Anl 5800 pdf

For groups the hafnium capture cross sections are small enough so that normal diffusion theory can be applied. The method used in this study treats hafnium as a non-diffusing medium for groups by applying a group-dependent internal boundary condition at the hafnium surfaces.

This internal boundary condition is the neutron current-to-flux ratio. Beginning with the same cylindrical model used to generate the WIMS cross sections, but with a much finer mesh within and about the hafnium rods, PIS16 transport calculations were used to determine numerical values for the internal boundary conditions.

For a few groups, these values were too large and thus were replaced by values corresponding to the limiting case of a "black" rod of the same radius. Figure 5, derived from data in Ref.

Within lo statistics both methods give the same values for the Hf control rod worths. For the purpose of these calculations, the "rods out" condition was taken to be that with the hafnium Figs. Because of the complicated geometry of these components Fig. The neutron absorption effect of all the fission products not present in the ANL MCNP library was represented in terms of equivalent, region-dependent, l0B concentrations.

Each fuel element was divided into a 3x3 array of equal-volume regions. As a first approximation, the l0B concentrations were chosen to match the combined thermal neutron absorption rates of , l35Xe, P,, and the lumped fission products.

Table 7 summarizes the results for the reactivity worths of the ANS reflector components. The MCNP calculations show that the worth of the reflector components at EOC with the control rods withdrawn is smaller but nearly equal to the BOC value with the control rods fully inserted.

Emphasis is placed on eigenvalues, peak thermal neutron fluxes, and core lifetime comparisons. For the purpose of these comparisons, the control rod "out" configuration corresponds to the inserted Hf rods Figs. In reality, however, the fully withdrawn control rods are parked in the upper reflector with the bottom of the hafnium located on the plane where the top of the hafnium would be for a fully inserted rod.

Figures show the control rods in the fully inserted position. The peak thermal neutron flux is a function of control rod elevation and is largest for fully inserted rods. However, the product of the flux and the eigenvalue is rather insensitive to control rod elevation and fuel bumup. Since the reactor would operate at an effective multiplication factor of unity, this product is a realistic estimate of the peak thermal neutron flux.

Icff Flux. Oo19 6. MCXP statistics are the la values. Maximum core lifetime estimates have been made using the EOC reactivity balance table shown in Table This table was constructed from data given in Tables Normally, a reactivity reserve would be included in this table.

This reserve is needed to overcome the buildup of l35Xe and restart the reactor within a short time after an unantic! This balance table determines the minimum unperturbed multiplication factor needed at EOC from which the maximum core lifetime can be found. Figure 7 shows how these lifetime estimates were obtained. The estimated maximum core life for the I Three-Element We have analyzed this "overlapping" core configuration for the With the control rods fully inserted, BOC MCNP Monte Carlo calculations showed that the peak unperturbed thermal neutron flux for this overlap core increased by a factor of 1.

For the MCNP perturbed calculations the peak flux for the overlap core decreased by a factor of 0. An estimate of the lifetime of this overlap core was determined, as described earlier, on the basis of an EOC reactivity balance table.

Using results in Table 7 as a guide for extrapolation, the EOC worth of the reflector components for this overlap configuration was estimated to be 8. This translates into a required EOC eigenvalue of 1. Figure 8 shows the unperturbed eigenvalues as a function of exposure in FPD as well as core lifetime estimates for the normal and overlap configurations for the This figure indicates that no significant improvement in core life results from the overlap configuration.

The increased worth of the reflector components offsets the advantages of larger eigenvalues for the overlap core. Sensitivity of Lifetime and Performance to the Choice of Structural Material For the A N S cores a significant reactivity loss is associated with parasitic neutron capture in the Al structural materials.

Sizeable reactivity gains would be achieved if some or dl of the Ab could be replaced with a material significantly less absorbing to neutrons. To investigate the potential effects of this type of substitution, some calculations were performed with magnesium replacing Al in some or all of the structures of selected A N S cores. The results show that dramatic improvements would result if substitutions of this type were feasible.

For example, substitution of all structural materials excluding fuel plates in the LEU Similar results were obtained for other A N S cores. In particular, for the The control rods are parked in the upper reflector. The H,O contamination in D,O is 0.

The peak thermal neutron flux for the unperturbed HEU reference core wa.. It LOO! Thus, the effect of using structural materials other than Al even in structures which will be present in the A N S core for only short periods of time, like the fuel element side plates, would be very significant. These considerations address only the neutronic effects of the substitution.

Whether high magnesium content alloys are available or can be developed to withstand the harsh environment of an A N S core is of fundamental importance, but is an issue not addressed in this study.

This table illustrates the size of neutron flux penalties associated with the use of larger volume cores required by fuels of reduced enrichment and also flux reductions caused by the reflector components.

Since the reactivity balance tables, upon which the core lifetime estimates are based, do not include allowances for in-core target facilities nor for any operational reserves, it is unlikely that any of these cores will operate for 17 days at MW of fission power.

However, the uranium density for the HEU core could be increased to offset these additional reacti-eity requirements. Similarly, the However, design improvements will be needed for obtaining substantial reactivity additions before the This suggests that perhaps some of the aluminum filler assigned to the LEU fuel plates is not needed and could be replaced with additional fuel meat.

Carefully placed burnable poisons will help limit initial power peaking values. But it seems likely that this LEU core design will require the development and certification of higher density fuels with improved thermal conductivities. Several important issues have not been addressed by this study.

Table 2 shows that for these three core designs the fuel plates in the upper element of the LEU core are the least stable. To reduce power peaking effects it is assumed that, like the HEU reference core, the LEU fuel will need to be graded in both the axial and radial directions. Whether the high density LEU fuel can be graded and fabricated with acceptable yields is another important issue outside the scope of this study. Department of Energy and Robert A.

Renier and J. This information allowed us to cany out this study using ANS models consistent with theirs. ANL, March Costescu, D. Cacuci, J. Deen, and W. Deen, W. Woodruff, and C. Templin, Ed. Gallmeier, et al. Use the facets available to narrow your search. Authors names may appear more than once because the Catalog indexes both the names of the authors and the various forms of their names that they use on their publications.

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