Increased axial and radial complexity in fuel designs becomes even more relevant in those iPWRs that rely on control rods or heavy BP loadings rather than soluble boron in the coolant to control the excess reactivity. Unlike large PWRs where there may be a control rod located in one out of three or four assemblies, the iPWRs have a much higher density of control rods, particularly if the control rods are the means by which the excess reactivity is controlled. For example, in the mPower and SMR-160 designs, every assembly contains a control rod. In a large PWR where boron is used to control the excess reactivity, the control rods are normally fully withdrawn. But in some of the iPWRs, these control rods, and the sequence in which they are maneuvered, allows for compensation in the reduction in excess reactivity (as the fissile material depletes), for axial power changes, and for xenon variations (radially and axially) — see Table 4.1. The removal of the need for boron simplifies the chemical volume control systems and results in capital and operational cost savings, e. g., waste handling of depleted soluble boron.

However, at the same time, inserting a control rod moves the power away from its location to somewhere else in the core, and excessive movements can thermally ‘cycle’ the fuel, potentially resulting in fuel failures. There is also a need to control the rate of withdrawal due to fuel performance (such as pellet clad mechanical interaction, or thermal cycling) and for reactivity insertion limits (as explained in Section 4.2). Therefore, in terms of nuclear design, this leads to the need not only to optimize the fuel and core design, but also to develop a control rod sequence for the various control rod groupings to ensure appropriate compensation in reactivity, while at the same time, also ensuring other limits, such as changes in reactivity or power ramp rates for the fuel rods, are not violated. This development of the control rod sequence includes:

• determining which control rods get moved together, in so-called ‘control rod banks’;

• when they get moved in or out, how far and at what rate;

• what the overlap is between one bank and the next one starting to move in/out.

Even with the use of soluble boron, similar challenges regarding control of excess reactivity are seen for those iPWRs looking to achieve long cycle lengths between refueling outages, as greater fissile loadings are required to achieve the cycle length of greater than 24 months.

Using control rod insertion throughout a cycle results in strongly varying axial power shapes; inserting control rods in an iPWR will skew the power to the bottom of the core and result in higher peaking factors. This effect is mitigated by using higher burnable poison loadings in the lower parts of the fuel, or varying the fissile enrichment axially. In addition to the axial power effects, those assemblies that have the control rods inserted during the cycle tend to have their peak pin powers pushed radially to the edge of the assembly, i. e., away from the inserted control rod locations. Therefore, additional radial BP loadings tend to be needed to reduce the radial peaking factors. Overall, this tends to lead to a heterogeneous nuclear design in terms of fuel assemblies, and areas of notable flux suppression in the fuel pins around the BPs and control rods, and results in power peaking around the inserted control rods and BP locations. The disadvantage of a more heterogeneous core is a non-uniform radial and axial depletion throughout the core, which results in either inefficient use of the fuel (as burnup limits are met in some, but not all fuel batches), or difficulty in achieving power peaking limits. Furthermore, for the iPWRs that use natural circulation (e. g., NuScale and SMR-160), the non-uniformity in the power and resulting moderator density will require further thermal hydraulic analysis to ensure sufficient cooling and an accurate prediction and coupling of the neutronic and thermal hydraulic feedback. This issue, which needs to be addressed for all natural circulation designs, is exacerbated by the fuel non-uniformity.

In the case of high gadolinium loadings being used, the iPWR’s nuclear designers will have to consider very carefully the fissile enrichment of the fuel pins carrying the gadolinium. As explained above, those pins will have a lower fissile enrichment compared to the non-BP fuel pins to assist in the licensing of the gadolinium pins, i. e., lower thermal conductivity, and they should not be limiting, considering the lack of irradiation data and fuel performance validation. This will mean that to offset the lower enrichments in these BP fuel pins, the other pins will need to have their enrichment increased correspondingly. The larger the number of gadolinium rods in the core, the greater the level of additional enrichment required in the other fuel pins; thus achieving as long an operating cycle as possible, which requires average fuel enrichments as high as possible, could become in conflict with the current licensing limit of 5 wt% U-235.

In large PWRs operating today, there is extensive experience with different BP types and loadings, as well as control rod sequences and calculation of control rod worths. However, the combination and extent of some of these does result in challenges for code predictions, and the associated validation of nuclear design tools in extremely heterogeneous cores, and regions of notable flux variation, in particular ensuring that BP and control rod worths are accurately predicted. Furthermore, with numerous axial zones in the fuel designs, and with significant control rod movements, advances in nodal methods may be required, including the need for variable size meshes to accommodate some of the nuclear designs.