摘要:

AbstractFor nodal light water reactor calculations, the reconstruction of local multigroup neutron flux and current distributions within a node is discussed. Interpolation schemes that exploit the typical behavior of multigroup neutron spectra are proposed. These spectral interpolation schemes will be seen to be more accurate than the usual polynomial approaches. They reduce the error of pin power interpolation to that of fast flux interpolation. Thus the accuracy of one node per assembly calculations is considerably enhanced. The improvement is demonstrated by several benchmark problems.

ISSN:0029-5639    

DOI:10.13182/NSE85-A27435

出版商:Taylor&Francis    

年代:1985

数据来源: Taylor

摘要:

AbstractNodal diffusion and transport methods are formulated variationally in terms of the even-parity form of the neutron transport equation and applied to problems in X-Y geometry. The resulting functional guarantees the satisfaction of nodal balance, regardless of the form of the space-angle trial function within the node or on its boundaries. Deletion of X-Y cross terms from the within-node flux approximations yields equations that are strikingly similar to conventional diffusion nodal methods; inclusion of the terms obviatesad hocapproximations to the transverse leakage. Transport and diffusion nodal methods differ only in the angular basis functions. In both cases the equations are first solved for partial current moments along nodal interfaces. Subsequently, the detailed flux distribution and the node-averaged scalar flux values are obtained from the spatial trial functions. Results are given for fixed-source two-dimensional problems in the P1and P3approximations. Code vectorization and generalization to three dimensions are discussed.

ISSN:0029-5639    

DOI:10.13182/NSE85-A27436

出版商:Taylor&Francis    

年代:1985

数据来源: Taylor

摘要:

AbstractA deuterium-tritium neutron source is amplified when emitted into a body of material with appreciable (n,2n), (n,3n), and (n,f) cross sections. This amplification is described by a simple theory, approximating the strict integral transport description of the process. The distribution of neutrons in energy, from 14 MeV down to the (n,2n) threshold, is approximated by a generalized slowing down equation, which is similar in form to the infinite medium slowing down equation, and with average collision probabilities taking up the role of scattering fractions.Following a few collisions, the collision source spatial distribution resembles the fundamental mode flux distribution of a critical reactor. The average collision probability for such a source is, in diffusion theory,∑tr/(∑tr+ DB2), where B2is the geometrical buckling of the system. This yields an expression of the form (αx+βx2)/(l +αx +βx2) for the average collision probability, where x is a representative optical thickness of the system. It has been shown by numerical means that this form for the average collision probability is generally true for centrally peaked sources in variously shaped bare bodies of any optical thickness.

ISSN:0029-5639    

DOI:10.13182/NSE85-A27437

出版商:Taylor&Francis    

年代:1985

数据来源: Taylor

摘要:

AbstractThe energy distribution of neutrons, from the source energy of 14 MeV down to the (n,2n) threshold, can be approximated by effective one-group cross sections for such high and medium mass number elements as lead, copper, zirconium, and iron. The same is true for238U, when the fast fission factorϵis applied in a special manner to account for the added multiplication below the (n,2n) threshold. Two groups are required to obtain a reasonable description of the amplification in beryllium.The scheme enables a very accurate determination ofσn,2n+ 2σn,3n+ (vf–1)σfat the source energy from measurements of total multiplications. If total leakages above the (n,2n) threshold are also measured, then the hardness of the secondaries spectra can be estimated.The one-group theory applied is based on a previous derivation. The accuracy of the theory is ascertained by energy space detailed transport calculations performed for beryllium, copper, zirconium, iron, lead, and238U.

ISSN:0029-5639    

DOI:10.13182/NSE85-A27438

出版商:Taylor&Francis    

年代:1985

数据来源: Taylor

摘要:

AbstractAluminum samples, together with sets of 12 flux monitor foils having different reaction thresholds, were irradiated in 6 different deuteron/beryllium neutron fields (Ed= 17.5 to 30 MeV). The shapes of the neutron spectra were determined by spectrum unfolding, using the known excitation functions of the monitor nuclides and their measured activities. In a second calculational step, the excitation function for the f(n, t)+(n, tn)] process on27Al was obtained from the neutron flux distributions and the measured tritium activities. At both calculational stages the iterative code SAND-II and the generalized least-squares unfolding code were applied, the latter yielding additionally the error covariance matrix. The excitation function thus obtained has a maximum cross-section value of∼8 mb at 25.5 MeV and compares well with the values obtained using monoenergetic neutrons in the region of 14 to 19 MeV.

