Microheterogeneous Thoria-Urania Fuels for Pressurized Water Reactors

Shwageraus, Eugene; Zhao, X; Driscoll, Michael J.; Hejzlar, Pavel; Kazimi, Mujid S.; Herring, J.S.

doi:10.13182/nt04-a3512

Cited by 32 publications

(42 citation statements)

References 5 publications

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“…The final strategy is a micro-heterogeneous concept, which is a compromise position between homogeneous and heterogeneous approaches and can be achieved in micro-heterogeneous ThO 2 -UO 2 duplex fuel. In this concept, the uranium and thorium components are not blended and are discretely interspersed on very small distance scales (Zhao, 2001, Shwageraus et al, 2004, Otto, 2013. Studies (Zhao, 2001, Shwageraus et al, 2004, Alam, 2018 found that the spatial separation between the fissile 235 U and the fertile thorium enables better moderation of the fission neutrons before they interact with the thorium.…”

Section: Rationale Behind the Selection Of The Duplex Fuelmentioning

confidence: 99%

“…In this concept, the uranium and thorium components are not blended and are discretely interspersed on very small distance scales (Zhao, 2001, Shwageraus et al, 2004, Otto, 2013. Studies (Zhao, 2001, Shwageraus et al, 2004, Alam, 2018 found that the spatial separation between the fissile 235 U and the fertile thorium enables better moderation of the fission neutrons before they interact with the thorium. This spectral shift encourages neutron absorption in the thorium component of the fuel at the beginning of life, helps controlling reactivity and promotes breeding of new fissile material 233 U. Thorium's advantages are best realized in micro-heterogeneous ThO 2 -UO 2 and heterogeneous seed-blanket geometries (MacDonald and Lee, 2004), although, as discussed, heterogeneous arrangements will not be considered in this study.…”

Section: Rationale Behind the Selection Of The Duplex Fuelmentioning

confidence: 99%

“…This spectral shift encourages neutron absorption in the thorium component of the fuel at the beginning of life, helps controlling reactivity and promotes breeding of new fissile material 233 U. Thorium's advantages are best realized in micro-heterogeneous ThO 2 -UO 2 and heterogeneous seed-blanket geometries (MacDonald and Lee, 2004), although, as discussed, heterogeneous arrangements will not be considered in this study. A limited number of studies on duplex fuel are available in the public domain, but its use has never been examined in the context of a SBF environment for long-life core (Zhao, 2001, Shwageraus et al, 2004, Alam, 2018, MacDonald and Lee, 2004. Therefore, this study will utilize duplex fuel for the ATF analysis.…”

Section: Rationale Behind the Selection Of The Duplex Fuelmentioning

confidence: 99%

“…This originates from the different magnitude of cladding captures with the candidate fuel (mathematical ratio of the total clad capture between duplex to UO 2 ) (Alam, 2018). The interference of the fertile resonances of duplex fuel (simultaneous presence of 238 U and 232 Th) with the resonances of cladding isotopes (resonance overlap) can lead to a higher overall cladding capture rate for duplex fuel than for UO 2 fuel (Zhao, 2001, Shwageraus et al, 2004. Fig.…”

Section: Reactivitymentioning

confidence: 99%

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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

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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Section: Rationale Behind the Selection Of The Duplex Fuelmentioning

confidence: 99%

Section: Rationale Behind the Selection Of The Duplex Fuelmentioning

confidence: 99%

Section: Rationale Behind the Selection Of The Duplex Fuelmentioning

confidence: 99%

Section: Reactivitymentioning

confidence: 99%

See 2 more Smart Citations

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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“…The main mechanism by which this is achieved is physical separation which reduces the competition for neutrons between Th and Pu, i.e. Th has a better chance of absorbing more neutrons to produce 233 U without competition from Pu isotopes (Shwageraus et al, 2004b).…”

Section: Introductionmentioning

confidence: 99%

Assembly-level analysis of heterogeneous Th–Pu PWR fuel

Zainuddin

Parks

Shwageraus

2017

Annals of Nuclear Energy

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This study compares homogeneous and heterogeneous thorium-plutonium (Th-Pu) fuel assemblies (with high Pu content-20 wt%), and examines whether there is an increase in Pu incineration in the latter. A seed-blanket configuration based on the Radkowsky thorium reactor concept is used for the heterogeneous assembly. This separates the thorium blanket from the uranium seed, or in this case a plutonium seed. The seed supplies neutrons to the subcritical thorium blanket which encourages the in-situ breeding and burning of 233 U, allowing the fuel to stay critical for longer, extending burnup of the fuel. While past work on Th-Pu seed-blanket units shows superior Pu incineration compared to conventional U-Pu mixed oxide fuel, there is no literature to date that directly compares the performance of homogeneous and heterogeneous Th-Pu assembly configurations. Use of exactly the same fuel loading for both configurations allows the effects of spatial separation to be fully understood. It was found that the homogeneous fuel with and without burnable poisons were able to achieve much higher Pu incinerations than the heterogeneous fuel configurations, while still attaining a reasonably high discharge burnup. This is because in the heterogeneous cases, 233 U breeding is faster, thereby contributing to a much larger fraction of total power produced by the assembly. In contrast, 233 U build-up is slower in the homogeneous case and therefore Pu burning is greater. This 233 U begins to contribute a significant fraction of power produced only towards the end of life, thus extending criticality, allowing more Pu to burn.

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Low enriched uranium and thorium fuel utilization under once‐through and offline reprocessing scenarios in small modular molten salt reactor

et al. 2019

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Summary Using low enrichment uranium as driver fuel under once‐through mode in molten salt reactor (MSR) attracts more and more attention because of its fuel availability, no new technology, and nuclear nonproliferation. It is regarded as a wise research and development road to shorten deployment time of MSRs and to prepare techniques and experiences for thorium‐uranium breeding of MSRs in the future. However, this fuel management is still faced with some different technical routes, such as the selection of carrier salts, the enrichment of uranium, with or without thorium, and the recycle necessary of spent nuclear fuel. Therefore, various fuel cycle schemes were compared and analyzed using an in‐house developed fuel management code MOBAT. Different graphite assemblies were optimized by changing the salt volume fraction in core and dimension to find a region for best fuel utilization and negative temperature reactivity coefficient. Prismatic block with 10% volume fraction of molten salt is considered as a good assembly type because of its significant space shielding effect of U‐238. For carrier salts, LiF‐BeF2 with 99.995% enriched Li‐7 displays higher fuel utilization and lower cost of fuel cycle than NaF‐BeF2, while the tritium production at the beginning of life will be two orders of magnitude higher. For fuel enrichment, 20% enriched uranium is recommended because the background of neutron absorption from carrier salt and graphite will be more significant in lower enrichment condition. Importantly, it shows that thorium is a good breed and burned fuel in situ and could improve the fuel utilization by 20%. Also, offline reprocessing to recover the uranium is a commendable scheme when the cost of offline reprocessing is lower than 400 $/kgHN.

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Microheterogeneous Thoria-Urania Fuels for Pressurized Water Reactors

Cited by 32 publications

References 5 publications

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

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

Assembly-level analysis of heterogeneous Th–Pu PWR fuel

Low enriched uranium and thorium fuel utilization under once‐through and offline reprocessing scenarios in small modular molten salt reactor

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