摘要:

AbstractUnder actual reactor conditions, the mechanical characteristics of nuclear fuel are particularly affected by thermal expansion of the fuel pellets. The results of an experiment studying the vibration behavior of nuclear fuel under reactor conditions are presented. The experiment was conducted in an experimental reactor in a two-phase flow test loop simulating a Canada deuterium uranium (CANDU) reactor fuel channel. The effect of fuel power in particular was investigated. The natural frequency of the fuel elements increased by∼50% at full power. The effect of the fuel channel parameters such as flow velocity, steam quality, and pressure were also investigated.

ISSN:0029-5639    

DOI:10.13182/NSE93-A24032

出版商:Taylor&Francis    

年代:1993

数据来源: Taylor

摘要:

AbstractAn intense reusable source of pulsed photoneutrons is developed that produces≍0.5 or 1.0×1014neutrons in an∼15-ns pulse from natural lead or depleted uranium, respectively, on the HERMES III electron accelerator. Corresponding to this source, a numerical model is developed that is applicable to other pulsed-power systems. If Vprepresents the peak voltage of HERMES III measured in megavolts, then model predictions show that over the range 12 MV< Vp<20 MV, the number of neutrons produced per incident electron is 7.2×10-6(VP—11)2.0and 1.2×10-6(VP—7.4)2 8in lead and uranium, respectively. Measurements using a set of nuclear activation foils confirm these predictions as well as predictions of the spatial and spectral distribution of the neutrons at Vp= 19 MV.

ISSN:0029-5639    

DOI:10.13182/NSE93-A24033

出版商:Taylor&Francis    

年代:1993

数据来源: Taylor

摘要:

AbstractThe transmitted photon spectra of133Ba,137Cs, and60Co sources are measured through different thicknesses of a soil medium. A multiple-scatter peak is observed at an energy of 100 keV independent of incident photon energy and thickness of the medium. The intensity of the peak decreases exponentially with the increase in thickness of the medium. A linear relationship between the effective atomic number of the medium and the multiple-scatter peak energy is observed.

ISSN:0029-5639    

DOI:10.13182/NSE93-A24034

出版商:Taylor&Francis    

年代:1993

数据来源: Taylor

摘要:

AbstractThe MCNP Monte Carlo transport code is used by the Los Alamos National Laboratory Health and Safety Division for a broad spectrum of radiation shielding calculations. One such application involves the determination of skyshine dose for a variety of photon sources. To verify the accuracy of the code, it was benchmarked with the Kansas State University (KSU) photon skyshine experiment of 1977. The KSU experiment for the unshielded source geometry was simulated in great detail to include the contribution of groundshine, in-silo photon scatter, and the effect of spectral degradation in the source capsule. The standard deviation of the KSU experimental data was stated to be 7%, while the statistical uncertainty of the simulation was kept at or under 1%. The results of the simulation agreed closely with the experimental data, generally to within 6%. At distances of under 100 m from the silo, the modeling of the in-silo scatter was crucial to achieving close agreement with the experiment. Specifically, scatter off the top layer of the source cask accounted for∼12% of the dose at 50 m. At distances>300 m, using the60Co line spectrum led to a dose overresponse as great as 19% at 700 m. It was necessary to use the actual source spectrum, which includes a Compton tail from photon collisions in the source capsule, to achieve close agreement with experimental data. These results highlight the importance of using Monte Carlo transport techniques to account for the nonideal features of even“simple experiments.”

ISSN:0029-5639    

DOI:10.13182/NSE93-A24035

出版商:Taylor&Francis    

年代:1993

数据来源: Taylor

摘要:

AbstractRay effects, an inherent problem in the formulation of the discrete ordinates approximation to the transport equation, are studied. In particular, the effectiveness of using Monte Carlo procedures to generate a first or second collision source is investigated. Monte Carlo procedures provide a general methodology that can be applied to the discrete ordinates solution of complex problems in either two-dimensional or three-dimensional geometries for which ray effects are likely to occur. The Monte Carlo method, which is intrinsically free of ray effects, performs the transport of the source particle to the first collision sites, at which estimates for the uncollided fluxes are made. The uncollided fluxes are then used to compute the first collided fluxes. The uncollided, collided, or first collided fluxes are calculated as first or second collision scattering sources in a format suitable for input into the DORT two-dimensional and TORT three-dimensional discrete ordinates codes. The computational time and precision requirements of the Monte Carlo calculation are analyzed. The results show that significant improvements are achieved in the solution of test problems when using the estimated first collision source and that ray effects are virtually eliminated when using the estimated second collision source.

ISSN:0029-5639    

DOI:10.13182/NSE93-A24036

出版商:Taylor&Francis    

年代:1993

数据来源: Taylor

摘要:

AbstractTwo methods are presented for optimally calculating spatial distributions of neutron flux in a nuclear reactor core. Both techniques, Kalman filtering and maximum likelihood estimation, simultaneously account for all initial information contained in the nominal core specifications and in-core measurements, as well as all of the uncertainties within the system, to provide a minimum variance estimate of neutron flux. These methods resolve discrepancies in the initial information in a statistically optimal manner, thereby providing valuable insight into the nature of the optimal solution obtained. Despite radically different algorithms, both methods yield the same minimum variance estimate for the quantity of interest. The algorithms have been successfully tested for one-dimensional axial and two-dimensional x-y flux mapping problems with simulated in-core data sets.

ISSN:0029-5639    

DOI:10.13182/NSE93-A24037

出版商:Taylor&Francis    

年代:1993

数据来源: Taylor

摘要:

AbstractThe neutron diffusion equation in reactor physics is solved on a multiple-instruction, multiple-data parallel computer network composed of five transputers. A parallel variant of the Schwarz alternating procedure for overlapping subdomains is used for domain decomposition. The parallel Schwarz algorithm with the concept of underrelaxation in pseudo-boundary conditions is applied to two types of reactor benchmark problems: fixed-source problems and eigenvalue problems. Results of parallel computation for these problems are reported and compared with results of sequential computation. The results show that a very high speedup can be achieved in fixed-source problems in spite of the small problem size and that a relatively high speedup, although lower than that of fixed-source problems, can be obtained in eigenvalue problems.

ISSN:0029-5639    

DOI:10.13182/NSE93-A24038

出版商:Taylor&Francis    

年代:1993

数据来源: Taylor