摘要:

AbstractLet the lattice consist of an infinite uniform distribution of clusters in an (external) moderator, and let the cluster consist of a uniform distribution of absorber lumps in an (internal) moderator. The lattice is characterized by the parameters: cluster mean chord length, L; probability of neutrons leaving a cluster to collide in the external moderator prior to crossing a cluster,Γ; adjustable Bell factor for the clusters, A; lump mean chord length, I; probability of neturons leaving a lump (in an infinite cluster) to collide in the internal moderator prior to crossing a lump,γ; adjustable Bell factor for the lumps, a; internal moderator volume fraction in the cluster,υm; internal moderator macroscopic cross section,Σm. The flux in (or resonance integral of) the absorber lump is equivalent to the flux in (or resonance integral of) an infinite medium consisting of the lump material, homogenously mixed with a moderator of cross sectionΣe, given bywhereThe expression forΣeis quite general, the only restriction on the lattice structure being that a cluster contain many lumps. The factorβcan be termed the“double heterogeneity”factor abbreviated“doublet.”In the limit of an infinite single clusterβ→1, yielding the correct single heterogeneity expression forΣe.In the limit of small lump volume fractions, the expression forΣereduces to the expression of Goldstein, as derived from the work of Lane et al. Goldstein’s formulation was successfully compared with the experimental data of Lewis and Conolly. The WIMS formulation for a single cluster is almost equivalent to the above formulas with a difference that becomes significant only if the cluster contains a small number of lumps. The equivalence formulations by Tsuchihashi et al., as well as by Stamatelatos, yield results which are discrepant with those of the formulations discussed above and, therefore, have to be judged unsatisfactory.

ISSN:0029-5639    

DOI:10.13182/NSE82-A20082

出版商:Taylor&Francis    

年代:1982

数据来源: Taylor

摘要:

AbstractSpectral and spatial distributions of neutrons and gamma rays were measured in a simple cavity-duct configuration to observe profiles of cavity streaming. The change of the profiles of neutrons and gamma rays is examined by blocking source neutrons with a 32-cm-thick aluminum plug. The following observations resulted from the experiments:Fast neutrons of several million electron volts energy streamed through ducts.The neutron and gamma-ray spectra were similar except in the source neutron beam.The gamma rays were predominantly those arising from neutron capture in iron.The aluminum plug greatly decreased the fast neutron flux but had only a limited effect on the low energy neutron flux.The applicability of the albedo Monte Carlo calculational method to this problem was examined with the following conclusions:For ducts of small radius, the calculations overestimate the streaming because the albedo data were given for plane geometry.Low energy neutrons were underestimated by the calculation due to the neutron penetration through the cavity wall.

ISSN:0029-5639    

DOI:10.13182/NSE82-A20083

出版商:Taylor&Francis    

年代:1982

数据来源: Taylor

摘要:

AbstractThe contribution of bremsstrahlung, annihilation, and fluorescence radiation to the spectra of scattered gamma radiation and to the dose buildup factors has been investigated as a function of source energy, atomic number, and sample thickness over an energy range of 0.1 to 8 MeV. The computations were performed with the one-dimensional transport code ASFIT modified to take into account all the secondary radiations. The required mathematical formulation, along with representative results obtained for uranium, lead, iron, and water, typifying materials of very high, high, medium, and low atomic number, are presented and discussed.A noticeable effect of including bremsstrahlung sources is the general softening of the scattered radiation spectra inside the medium and at the exit. This effect is more pronounced in materials of high atomic number. The bremsstrahlung contribution is seen most prominently in the reflection spectra above 0.511 MeV, where the contribution from other processes is insignificant.The effect of annihilation radiation is significant in the region between 0.511 MeV and the K edge, below which the effects of fluorescence radiation overshadow all others. Peaks and discontinuities characteristic of single scatterings of these radiations are seen in the reflection spectra, gradually disappearing with depth in the medium. The effect of fluorescence on the dose buildup factor is spectacular for source energies close to the K edge and falls off rapidly thereafter. The impact of bremsstrahlung, on the other hand, steadily rises with source energy. The influence of annihilation radiation is comparatively modest and is significant only for systems of intermediate atomic numbers.

ISSN:0029-5639    

DOI:10.13182/NSE82-A20084

出版商:Taylor&Francis    

年代:1982

数据来源: Taylor

摘要:

AbstractThe absolute235U neutron-induced fission cross section has been measured at the U.S. National Bureau of Standards (NBS) 3-MV Van de Graaff Laboratory from 0.2- to 1.2-MeV neutron energy. The mass of the235U contained in a large volume multiplated fission ionization chamber was measured relative to the NBS fissionable isotope mass standards. Pulsed beam time-of-flight techniques were used with neutrons from the7Li(p, n)7Be reaction while the neutron flux was monitored with a large plastic scintillator whose efficiency was both calculated and measured with the associated-particle technique. The cross sections, which were measured with a typical uncertainty of 2.3%, are∼2% lower than the ENDF/B-V evaluation.

