摘要:

AbstractCalculations of fission neutron spectra and fission spectrum integrals for six fissioning systems are reported. The systems studied are thermal-neutron-induced fission of233U,235U, and239Pu; 2-MeV neutron fission of232Th and238U; and spontaneous fission of252Cf The Madland-Nix formalism is used, but is extended to take account of the spin of the fission fragment. Also, more recent values for the average energy release in fission are used. The results of the fission neutron spectrum calculations are given in tabular and graphical form for applied purposes and are compared with experimental data. Similar deviations from a Maxwellian spectrum are seen for each system. The effect on the fission neutron spectra of233U,235U, and239Pu from inclusion of alternative values for the average total fragment kinetic energy is shown.

ISSN:0029-5639    

DOI:10.13182/NSE89-A23637

出版商:Taylor&Francis    

年代:1989

数据来源: Taylor

摘要:

AbstractA method based on a forward dynamic programming technique is applied to load-following control of a boiling water reactor. The control strategy obtained is optimal and satisfies operation constraints. A coarse-mesh, one-dimensional model using the two-group diffusion theory with Doppler, void, and xenon feedbacks is developed to reduce computer time. The control rods are assumed to be fixed during load maneuvers, and variations in core power are accomplished through core flow.An off-line daily load-following analysis needs∼2000 CPU s on a PRIME 9950 computer. With some relaxation, computation time can be reduced to several hundred seconds. Thus, an on-line calculation that leads to an approximate closed-loop control is feasible.

ISSN:0029-5639    

DOI:10.13182/NSE89-A23638

出版商:Taylor&Francis    

年代:1989

数据来源: Taylor

摘要:

AbstractVariational nodal methods for neutron transport are modified to reduce the angular coupling between spatial nodes without a commensurate loss of accuracy. In both one and two dimensions, the interface conditions of the variational principle allow near Pnaccuracy to be obtained with only Pn−2interface coupling. As a result, the dimension of the nodal response matrix is reduced by a factor of 2, and the number of arithmetic operations required for solution by a factor of 4. In the small spatial mesh limit, the resulting Pn, n−2approximation retains accuracy near the Pnapproximation used within the node rather than reverting to the Pn−2interface approximation. Two-dimensional P3,1transport calculations demonstrate that the variational nodal approximations are not subject to the flux depression suffered by other interface current nodal transport methods in problems dominated by streaming diagonal to the coordinate directions.

ISSN:0029-5639    

DOI:10.13182/NSE89-A23639

出版商:Taylor&Francis    

年代:1989

数据来源: Taylor

摘要:

AbstractAn expert system for diagnosing operational transients in a nuclear power plant is discussed. Hypothesis and test is used as the problem-solving strategy with hypotheses generated by an expert system that monitors the plant for patterns of data symptomatic of known failure modes. Fuzzy logic is employed as the inferencing mechanism with two complementary implication schemes to handle scenarios involving competing failures. Hypothesis testing is performed by simulating the behavior of faulted components using numerical models. A filter has been developed for systematically adjusting key model parameters in an attempt to obtain agreement between simulations and actual plant data. Pattern recognition is employed as a decision analysis technique for choosing among several hypotheses based on simulation results. An artificial intelligence framework based on a critical functions approach is used to deal with the complexity of a nuclear plant. A prototype system for diagnosing transients in the reactor coolant system of a pressurized water reactor has been developed to test the algorithms described here. Results are presented for the diagnosis of data from the Three Mile Island Unit 2 loss-of-feedwater/small-break loss-of-coolant accident.

ISSN:0029-5639    

DOI:10.13182/NSE89-A23640

出版商:Taylor&Francis    

年代:1989

数据来源: Taylor

摘要:

AbstractTime-dependent integral transport equation single-collision kernels for one-dimensional geometries corresponding to the steady-state single-collision kernels found in the available literature have been calculated by making use of the Laplace transform technique, simple geometric transformation relationships, and point kernel integrations. Using the convolution theorem, the time-dependent scalar flux is obtained by convoluting the single-collision kernel with the time-dependent source. Using the multiple collision formulation of the integral transport solution, isotropic sources that are delta distributions in time are considered in several examples. Analytical solutions for the uncollided and first-collided scalar fluxes are obtained for a boundary source having an isotropic angular distribution directed into a semi-infinite medium and into a slab of thickness b and for a point source at the origin of an infinite medium and finite sphere of radius a. A closed form solution is obtained for the simple problem of uniformly distributed sources within an infinite medium.

ISSN:0029-5639    

DOI:10.13182/NSE89-A23641

出版商:Taylor&Francis    

年代:1989

数据来源: Taylor

摘要:

AbstractAn expression is developed in diffusion theory for estimating the first-order change in a ratio of linear functionals due to a perturbation in the current-to-flux ratio boundary condition of a system. One numerical example is given. Additionally, it is shown that the perturbation formalism may be used in the spatial homogenization process to account for interassembly effects such as gross flux tilts and spectral effects caused by neighboring assemblies in the reactor core.

ISSN:0029-5639    

DOI:10.13182/NSE89-3

出版商:Taylor&Francis    

年代:1989

数据来源: Taylor

摘要:

AbstractEmpirical formulas are derived that describe neutron flux distributions in straight and bent cylindrical steel-walled ducts in symmetrical and asymmetrical geometries. Asymmetry is defined by the slant angle between the duct axis and a line passing through the center of the duct mouth and source. Streaming neutrons are divided into direct, albedo, and penetration components. The first two components are described by a function of the axial distance in units of the square root of the line-of-sight area. The last component is described by a function of the neutron flux distribution in the shield in the absence of the duct. Formulas are applicable to thermal, epithermal, intermediate, and fast neutrons, respectively, in the following range: (a) duct diameter is 5 to 20 cm, (b) duct length is up to 240 cm, (c) slant angle is 0 to 90 deg, (d) steel wall thickness is 0 to 1.0 cm, (e) bent angle is 45 to 90 deg, and (f) surrounding medium of the duct is water or ordinary concrete. Calculations by formulas agree with experimental data, in general, within an accuracy of±30%.

ISSN:0029-5639    

DOI:10.13182/NSE89-A23643

出版商:Taylor&Francis    

年代:1989

数据来源: Taylor