Relative assembly power distribution

The maximum-to-average fuel bundle peaking or radial distribution is reduced in a BWR core because of greafer sfeam voids in the center bundles of the care. A control rod operating procedure is also used to maintain approximately the same radial power shape throughout an operating cycle. 

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More dilute fuel could be burned more completely in shorter cycles. However, capacity factors would likely be higher with longer cycles. The fuel could be managed similar to a commercial LWR. Cycles could last one or two years, then one-third or one-fourth of the core could be replaced with fresh fuel. Thus, the typical fuel assembly would remain in the core several years, until a high exposure level is achieved. Commercial LWR fuel management strategies could be used to maintain relatively flat power distributions. 

Each rod cluster control assembly consists of 24 absorber rods fastened at the top end to a common hub (or spider) assembly. The rod cluster control assemblies are used to make relatively rapid changes in reactivity and to control the axial power distribution. Figure 4.2-9 of Reference 6.1 shows a rod cluster control assembly, with Figure 4.2-10 giving detail of an individual absorber rod. 

Figure XXIV-4 illustrates the thermal-hydraulic characteristics of the PEACER core. The temperature distribution shows that the maximum coolant temperature of 460°C exceeds the specified design temperature range (300-A00°C). Figure XXIV-5 shows the relative fuel rod power distribution in the hottest assembly.

ASMBURN uses the macro-cross section sets of the fuel cells, which are prepared by the cell burnup calculations in advance as described above. Those cell bumup calculations assume that one fuel cell is surrounded by the same fuel cells. Therefore, the ASMBURN modeling is not applicable when different types of fuel rods are aligned in large irregularities. ASMBURN first interpolates the macrocross section of each fuel cell by the burnup as shown in Fig. 2.18 [9]. Then the neutron flux distribution is calculated and normalized by the thermal power of the fuel assembly at each bumup step. The bumup increase of each fuel cell is calculated by multiplying the relative power distribution by the time exposure at each bumup step, and the calculation proceeds to the next bumup step. 

From these calculations, the interfuel assembly gap size is determined to be 4.0 mm. In this case, the local power peaking factor takes the lowest value of 1.06 without fuel rod enrichment controls. The relative fuel rod power distribution for the case with inter-fuel assembly gap size of 4.0 mm is shown in Fig. 2.38 [9]. The pin number (from 1 to 46) on the x axis of this figure corresponds to the pin number position shown in Fig. 2.37 [9]. Although the pin powers tend to be relatively high near the middle of the water rods, and relatively low at the corners of the water rods, the overall power distribution is flat. 

The control rod patterns are determined for each of the 15 bumup steps of the equilibrium cycle (cycle bumup exposure of 0-14.8 GWd/t). Figure 2.54 [9] shows the control rod patterns for the equilibrium core (1/4 core symmetry). Each box represents a fuel assembly and the value in the box represents the control rod withdrawn rate out of 40. A blank box represents a fuel assembly with control rods completely withdrawn. While the control rod patterns are adjusted at every 1.1 GW/t throughout most of the cycle, the fine adjustment of the control rod pattern at a cycle bumup of 0.22 GWd/t is necessary to compensate for a rapid drop of BOC excess reactivity. The excess reactivity drop is relatively fast with respect to the bumup at BOC because of the initial build up of xenon gas and other fission products. The concentration of xenon reaches eqmlibrium shortly after operation commences and from there, the rate of the excess reactivity drop becomes lower and almost constant. The control rod patterns are determined by considering control of the core power distributions while keeping the core critical. The radial core... 

The above relationships assume that the maximum power point always appears in the maximum power fuel assembly. Such an assumption may be acceptable when the core power distribution is relatively smooth, and it seems to be acceptable for the Super LWR core design as far as the three-dimensional core calculation results are concerned. 

Since the transient subchannel analysis code does not have the functions prepared in typical system analysis codes, several parameters are taken from the calculation results by the single channel safety analyses performed in Sect. 6.7. These parameters are the flow rate, temperature and pressure at the inlet of the hot fuel assembly, and the relative power. The radial and axial power distributions are assumed not to change with time. This is reasonable because the reactivity is not locally changed at the flow decreasing events. 

Figiue 7.28 [1] compares radial power distributions with ordinary blanket fuel rods or thick walled duct tubes in the outer region of blanket assemblies. Simplified R-Z calculations for radial heterogeneous cores are used to clarify the effect of the duct tubes. The relative power peak near the seed and blanket interface is significantly reduced by replacing the blanket fuel rods with the duct tubes. 

In order to evaluate the influence of the subchannel heterogeneity, the pin power distribution is set as imiform and the axial power distribution is set as cosine with the maximum linear heat generation rate of 39 kW/m. Figure 7.43 [1] shows the mass flux distribution at the assembly outlet. Due to a relatively large hydraulic... 

Fig. 7.44 Relative distribution of outlet coolant and peak cladding temperatures in seed fuel assembly under uniform pin power distribution. (Taken from [1])...

However, when calculating a thermal-spectrum core with large heterogeneities, the R-Z two-dimensional model is inadequate for design purposes. In a thermal-spectrum core, the spatial dependence of the thermal neutron flux is large. The fuel assemblies are loaded with a relatively complex pattern to flatten the neutron flux distributions. Hence, the calculation of such a core requires the modeling of each fuel assembly with a three-dimensional model as shown in Fig. 2.31. To conserve computational power, symmetric boundary conditions can be applied. 

Figure 2.68 [9] shows the relative coolant flow rate. The distribution is determined by the inlet orifice attached to each fuel assembly for the 1/4 symmetric core (the flow rate is not normalized, and the average is 0.99). The outer (or peripheral) fuel assemblies are cooled by downward flow. A relatively large flow rate can be distributed to the outer fuel assemblies compared with the expected power generation because the outlet coolant temperature does not need to be high. By eliminating... 

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