ISSN:0029-5639    

DOI:10.13182/NSE82-A20085

出版商:Taylor&Francis    

年代:1982

数据来源: Taylor

摘要:

AbstractOn the basis of standard nuclear evaporation theory, we calculate the prompt fission neutron spectrum N(E) as a function of both the fissioning nucleus and its excitation energy. To simulate the initial distribution of fission-fragment excitation energy and the subsequent cooling of the fragments as neutrons are emitted, we take the distribution of fission-fragment residual nuclear temperature to be triangular in shape, extending linearly from zero to a maximum value Tm. This maximum temperature is determined from the average energy release, the separation energy and kinetic energy of the neutron inducing fission, the total average fission-fragment kinetic energy, and the level density parameter of the Fermi gas model. The neutron energy spectrum for fixed residual nuclear temperature is integrated over this triangular distribution to obtain the neutron energy spectrum in the center-of-mass system of a given fission fragment, which is then transformed to the laboratory system. When the cross sectionσcfor the inverse process of compound nucleus formation is assumed constant, N(E) is the sum of a four-term closed expression involving the exponential integral and the incomplete gamma function for the light fragment and an analogous result for the heavy fragment. We also calculate N(E) by numerical integration for an energy-dependent cross sectionσcthat is obtained from an optical model; this shifts the peak in N(E) to somewhat lower neutron energy and changes the overall shape slightly. The spectra calculated for both a constant cross section and an energy-dependent cross section reproduce recent experimental data for several fissioning nuclei and excitation energies for a single choice of the nuclear level density parameter and without the use of any further adjustable parameters. However, the spectra calculated with an energy-dependent cross section agree somewhat better with the experimental data than do those calculated with a constant cross section. Our approach is also used to calculate, the average number of prompt neutrons per fission, as a function of excitation energy for several fissioning nuclei. At high excitation energy, where fission following the emission of one or more neutrons is possible, we take into account the effects of and competition between first-, second-, and third-chance fission when calculating both N(E) and. For ease of computation, we present finally an approximate way to simulate the energy dependence of the compound nucleus cross section through a slight readjustment of the value of the level density parameter.

ISSN:0029-5639    

DOI:10.13182/NSE82-5

出版商:Taylor&Francis    

年代:1982

数据来源: Taylor

摘要:

AbstractA method is presented for calculating the nodal flux distribution and the pin power distribution, as well as the effective multiplication, in a nuclear power reactor described by the one-dimensional, two-group diffusion equation.The method is based on the use of Green’s functions in a nodal reactor description, and it extends the work of previous authors by including burnup-induced heterogeneities and by calculating local pin power distributions from spatial flux distributions within the node obtained by piecewise polynomial interpolation. An advantage of the method is that one obtains power and exposure distributions at fine mesh points, while retaining the economy characteristic of solutions of the neutron diffusion equation in the nodal framework.In numerical calculations carried out on model problems, good agreement is achieved between the results of the extended nodal Green’s function method and those obtained using the CITATION finite difference code.

ISSN:0029-5639    

DOI:10.13182/NSE82-A20087

出版商:Taylor&Francis    

年代:1982

数据来源: Taylor

摘要:

AbstractA method is presented for“statistical screening”of input variables in a complex computer code. The object is to determine the“effective”or important input variables by estimating the relative magnitudes of their associated sensitivity coefficients. This is accomplished by performing a numerical experiment consisting of a relatively small number of computer runs with the code followed by a statistical analysis of the results. A formula for estimating the sensitivity coefficients is derived. Reference is made to an earlier work in which the method was applied to a complex reactor code with good results.

ISSN:0029-5639    

DOI:10.13182/NSE82-A20088

出版商:Taylor&Francis    

年代:1982

数据来源: Taylor

摘要:

AbstractIn this investigation, variance measurements of the secondary electron emission current from an evacuated chamber have been performed and a formula to calculatein these experiments has been derived. The value obtained foris within 50% ofobtained with a gas detector and a factor of 2 greater than calculated values. The mean track length and the Vtrackterm were also calculated for this irradiation geometry, Vtrackbeing of the same order of magnitude as Vtrackfor a sphere.

ISSN:0029-5639    

DOI:10.13182/NSE82-A20089

出版商:Taylor&Francis    

年代:1982

数据来源: Taylor

摘要:

AbstractA short description of the reactor dynamics program KINE is given. The KINE code is a one-dimensional code solving the time-dependent neutron diffusion equations in a two-group representation, taking into account six groups of delayed neutron precursors, one heating channel composed of fuel, canning, and a coolant region in which boiling may occur. Special attention is given to radial averaging of coolant density and fuel temperature in a core in which partial boiling can exist. The method of solution is a modified backward extrapolation procedure. Stationary starting conditions are achieved by adjustment of control rod position or liquid poison concentration.

ISSN:0029-5639    

DOI:10.13182/NSE82-A20090

出版商:Taylor&Francis    

年代:1982

数据来源: Taylor

摘要:

AbstractA set of pressure data recorded from the primary coolant system of a pressurized water reactor under normal operation is examined. For pressure sensors under study, it is found that the range of sampling intervals adequate to abstract pressure dynamics is∼0.2 to 0.8 ms. The estimated time constant of the pressure sensor for operating data is 1.83 ms.

ISSN:0029-5639    

DOI:10.13182/NSE82-A20091

出版商:Taylor&Francis    

年代:1982

数据来源: Taylor