摘要:

AbstractThe combination of a gaseous core fission reactor with a magnetohydrodynamic (MHD) generator can lead to more efficient conversion of fission energy to electricity than can conventional conversion systems. A system concept currently being investigated utilizes uranium tetrafluoride (UF4) as fuel and potassium or potassium fluoride (KF) as the working fluid. The electrical conductivity of the gas greatly influences the performance of the MHD generator. It is possible to enhance the electrical conductivity by taking advantage of fission fragment ions born in the fissile gas-working gas mixture. To study and quantify this effect, a chemical reaction model as well as a physical model are developed. The governing rate equations and an electron energy balance equation are numerically solved for steady-state and spatially homogeneous cases. The electrical conductivity of a UF4-K/KF gaseous mixture is shown to be a function of neutron flux at representative gas conditions (2500 K and 1 atm). The enhancement is achieved because of the rise in electron temperature due to fission fragment heating.

ISSN:0029-5639    

DOI:10.13182/NSE92-A23881

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractThe photodisintegration of232Th,233U,238U,237Np, and239Pu has been studied with monochromatic photons produced by neutron capture using the IEA-R1 reactor. The ratiosσγf(X)/σγf(238U) (relative fissionability),σγf/σα(photofission branching ratio), andΓn/Γf(neutron emission and fission competition) are shown to be independent of excitation energy, within the experimental errors, in the 6.73- to 9.72-MeV energy interval. Some correlations of these ratios with Z2/A and (E′f–B′n) are performed, and the results are in reasonable agreement with the theoretical model predictions.

ISSN:0029-5639    

DOI:10.13182/NSE92-A23882

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractGamma-ray and neutron dose-equivalent buildup factors are calculated for six common shielding materials in a point-source, infinite-slab, point-detector geometry using a decomposition of the solution to the transport problem into single- and multiple-scatter components. A rigorous solution for the single-scatter component is constructed and a Monte Carlo model for the multiple-scatter component is employed. Simplified models are fit to the calculated buildup factors as functions of slab thickness and source-detector separation, and model constants are evaluated for each of several source energies. Single-scatter and total slab buildup factors are presented, both in tabular form and in graphs that also show the fitted models, for six materials. The models are demonstrated for a sample problem.

ISSN:0029-5639    

DOI:10.13182/NSE92-A23883

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractExactly describing the time of thermal neutron collisions with water molecules, orientation averaging is performed by an exact method (EOAK) and four approximate methods (two well known and two less known). Expressions for the microscopic scattering kernel are developed. The two well-known approximate orientation averaging methods are Krieger-Nelkin (K-N) and Koppel-Young (K-Y). The results obtained by one of the two proposed approximate orientation averaging methods agree best with the corresponding results obtained by EOAK. The largest discrepancies between the EOAKresults and the results of the approximate methods are obtained using the well-known K-N approximate orientation averaging method.

ISSN:0029-5639    

DOI:10.13182/NSE92-A23884

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractIt is argued that79Kr is uniquely suited for an intense positron source. It can be produced by neutron activation of a rare, but available, stable isotope78Kr; it has a convenient 35-h half-life; as a nonreactive gas, it can be transported and recycled in a closed system using automated valves without exposure of personnel; and it can be vapor deposited easily on a large area cold surface using a solid neon moderator to make a slow positron source with intensity (≈1011s−1) limited only by the availability of neutrons and cryogenic refrigeration.

ISSN:0029-5639    

DOI:10.13182/NSE92-A23885

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractThere is a problem in the neutron flux calculation in regions with a strong spectral transition from epithermal toward thermal. Space-dependent group constants are developed for the thermal range to treat the highly nonseparable space- and energy-dependent flux distribution that characterizes the transition of fast neutron spectra into partially thermalized spectra. The weighting spectra are obtained from a parametric application of the heavy gas model for scattering with absorption cross sections that include the resonances near and below 1 eV. A space dependence is introduced into weighting spectra by relating the parametric solution of the zero-dimensional spectral equation to thermal and epithermal group fluxes obtained from a prior one-dimensional diffusion calculation. Subsequently, space-dependent thermal group constants are generated. The method is implemented in a standard multigroup diffusion code, executed iteratively. This procedure was applied to compact liquid-metal reactor designs having thermalizing reflector regions. The results indicate the effect of global parameters such as the size of the thermalizing reflector on the group constants, which are considerably different from the classical local group constants.

ISSN:0029-5639    

DOI:10.13182/NSE92-A23886

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractThe accuracy of the nuclear design code system for the High-Temperature Engineering Test Reactor (HTTR) is evaluated for the neutronic characteristics that depend on core temperature by analyzing the overall temperature coefficients of reactivity and the effective multiplication factors obtained by an experiment in which the Very High Temperature Reactor Critical Assembly (VHTRC) is heated from ambient temperature to 200°C. The core of the VHTRC consists of block-type fuel containing low-enriched uranium (LEU).The nuclear design code system for the HTTR includes the DELIGHT, TWOTRAN-2, and CITATION-1000VP computer codes. The DELIGHT code is a one-dimensional cell burnup code developed to evaluate the nuclear characteristics of HTTR fuel and to calculate the group constants.The calculated overall temperature coefficients of reactivity between ambient temperature and 200°C agree well with the measured coefficients, and the calculated effective multiplication factors for different temperatures agree with measured factors within an uncertainty of 0.6%. From the results, it is concluded that the nuclear design code system for the HTTR predicts well the temperature-dependent neutronic characteristics of a core containing LEU fuel.

ISSN:0029-5639    

DOI:10.13182/NSE92-A23887

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractInterrogation neutrons from 3 eV to 3 keV are used to determine the relative sensitivity of a spent light water reactor fuel assembly assay system. The fuel assay system used for this measurement consists of three threshold fission chambers installed in the Rensselaer intense neutron spectrometer, a 75-t lead slowing-down-time spectrometer at the Gaerttner Linac Laboratory. The fission chambers detect fission neutrons from a simulated fuel assembly, an aluminum enclosure filled with depleted uranium oxide (0.2%235U), and a235U (93%) metal foil sample placed at various locations throughout the assembly. The measurements with the assembly are compared with a Monte Carlo analysis of an homogenized pressurized water reactor fuel assembly. This is concluded to be a practical method for the assay of spent fuel.

ISSN:0029-5639    

DOI:10.13182/NSE92-A23888

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractThe vectorial form of the neutron streaming term used in the Boltzmann equation has been derived. With the aid of this vectorial form, the explicit expression of the neutron streaming term in a generalized coordinate system can readily be obtained. Examples of applications in the orthogonal curvilinear and the elliptical-elliptical toroidal coordinate systems are given for illustration.

ISSN:0029-5639    

DOI:10.13182/NSE92-A23889

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractThe dynamics of continuously fueled fission reactors based on an isothermal and nonlinear description of neutron/nuclide interactions is considered. In general, and for a plausible range of operational fueling conditions, such systems are absolutely stable, at most displaying converging spiral phase-space trajectories.

ISSN:0029-5639    

DOI:10.13182/NSE92-A23890

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor