[en] Simple diffusion theory cannot be used to evaluate control rod worths in thermal neutron reactors because of the strongly absorbing character of the control material. However, reliable control rod worths can be obtained within the framework of diffusion theory if the control material is characterized by a set of mesh-dependent effective diffusion parameters. For thin slab absorbers the effective diffusion parameters can be expressed as functions of a suitably-defined pair of blackness coefficients. Methods for calculating these blackness coefficients in the P₁, P₃, and P₅ approximations, with and without scattering, are presented. For control elements whose geometry does not permit a thin slab treatment, other methods are needed for determining the effective diffusion parameters. One such method, based on reaction rate ratios, is discussed

[en] Reasonable agreement between calculated and measured differential shim rod worths in the Oak Ridge Research Reactor (ORR) has been achieved by taking into account the combined effects of negative reactivity contributions from changing fuel-moderator temperatures and of delayed photoneutrons. A method has been developed for extracting the asymptotic period from the shape of the initial portion of the measured time-dependent neutron flux profile following a positive reactivity insertion. In this region of the curve temperature-related reactivity feedback effects are negligibly small. Results obtained by applying this technique to differential shim rod worth measurements made in a wide variety of ORR cores are presented

[en] The WIMS-D4 code has been modified (WIMS-D4M) to produce microscopic isotopic cross sections in ISOTXS format for use in diffusion and transport calculations. Beginning with 69-group libraries based on ENDF/B-V data, numerous cell calculations have been made to prepare a set of broad group cross sections for use in diffusion calculations. Global calculations have been made for two control rod states of the Romanian steady state TRIGA reactor with 29 fresh HEU fuel clusters. Detailed Monte Carlo calculations also have been performed for the same reactor configurations using data based on ENDF/B-V. Results from these global calculations are compared with each other and with the measured excess reactivities. Although region-averaged macroscopic principal cross sections obtained from WIMS-D4M are in good agreement with the corresponding Monte Carlo values, problems exist with the high energy (E > 10 keV) microscopic hydrogen transport cross sections

[en] Simple diffusion theory cannot be used to evaluate control rod worths in thermal reactors because of the strongly absorbing character of the control material. However, good results can be obtained from a diffusion calculation by representing the absorber slab by means of a suitable pair of internal boundary conditions, α and β, which are ratios of neutron flux to neutron current. Methods for calculating α and β in the P₁, P₃, and P₅ approximations, with and without scattering, are presented. By appropriately weighting the fine-group blackness coefficients, broad group values, <α> and <β>, are obtained. The technique is applied to the calculation of control rod worths of Cd, Ag-In-Cd, and Hf control elements. Results are found to compare very favorably with detailed Monte Carlo calculations. For control elements whose geometry does not permit a thin slab treatment, other methods are needed for determining the effective diffusion parameters. One such method is briefly discussed and applied to the calculation of control rod worths in the Ford Nuclear Reactor at the University of Michigan. Calculated and measured worths are found to be in good agreement

[en] The Ford Nuclear Reactor (FNR) went critical on December 8, 1981 with 23 LEU fuel elements. Five of these 23 elements were fabricated by CERCA and the others by NUKEM. Since that time a substantial data base of experimental results for LEU cores has been accumulated by the University of Michigan FNR staff. This paper compares some of the experimental data with analytical calculations based, for the most part, on three-dimensional diffusion theory. The critical configuration, control rod worths, axial rhodium reaction rate profiles and thermal flux distributions have been calculated and compared with measurements

[en] Analytical methods used to analyze neutronic data from the whole-core LEU fuel demonstration in the Oak Ridge Research Reactor are briefly discussed. Calculated eigenvalues corresponding to measured critical control rod positions are presented for each core used in the gradual transition from an all HEU to an all LEU configuration. Some calculated and measured results, including β/sub eff//l/sub p/, are compared for HEU and LEU fresh fuel criticals. Finally, the perturbing influences of the six voided beam tubes on certain core parameters are examined. For reasons yet to be determined, differential shim rod worths are not well-calculated in partially burned cores

[en] The ORR Whole-Core LEU Fuel Demonstration, conducted as part of the US Reduced Enrichment Research and Test Reactor Program, has been successfully completed. Using commercially-fabricated U₃Si₂-Al 20%-enriched fuel elements (4.8 g U/cc) and fuel followers (3.5 g U/cc), the 30-MW Oak Ridge Research Reactor was safely converted from an all-HEU core, through a series of HEU/LEU mixed transition cores, to an all-LEU core. There were no fuel element failures and average discharge burnups were measured to be as high as 50% for the standard elements and 75% for the fuel followers. Experimental results for burnup-dependent critical configurations, cycle-averaged fuel element powers, and fuel-element-averaged ²³⁵U burnups validated predictions based on three-dimensional depletion calculations. Calculated values for plutonium production and isotopic mass ratios as functions of ²³⁵U burnup support the corresponding measured quantities. In general, calculations for reaction rate distributions, control rod worths, prompt neutron decay constants, and isothermal temperature coefficients were found to agree with corresponding measured values. Experimentally determined critical configurations for fresh HEU and LEU cores radially reflected with water and with beryllium are well-predicted by both Monte Carlo and diffusion calculations. 17 refs

[en] The ORR Whole-Core LEU Fuel Demonstration, conducted as part of the US Reduced Enrichment Research and Test Reactor Program, has been successfully completed. Using commercially-fabricated U₃Si₂-Al 20%-enriched fuel elements (4.8 g U/cc) and fuel followers (3.5 g U/cc), the 30-MW Oak Ridge Research Reactor was safely converted from an all-HEU core, through a series of HEU/LEU mixed transition cores, to an all-LEU core. There were no fuel element failures and average discharge burnups were measured to be as high as 50% for the standard elements and 75% for the fuel followers. Experimental results for burnup-dependent critical configurations, cycle-averaged fuel element powers, and fuel-element-averaged ²³⁵U burnups, validated predictions based on three-dimensional depletion calculations. Calculated values for plutonium production and isotopic mass ratios as functions of ²³⁵U burnup support the corresponding measured quantities. In general, calculations for ⁶⁰Co and ¹⁹⁸Au reaction rate distributions, differential and integral control rod worths, prompt neutron decay constants, and isothermal temperature coefficients were found to agree with corresponding measured values. Experimentally determined critical configurations for fresh HEU and LEU cores radially reflected with water and with beryllium are well-predicted by both Monte Carlo and diffusion calculations. 44 refs., 57 figs., 45 tabs

[en] More than two years ago the government of Indonesia announced plans to purchase a research reactor for the Puspiptek Research Center in Serpong Indonesia to be used for isotope production, materials testing, neutron physics measurements, and reactor operator training. Reactors using low-enriched uranium (LEU) plate-type and rod-type fuel elements were considered. This paper deals with the neutronic evaluation of the rod-type 25-MW LEU TRIGA Multipurpose Research Reactor (MPRR) proposed by the General Atomic Company of the United States of America

[en] At the Reduced Enrichment for Research and Test Reactors (RERTR) meeting in September 1994, Durand reported that the maximum uranium loading attainable with U₃Si₂ fuel is about 6.0 g U/cm³. The French Commissariat a l'Energie Atomique (CEA) plan to perform irradiation tests with 5 plates at this loading. Compagnie pour L'Etude et La Realisation de Combustibles Atomiques (CERCA) has also fabricated a few uranium nitride (UN) plates with a uranium density in the fuel meat of 7.0 g/cm³ and found that UN is compatible with the aluminum matrix at temperatures below 500 C. High density dispersion fuels proposed for development include U-Zr(4 wt%)-Nb(2 wt%), U-Mo(5 wt%), and U-Mo(9 wt%). The purpose of this note is to examine the relative neutronic behavior of these high density fuels in a typical light water-reflected and water-moderated MTR-type research reactor. The results show that a dispersion of the U-Zr-Nb alloy has the most favorable neutronic properties and offers the potential for uranium densities greater than 8.0 g/cm³. On the other hand, UN is the least reactive fuel because of the relatively large ¹⁴N(n,p) cross section. For a fixed value of k_eff, the required ²³⁵U loading per fuel element is least for the U-Zr-Nb fuel and steadily increases for the U-Mo(5%), U-Mo(9%), and UN fuels. Because of volume fraction limitations, the UO₂ dispersions are only useful for uranium densities below 5.0 g/cm³. In this density range, however, UO₂ is more reactive than U₃Si₂