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

Alam, Syed Bahauddin; Goodwin, Cameron S.; Parks, Geoffrey T.

doi:10.1016/j.pnucene.2018.10.011

Cited by 33 publications

(12 citation statements)

References 26 publications

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“…In addition to satisfying the neutronics and TH requirements and constraints, additional burnup requirement needs to be satisfied. Cladding oxide thickness of 100 μm currently limits discharge burnup up to 62 GWd/tHM, and therefore new accident-tolerant cladding will be required (Alam et al, 2019(Alam et al, , 2018a. Fission gas release is limited to a few percents for up to 50-60 GWd/tHM, but can easily exceed 30% for the higher burnup.…”

Section: Discussion On Practical Considerations For the Duplex Fuel mentioning

confidence: 99%

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

Alam

Oliveira

Goodwin³

et al. 2019

Annals of Nuclear Energy

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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.

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Section: Discussion On Practical Considerations For the Duplex Fuel mentioning

confidence: 99%

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

Alam

Oliveira

Goodwin³

et al. 2019

Annals of Nuclear Energy

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“…In this study, we have used deterministic reactor physics code WIMS 10 20,21 for the lattice-level calculation while considering JEF 2.2 nuclear data library and 6 energy group structure 9–12,17,22–24 . WIMS performs deterministic neutron transport calculations for every pin in a fuel assembly using an established calculation route through sequence of separate modules 19,20,22,25,26 .…”

Section: Design and Calculational Methodsmentioning

confidence: 99%

“…For the analyses of alternative burnable poison, this study considers 15% U-235 enriched homogeneously mixed UO 2 fuel enriched in in a 13 × 13 assembly while assuming 7.5% neutron leakage and discharge burnup limit of 100 GWd/tonne in the WIMS model 22 . The detailed computational and design methods are reported in our previous studies 9–12,17,22,30 . Table 1 shows the 2D lattice-level (WIMS) and 3D whole-core (PANTHER) system parameter values 12,19,30–38 .…”

Section: Design and Calculational Methodsmentioning

confidence: 99%

“…We set a target operational life of at least 15 years under high irradiation of around 100 GWd/tonne and constrain uranium enrichment at 20%– in line with proliferation resistance related limits on low-enriched uranium (LEU) 9–12 . It is noteworthy to mention that reactor cores for the purpose of civil marine applications cannot be conceivably as highly enriched as the reactor cores used in military vessels 9,10,17 . Civil shipping reactors will require a LEU core design but should also have a long operating lifetime.…”

Section: Introductionmentioning

confidence: 99%

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Neutronic investigation of alternative & composite burnable poisons for the soluble-boron-free and long life civil marine small modular reactor cores

