewed. If these inconsistencies were cleared up, use of the generalized method might very well challenge the need to perform further Monte Carlo studies of radial resonance captures for isolated uranium-oxide fuel rods.

ISSN:0029-5639    

DOI:10.13182/NSE96-A24185

出版商:Taylor&Francis    

年代:1996

数据来源: Taylor

摘要:

AbstractThe Accident Sequence Precursor (ASP) program methods and results are reviewed. Generally, it is concluded that the ASP program is useful and that continued methods improvement efforts currently under way should be continued. More care is needed in the interpretation of results. Alternative methods and treatments for the analysis of operational events and the use of ASP results are determined.

ISSN:0029-5639    

DOI:10.13182/NSE96-A24186

出版商:Taylor&Francis    

年代:1996

数据来源: Taylor

摘要:

AbstractThe trend toward increased reliance on passive features for power reactor safety makes it important to obtain the characteristics of the reactor system from measurements on the system. A method is described for solving for the delayed neutron parameters in a liquid-metal power reactor by fitting an analytic solution of the point-kinetics equations to the flux die-away from a dropped rod in an initially critical core. The method includes treatment of those conditions found in a power reactor that depart from those in a critical assembly experiment. These include a comparatively long rod drop time and a detector signal that instead of providing an integrated count rate is a sampled data signal proportional to the instantaneous fission power. The delayed neutron parameter values calculated from a rod drop experiment in the Experimental Breeder Reactor II are in agreement with values calculated using first principles and knowledge of core material composition and nuclear cross sections.

ISSN:0029-5639    

DOI:10.13182/NSE96-A24187

出版商:Taylor&Francis    

年代:1996

数据来源: Taylor

摘要:

AbstractGamma-ray production data of carbon, fluorine, aluminum, silicon, titanium, chromium, manganese, cobalt, copper, niobium, molybdenum, tungsten, and lead in JENDL-3.2 and FENDL/E-1.0 induced by 14-MeV neutrons are tested through benchmark analyses of leakage gamma-ray spectrum measurements conducted at the OKTAVIAN deuterium-tritium neutron source facility. The MCNP transport code is used along with the flagging method for detailed analyses of the spectra. As a result, several moderate problems are pointed out for secondary gamma-ray data of titanium, chromium, manganese, and niobium in JENDL-3.2 and for the data of titanium, chromium, manganese, cobalt, niobium, and lead in FENDL/E-1.0. Because no fatal errors are found, however, secondary gamma-ray data for the 13 elements in both libraries are reasonably well validated through these benchmark tests as far as 14-MeV neutron incidence is concerned.

ISSN:0029-5639    

DOI:10.13182/NSE96-A24188

出版商:Taylor&Francis    

年代:1996

数据来源: Taylor

摘要:

AbstractThe line-beam response function, needed for skyshine analyses based on the integral linebeam method, was evaluated with the MCNP Monte Carlo code for photon energies from 20 to 100 MeV and for source-to-detector distances out to 1000 m. These results are compared with pointkernel results, and the effects of bremsstrahlung and positron transport in the air are found to be important in this energy range. The three-parameter empirical formula used in the integral line-beam skyshine method was fit to the MCNP results, and values of these parameters are reported for various source energies and angles.

ISSN:0029-5639    

DOI:10.13182/NSE96-A24189

出版商:Taylor&Francis    

年代:1996

数据来源: Taylor