摘要:

AbstractFission-gas release from pyrolytic-carbon-coated uranium carbide particles was studied as part of a fuel-development program for gas-cooled reactors. The particles were contained in a test capsule between concentric cylinders of porous graphite and were heated by fission heat. A small flow of helium was used to sweep the fission gas from the test capsule.Uranium carbide particles coated with three types of pyrolytic carbon (laminar, columnar, and duplex), as well as uncoated uranium carbide particles, were irradiated at temperatures up to 1800 F. The steady-state fission-gas release rates were measured as a function of temperature and burnup. All three coating types greatly reduced the fission-gas release rate from uranium carbide particles; the duplex coating was much better than the laminar or the columnar coatings.Post-irradiation examination revealed less than 0.1% broken coatings for the duplex coating. A radiation-induced reaction zone was observed at the fuel/coating interface for all three types. A correlation was made between the number of broken coatings and fission-gas release rate.

ISSN:0029-5639    

DOI:10.13182/NSE64-A20051

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor

摘要:

AbstractExperiments have been performed in-pile, using the Transient Reactor Test Facility, to study the meltdown behavior under transient heating of metallic Experimental Breeder Reactor-II fuel elements contained in stagnant sodium. Threshold of failure, modes of failure, and post-experiment distribution of fuel were obtained for a range of experimental conditions including uniform axial power and axial-power profile shaped to approximate a typical power profile of a fast-reactor core. Samples were exposed both with sodium initially at saturation conditions, and with sodium pressurized to inhibit boiling. Although the presence of stagnant sodium was found to modify qualitatively the results found previously for dry EBR-II samples, the changes were not great, and results were consistent with those for dry elements.

ISSN:0029-5639    

DOI:10.13182/NSE64-A20052

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor

摘要:

AbstractMany nuclear reactors are constructed of arrays of parallel channels. In order to carry out heat-transfer and flow-redistribution calculations for such arrays, proper boundary conditions must be assigned. These boundary conditions are not always obvious.In this paper, a method of formulating boundary conditions is discussed in which the stagnation streamline is used to define fictitious channel extensions upstream and downstream of the original parallel-channel configuration. This procedure is equivalent to defining a new parallel-channel configuration for which the boundary conditions are more clearly defined.A comparison of the calculated hydraulic parameters with the associated experimental results shows that the method is essentially correct.

ISSN:0029-5639    

DOI:10.13182/NSE64-A20053

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor

摘要:

AbstractA set of one-group space-dependent neutron kinetics equations for reactor cores with spatially variable moderator density is developed. The solution to this set of differential equations is obtained numerically using an IBM-7094 digital computer. Employing the variational technique of von Neumann, a numerical stability criterion for space-dependent neutron kinetics equations is established. The present analysis is useful in the determination of the core open-loop response as well as the reactor system transient behavior. The open-loop response of a typical boiling-water reactor core for several values of step change in reactivity was determined using the present analysis. These are shown to be in agreement with the results of the classical space-independent neutron kinetics equations. The open-loop characteristic of the reactor core due to a step change in density distribution is also presented. The main distinguishing feature of the present study is the ability to determine the open-loop response due to disturbances (such as a series of successive step changes in density distribution) for which the classical space-independent approach provides no solution. Characteristics of this type are necessary in the dynamic analysis of boiling-water reactors where the system density distribution varies in time and space. A simple approximate method for the solution of the space-dependent neutron kinetics equations is also presented.

ISSN:0029-5639    

DOI:10.13182/NSE64-A20054

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor

摘要:

AbstractThe dependence of the decay time constant of a thermalized neutron pulse in H2O has been calculated both as a function of buckling and of temperature for the range of temperatures between 23 C and 300 C. Fair agreement between results for two moderator models and experiment has been found for the dependence of the diffusion coefficient on temperature. For higher coefficients in the buckling expansion the agreement is poorer.A new iterative method applicable to any moderator model has been used for the solution of the eigenvalue problem. This method is suited to high-order approximations to the transport equation, aP11approximation having been used in the present calculations. Convergence is rapid. An advantage is that the diffusion-cooled neutron fluxes are given accurately; these are presented and discussed.

ISSN:0029-5639    

DOI:10.13182/NSE64-A20055

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor

摘要:

AbstractAn iterative approach to the reactor burnup problem is developed on the basis of analytical solutions for the variable-coefficient burnup equations. The time dependence of the depletion matrices,A(t), is approximated by a polynomial representation. The number of basic time points for which spatial-diffusion calculations during burnup are required is determined only by the order of approximation necessary to give a reasonably good fit for the time dependence ofA(t). Usually a low-order approximation is sufficient, so the number of diffusion calculations is reduced to a minimum. The method is applicable both to survey-type calculations and to detailed reactor-burnup studies.A comparison of some results obtained with the method described in this paper and with standard calculational methods is given for a typical example. The results show the rapid convergence and accuracy of the proposed procedure.

ISSN:0029-5639    

DOI:10.13182/NSE64-A20056

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor

摘要:

AbstractAn iterative method has been developed for the solution of neutron-transport problems. The method is formulated within the framework of a spherical-harmonics method. The method is developed first for one-group problems, then for more general velocity-dependent problems. The method is illustrated by application to the Milne problem for the one-group case and also to a velocity-dependent variation of the Milne problem. Good accuracy is obtained for both cases.

ISSN:0029-5639    

DOI:10.13182/NSE64-A20057

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor

摘要:

AbstractNeutron-spectrum measurements in several moderators have been made with geometrical conditions approximating an infinite medium. The moderators studied were water, polyethylene and benzene, and the absorbers dispersed in the moderators were boron, samarium, erbium and gadolinium. The measurements are compared with predictions of the neutron spectra utilizing scattering models which take into consideration chemical binding. Measurements of scalar and angular neutron-energy spectra were made in a slab geometry water-moderated multiplying assembly at room temperature. The measured spectra are compared with DSN transport-theory calculations utilizing the Nelkin bound-hydrogen scattering model for water.

ISSN:0029-5639    

DOI:10.13182/NSE64-A20058

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor

ISSN:0029-5639    

DOI:10.13182/NSE64-A20060

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor

ISSN:0029-5639    

DOI:10.13182/NSE64-A20061

出版商:Taylor&Francis    

年代:1964

数据来源: Taylor