Alam

Almutairi

Ridwan

et al. 2019

Sci Rep

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Concerns about the effects of global warming provide a strong case to consider how best nuclear power could be applied to marine propulsion. Currently, there are persistent efforts worldwide to combat global warming, and that also includes the commercial freight shipping sector. In an effort to decarbonize the marine sector, there are growing interests in replacing the contemporary, traditional propulsion systems with nuclear propulsion systems. The latter system allows freight ships to have longer intervals before refueling; subsequently, lower fuel costs, and minimal carbon emissions. Nonetheless, nuclear propulsion systems have remained largely confined to military vessels. It is highly desirable that a civil marine core not use soluble boron for reactivity control, but it is then a challenge to achieve an adequate shutdown margin throughout the core life while maintaining reactivity control and acceptable power distributions in the core. High-thickness ZrB2 150 μm Integral Fuel Burnable Absorber (IFBA) is an excellent burnable poison (BP) candidate for long life soluble-boron-free core. However, in this study, we want to minimize the use of 150 μm IFBA since B-10 undergoes an (n, α) capture reaction, and the resulting helium raises the pressure within the plenum and in the cladding. Therefore, we have considered several alternative and novel burnable BP design strategies to minimize the use of IFBA for reactivity control in this study: (Case 1) a composite BP: gadolinia (Gd2O3) or erbia (Er2O3) with 150 μm thickness ZrB2 IFBA; (Case 2) Pu-240 or Am-241 mixed homogeneously with the fuel; and (Case 3) another composite BP: Pu-240 or Am-241 with 150 μm thickness ZrB2 IFBA. The results are compared against those for a high-thickness 150 μm 25 IFBA pins design from a previous study. The high-thickness 150 μm 25 IFBA pins design is termed the “IFBA-only” BP design throughout this study. We arrive at a design using 15% U-235 fuel loaded into 13 × 13 assemblies with Case 3 BPs (IFBA+Pu-240 or IFBA+Am-241) for reactivity control while reducing 20% IFBA use. This design exhibits lower assembly reactivity swing and minimal burnup penalty due to the self-shielding effect. Case 3 provides ~10% more initial (beginning-of-life) reactivity suppression with ~70% less reactivity swing compared to the IFBA-only design for UO2 fuel while achieving almost the same core lifetime. Finally, optimized Case 3 assemblies were loaded in 3D nodal diffusion and reactor model code. The results obtained from the 3D reactor model confirmed that the designed core with the proposed Case 3 BPs can achieve the target lifetime of 15 years while contributing to ~10% higher BOL reactivity suppression, ~70% lower reactivity swings, ~30% lower radial form factor and ~28% lower total peaking factor compared to the IFBA-only core.

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“…Several researchers have studied the neutronic performance of the cladding materials under consideration. S.B.Allam et al, [14] have calculated reactivity, radial power distribution and spectral hardening for small modular reactor assembly using micro-heterogeneous duplex fuel. They performed their calculations by WIMS-10 Lattice physics code (deterministic method) using nuclear data from the JEF 2.2 database available from the IAEA.…”

Section: Introductionmentioning

confidence: 99%

Investigation of the Safety Features of Advanced PWR Assembly Using SiC, Zr, FeCrAl and SS-310 As Cladding Materials

Mustafa

2020

Preprint

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In this work, SiC (Silicon carbide), FeCrAl (ferritic), SS-310 (stainless steel 310) and Zirconium are simulated by MCNPX code as cladding materials in advanced PWR assembly. A number of reactor safety parameters are evaluated for the candidate cladding materials as reactivity, cycle length, radial power distribution of fuel pellet, reactivity coefficients, spectral hardening, peaking factor, thermal neutron fraction and delayed neutron fraction. The neutron economy presented by Zr and SiC models is analyzed through the burnup calculations on the unit cell and assembly levels. The study also provided the geometric conditions of all cladding materials under consideration in terms of the relation between fuel enrichment and cladding thickness from the viewpoint to achieve the same discharge burnup as the Zircaloy cladding. It was found that the SiC model participated in extending the life cycle by 2.23% compared to Zr. The materials other than SiC largely decreased discharge burnup in comparison with Zircaloy. Furthermore, the claddings with lower capture cross-sections (SiC and Zr) exhibit higher relative fission power at the pellet periphery. The simulation also showed that using SiC with a thickness of 571.15 µm and 4.83% U-235 can satisfy the EOL irradiation value as Zr. For reactivity coefficient, the higher absorbing materials (SS-310 and FeCrAl) exhibit more negative FTCs, MTCs and VRCs at the BOL But, at the intermediate stages of burnup Zr and SiC have a strong trend of negative reactivity coefficients. Finally, the delayed neutron fraction of SiC and Zr models is the highest among all the four models.

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

Cited by 33 publications

References 26 publications

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

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

Neutronic investigation of alternative & composite burnable poisons for the soluble-boron-free and long life civil marine small modular reactor cores

Investigation of the Safety Features of Advanced PWR Assembly Using SiC, Zr, FeCrAl and SS-310 As Cladding Materials

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