ISSN:0029-5639    

DOI:10.13182/NSE85-A27439

出版商:Taylor&Francis    

年代:1985

数据来源: Taylor

摘要:

AbstractTritium production cross sections have been inferred from direct measurements of tritium generated in wafers of6LiH and7LiH under bombardment by 15-MeV neutrons produced at the Lawrence Livermore National Laboratory’s Rotating Target Neutron Source-I facility. Sealed in a thin-walled lead container, each hydride wafer was immersed in boiling mercury that first amalgamated the lead and then dissociated the LiH. The hydrogen, acting as a carrier, was directed to an electronic counter and mixed carefully with methane. The counting procedure provided an accurate measure of tritium originally generated in each wafer. The TART Monte Carlo code was employed in the analysis of the data. The tritium production cross section for6Li exposed to 14.92-MeV neutrons is 32±3 mb and that for7Li exposed to 14.94-MeV neutrons is 302±18 mb.

ISSN:0029-5639    

DOI:10.13182/NSE85-A27440

出版商:Taylor&Francis    

年代:1985

数据来源: Taylor

摘要:

AbstractNeutron inelastic scattering from232Th has been studied for states above 700 keV in excitation using the (n,n'γ) technique at incident energies in the 0.77- to 2.10-MeV range. The gamma-ray measurements employing a high-resolution Ge(Li) spectrometer show 74 transitions from 46 levels above the first excited state. Gamma-ray branching ratios and production cross sections have been determined. Inferred level cross sections are compared to compound-nucleus statistical model calculations, which generally represent the measurements fairly well at lower incident energies but become increasingly discrepant at higher bombarding energies, and to the ENDF/B-V evaluation.

ISSN:0029-5639    

DOI:10.13182/NSE85-A27441

出版商:Taylor&Francis    

年代:1985

数据来源: Taylor

摘要:

AbstractThe analysis of two91Zr neutron transmission experiments performed some years ago at the Oak Ridge National Laboratory and Geel electron Linacs showed systematic discrepancy in the resonance neutron width estimation. The two sets of data are comprehensively reanalyzed by a shape least-squares method that builds up a single system of normal equations. An accurate resolution function description was obtained by Monte Carlo calculation of the moderation and detection contributions. The two experiments are found to be in complete agreement and a single set of resonance neutron widths is obtained.

ISSN:0029-5639    

DOI:10.13182/NSE85-A27442

出版商:Taylor&Francis    

年代:1985

数据来源: Taylor

摘要:

AbstractIn the performance of probabilistic risk assessments, in which there are inevitably large uncertainties, it is customary to characterize the computed probabilities in terms of their medians. When this is done, it is incorrect to add the probabilities of different accident sequences to find an overall probability of some consequence (like core melt), or to add the risks of the members of a population of reactors to find the societal risk. The error is not only one in principle, but is substantial when the uncertainties are large. In addition, the uncertainties are reduced when the probabilities are combined properly. Some examples are given.

ISSN:0029-5639    

DOI:10.13182/NSE85-A27443

出版商:Taylor&Francis    

年代:1985

数据来源: Taylor

摘要:

AbstractMultiple-collision solutions of the time-, space-, and angle-dependent neutron transport equation in slab geometry are given. Two different monodirectional sources have been used: (a) aδ(t)-shaped pulse of neutrons [δ(t): Dirac delta distribution] impinging on the slab at time t = 0, and (b) a“rectangular”source, emitting neutrons for a time intervalΔt, describing a somewhat more realistic situation. Detailed results up to collision order three are discussed and exhibited in several figures. Interestingly, the“scalar”flux of one-time-scattered neutrons for the slab problem turns out to be independent of space in the region influenced by the slab boundaries.

ISSN:0029-5639    

DOI:10.13182/NSE85-A27444

出版商:Taylor&Francis    

年代:1985

数据来源: Taylor