Coupled neutronic/thermal-hydraulic hot channel analysis of high power density civil marine SMR cores

Abstract

Core average power density of standard small modular reactors (SMR) are generally limited to 60-65 MW/m 3 , which is 40% lower than for a standard civil PWR in order to accommodate better thermal margins. While designing a SMR core for civil marine propulsion systems, it is required to increase its power density to make more attractive for future deployment. However, there are obvious thermal-hydraulic (TH) concerns regarding a high power density (HPD) core, which needs to be satisfied in order to ensure safe operation through accurate prediction of the TH parameters. This paper presents a coupled neutronic/thermal-hydraulic (TH) hot channel analysis of a HPD 375 MWth soluble-boron-free PWR core using 19.25% 235 U enriched microheterogeneous ThO 2-UO 2 duplex fuel and 16% 235 U enriched homogeneously mixed all-UO 2 fuel with a 15 effective full-power-years (EFPY) core life. To perform this analysis the hybrid Monte Carlo reactor physics code MONK is coupled with sub-channel analysis TH code COBRA-EN. This approach is used to investigate the feasibility of different HPD marine PWR concepts and to identify the main TH challenges characterising these designs. To design HPD cores of between 82 and 111 MW/m 3 , three cases were chosen by optimizing the fuel pin diameter, pin pitch and pitch-to-diameter ratio. These cases have been studied to determine whether TH safety limits are satisfied by evaluating key parameters, such as minimum departure from nucleate boiling ratio, surface heat flux, critical heat flux, cladding inner surface and fuel centreline temperatures, and pressure drop. The results show that it is possible to achieve a core power density of 100 MW/m 3 for both the candidate fuels, a ∼50% improvement on the reference design (63 MW/m 3), while meeting the target core lifetime of 15 EFPY and remaining within TH limits. The size of the pressure vessel can therefore be reduced substantially and the economic competitiveness of the proposed civil marine PWR reactor core significantly improved.

Review of the microheterogeneous thoria‐urania fuel for micro‐sized high temperature reactors

“…A higher enrichment was expected due to the lower UO 2 volume and the higher thermal neutron absorption of fertile 232 Th 11 . The use of duplex pellet designs also improves the ratio of neutron capture in fertile material to neutron absorption in fissile material which leads to an extended fuel burnup and core lifetime 11,57,58 . This has been demonstrated by previous work on duplex pellet utilization in PWR combined with axial pellet arrangement and checkboard rod configuration, which shows a 15% increase in burnup 14 .…”

“…This indicates a higher breeding capacity, smaller core reactivity swing and longer core cycle 11 . The duplex is also used for the design of the SMR core with a high power density of 100 MW/m 3 58 …”

Neutronic feasibility of civil marine small modular reactor core using mixed D2O+H2O coolant

“…• High power density SMR cores (Alam et al, 2019f) experience a core life penalty due to the associated reduction in fuel mass. Since this mixed coolant exhibits promise in offering longer core lifetime than the H 2 O coolant (Alam et al, 2019c), it has the potential to compensate this core life penalty.…”

Small modular reactor core design for civil marine propulsion using micro-heterogeneous duplex fuel. Part II: whole-core analysis

“…The major challenge for the micro-heterogeneous duplex fuel arrangements is to meet the thermal-hydraulic margins since the most severe situation will be in the duplex pellet case where most of the power is generated in the UO 2 part of the fuel pellet. In order to confirm that all the thermal-hydraulic constraints are satisfied for the duplex fuel, 3D neutronic/thermal-hydraulic coupling of hybrid monte carlo MONK with sub-channel analysis COBRA-EN (Basile, 1999) code for hot channel analysis (Alam et al, 2019d) has been performed to evaluate key TH parameters such as: minimum departure from nucleate boiling ratio (MDNBR), heat flux, cladding, inner surface and fuel centreline temperatures, and pressure drop. Our study confirmed that thermal-hydraulic design requirements for the duplex fuel can be met and there will be no melting in the UO 2 region of the duplex pellet.…”