Civilian marine reactors face a unique set of design challenges in addition to the usual irradiation and thermal-hydraulic limits affecting all reactors. These include requirements for a small core size, long core lifetime, a 20% cap on fissile loading, and limitations on the use of soluble boron. One way to achieve higher burnup/longer core life is to alter the neutron spectrum by changing the hydrogen-to-heavy-metal ratio, thus increasing the conversion of fertile isotopes in the fuel. In this reactor physics study, we optimize the two-dimensional lattice geometry of a 333 MWth soluble-boron-free marine PWR for 18% 235 U enriched micro-heterogeneous ThO 2 -UO 2 duplex fuel and 15% 235 U enriched homogeneously mixed all-UO 2 fuel. We consider two types of coolant: H 2 O and mixed 80% D 2 O + 20% H 2 O. We aim to observe in which spectrum discharge burnup is maximized in order to improve uranium utilization, while satisfying the constraint on moderator temperature coefficient. It is observed that higher discharge burnup for the candidate fuels is achievable by using either a wetter lattice or a much drier lattice than normal, while epithermal lattices are distinctly inferior performers. The thorium-rich duplex fuel exhibits higher discharge burnup potential than the all-UO 2 fuel for all moderation regimes for both coolants. The candidate fuels exhibit higher initial reactivity and discharge burnup with the mixed D 2 O-H 2 O coolant than with the H 2 O coolant in the under-moderated regime, whereas these values are lower for the D 2 O-H 2 O coolant in the over-moderated regime.

Civil marine reactors face a unique set of design challenges. These include requirements for a small core size and long core lifetime, a 20% cap on fissile loading, and limitations on using soluble neutron absorbers. In this reactor physics study, we seek to design a core that meets these requirements over a 15 effective full-power-years (EFPY) life at 333 MWth using homogeneously mixed all-UO 2 and micro-heterogeneous ThO 2 -UO 2 duplex fuels. In a companion (Part I) paper, we found assembly designs using 15% and 18% 235 U for UO 2 and duplex fuels, respectively, loaded into 13×13 pin arrays. High thickness (150 μm) ZrB 2 integral fuel burnable absorber (IFBA) pins and boron carbide (B 4 C) control rods are used for reactivity control. Taking advantage of self-shielding effects, these designs maintain low and stable assembly reactivity with little burnup penalty.In this paper (Part II), whole-core design analyses are performed for small modular reactor (SMR) to determine whether the core remains critical for at least 15 EFPY with a reactivity swing of less than 4000 pcm, subject to appropriate constraints. The main challenge is to keep the radial form factor below its limit (1.50). Burnable poison radial-zoning is examined in the quest for a suitable arrangement to control power peaking. Optimized assemblies are loaded into a 3D reactor model in PANTHER. The PANTHER results confirm that the fissile loadings of both fuels are well-designed for the target lifetime: at the end of the ∼15-year cycle, the cores are on the border of criticality. The duplex fuel core can achieve ∼4% longer core life, has a ∼3% lower initial reactivity and ∼30% lower reactivity swing over life than the final UO 2 core design. The duplex core is therefore the more successful design, giving a core life of ∼16 years and a reactivity swing of less than 2500 pcm, while satisfying all the neutronic safety parameters. In particular, one of the major objectives of this study is to offer/explore a thorium-based candidate alternative fuel platform for the proposed marine core. It is proven by literature reviews that the ability of the duplex fuel was never explored in the context of a single-batch, LEU, SBF, long-life SMR core. In this regard, the motivation of this paper is to understand the underlying physics of the duplex fuel and 'open the option' of designing the functional cores with both the duplex and UO 2 fuel cores.

