Neutronic analysis for accident tolerant cladding candidates in CANDU-6 reactors

Naceur, Ahmed; Marleau, G.

doi:10.1016/j.anucene.2017.11.016

Cited by 36 publications

(7 citation statements)

References 13 publications

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“…For maintaining the safety standards of advanced PWRs, it is necessary that each cladding material produces negative fuel temperature coe cients especially at the earlier stages of fuel irradiation. For estimating the FTC, the temperature of moderator and clad was xed at 600 K but the fuel temperature is increased from 900 K (operational temperature) to 1200 K i.e., temperature change is 300 K. As can be noted in gure (15), the FTC values for all cases are less negative at the BOL (0 GWd/ton) and more negative as the burnup increases to about 15 GWd/ton due to the changes in isotopic composition and the increased variety of isotopes present. Table (7) shows that the BOL FTC values are negative for all the candidate claddings.…”

Section: Fuel Temperature Coe Cientmentioning

confidence: 99%

“…A. Naceur and G. Marleau., [15] compared the neutronic performance of (SS-304 and SS-310), (FeCrAl and APTAM) and SiC with Zirconium alloy using burnup calculations, isotope depletion analyses, absorption rate ratios, spectral and spatial self-shielding analysis, plutonium radial profile and reactivity perturbations. Their calculations are based on the 3D lattice transport-theory code DRAGON5 which is part of the Industry Standard Tool-set (IST) of Codes for CANDU reactor core design and analysis.…”

Section: Introductionmentioning

confidence: 99%

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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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Section: Fuel Temperature Coe Cientmentioning

confidence: 99%

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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“…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. Naceur and Marleau 15 compared the neutronic performance of (SS-304 and SS-310), (FeCrAl and APTAM) and SiC with Zirconium alloy using burnup calculations, isotope depletion analyses, absorption rate ratios, spectral and spatial self-shielding analysis, plutonium radial profile and reactivity perturbations. Their calculations are based on the 3D lattice transport-theory code DRAGON5 which is part of the Industry Standard Tool-set (IST) of Codes for CANDU reactor core design and analysis.…”

Section: Introductionmentioning

confidence: 99%

Investigation of the safety features of advanced PWR assembly using SiC, Zr, FeCrAl and SS-310 as cladding materials

Mustafa

2021

Sci Rep

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In this work, SiC (Silicon carbide), FeCrAl (ferritic), SS-310 (stainless steel 310) and Zirconium are simulated by MCNPX (Monte Carlo N‐Particle eXtended) code as cladding materials in advanced PWR (Pressurized Water Reactor) 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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“…Other recent studies (Wu et al, 2015, Terrani et al, 2014, Pint et al, 2013 performed neutronics and fuel performance assessments of the ATF concept, which was limited only to advanced oxidation-resistant iron-based claddings with UO 2 fuel. There is also a recent study (Naceur and Marleau, 2018) conducted byÉcole Polytechnique de Montréal on potential accident tolerant cladding concepts in a CANDU-6 (Canada Deuterium Uranium) reactor, where the analyses were also limited to the UO 2 fuel only. Furthermore, a Brookhaven National Laboratory team performed an extensive screening of advanced ATF concepts and neutronic evaluations of UN-U 3 Si 5 relative to the UO 2 reference fuel with ATF claddings (Brown 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

Alam

Goodwin²,

Parks

2019

Progress in Nuclear Energy

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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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Neutronic analysis for accident tolerant cladding candidates in CANDU-6 reactors

Cited by 36 publications

References 13 publications

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

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

Investigation of the safety features of advanced PWR assembly using SiC, Zr, FeCrAl and SS-310 as cladding materials

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

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