RITM-200: new-generation reactor for a new nuclear icebreaker

Annals of Nuclear Energy

Oliveira

Goodwin³

et al. 2019

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.

Section: Design Goals and Constraintsmentioning

confidence: 99%

“…OK-900, have employed with more than 20% 235 U enrichment (Bukharin, 2006), but a new generation of icebreaker cores (e.g. are reported to use less than 20% 235 U (Zverev et al, 2013) enrichment.…”

Section: Introductionmentioning

confidence: 99%

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

Annals of Nuclear Energy

Oliveira

Goodwin³

et al. 2019

“…OK-900, have utilised a cermet fuel with uranium enriched to more than 20% 235 U (Bukharin, 2006), but a new generation of icebreaker cores (e.g. are reported to use cermet fuel consisting of zirconium-based alloy host material embedded with UO 2 particles enriched to less than 20% 235 U (Zverev et al, 2013). These reactors will have a relatively low maximum power output of 174 MWth, a low capacity factor (around 65%) and require refueling every 7 years (Bukharin, 2006, Prasad et al, 2015.…”

Section: Introductionmentioning

confidence: 99%

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

Nuclear Engineering and Design

Ridwan

Kumar

et al. 2019

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.

“…OK-900, have utilised a cermet fuel with uranium enriched to more than 20% 235 U (Bukharin, 2006), but a new generation of icebreaker cores (e.g. are reported to use cermet fuel consisting of zirconium-based alloy host material embedded with UO 2 particles enriched to less than 20% 235 U (Zverev et al, 2013). The reactor will have a relatively low maximum power output of 174 MWth, a low capacity factor (around 65%) and require refueling every 7 years (Bukharin, 2006, Prasad et al, 2015.…”

Section: Introductionmentioning

confidence: 99%

Assembly-level analyses of accident-tolerant cladding concepts for a long-life civil marine SMR core using micro-heterogeneous duplex fuel

Progress in Nuclear Energy

Goodwin²,

Parks

2019

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.