In an e↵ort to de-carbonise commercial freight shipping, there is growing interest in the possibility of using nuclear propulsion systems. In this reactor physics study, we seek to design a soluble-boron-free (SBF) and low-enriched uranium (LEU) (<20% 235 U enrichment) civil nuclear marine propulsion small modular reactor (SMR) core that provides at least 15 e↵ective full-power-years (EFPY) life at 333 MWth using 18% 235 U enriched micro-heterogeneous ThO 2 -UO 2 duplex fuel and 15% 235 U enriched homogeneously mixed all-UO 2 fuel. We use WIMS to develop subassembly designs and PANTHER to examine whole-core arrangements.The assembly-level behaviours of candidate burnable poison (BP) materials and control rods are investigated. We examine gadolinia (Gd 2 O 3 ), erbia (Er 2 O 3 ) and ZrB 2 integral fuel burnable absorber (IFBA) as BPs. We arrive at a design with the candidate fuels loaded into 13⇥13 assemblies using IFBA pins for reactivity control. Taking advantage of self-shielding e↵ects, this design maintains low and stable assembly reactivity with relatively little burnup penalty. Thorium-based duplex fuel o↵ers better performance than all-UO 2 fuel with all BP options considered. Duplex fuel has ⇠20% lower reactivity swing and, in consequence, lower initial reactivity than all-UO 2 fuel. The lower initial reactivity and smaller reactivity swing make the task of reactivity control through BP design easier in the thorium-rich duplex core. For control rod design, we examine boron carbide (B 4 C), hafnium, and Ag-In-Cd alloy. All the candidate materials exhibit greater rod worth for the duplex design. For both fuels, B 4 C has the highest rod worth. In particular, one of the major objectives of this study is to o↵er/explore a thorium-based candidate alternative fuel platform for the proposed marine core. It is proven by literature reviews that the ability of the duplex fuel was never explored in the context of a single-batch, LEU, SBF, long-life SMR core. In this regard, the motivation of this paper is to observe the neutronic performance of the proposed duplex fuel with respect to the UO 2 fuel and 'open the option' of designing the functional cores with both the duplex and UO 2 fuel cores.A companion paper will examine key physics and core safety analysis parameters in the whole-core environment.

In this reactor physics study, we examine the neutronic performance of accident-tolerant fuel (ATF) claddings-austenitic type 310 stainless steel (310SS), ferritic Fe-20Cr-5Al (FeCrAl), advanced powder metallurgic ferritic (APMT), and silicon carbide (SiC)-based materials-as alternative cladding materials compared with Zircaloy-4 (Zr) cladding. The cores considered use 18% 235 U enriched micro-heterogeneous ThO 2-UO 2 duplex fuel and, for purposes of comparison, 15% 235 U enriched homogeneously mixed all-UO 2 fuel, loaded into 13×13 pin arrays. A constant cladding coating thickness of 655 μm is assumed. We use the WIMS reactor physics code to analyse the associated reactivity, achievable discharge burnup, spectral variations, rim effect and reactivity feedback parameters for the candidate cladding materials at the assembly level. The results show that candidate fuels with 310SS cladding exhibit a ∼13% discharge burnup penalty compared to Zr due to the presence of a very high nickel (Ni) concentration. The high neutron absorption cross-sections of iron (Fe) in the FeCrAl and APMT claddings also lead to a ∼10% discharge burnup penalty. The fuels with SiC cladding can achieve a ∼1% higher discharge burnup compared to Zr due to the low thermal neutron absorption cross-sections of its constituents and the softer neutron spectrum. The claddings with lower capture cross-sections (SiC and Zr) exhibit higher relative fission power at the pellet periphery. For both candidate fuels, the end-of-life 239 Pu (for UO 2 fuel) and 233 U (for duplex fuel) inventories are higher for the claddings (Fe-based: FeCrAl, APMT and steel-based: 310SS) with higher thermal capture cross-sections, unlike for SiC and Zr, where SiC provides higher end-of-life 239 Pu and 233 U inventories despite having lower capture cross-section than that of the Zr. Reactivity feedback parameter values (moderator and fuel temperature coefficients) are more negative for the duplex fuel than the UO 2 fuel for all the candidate claddings, with claddings with harder spectra exhibiting more negative values. The duplex fuel yields a softer spectrum than the UO 2 fuel with the candidate claddings, which improves neutron economy and thus discharge burnup.

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