Comparative Studies on Plutonium and Minor Actinides Utilization in Small Molten Salt Reactors with Various Powers and Core Sizes

Waris, Abdul; Aji, Indarta Kuncoro; Pramuditya, Syeilendra; Novitrian, Novitrian; Permana, Sidik; Su’ud, Zaki

doi:10.1016/j.egypro.2014.11.855

Cited by 27 publications

(13 citation statements)

References 9 publications

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“…LEU enrichment is set at 20%, to minimise 238 U content in the fuel. Plutonium and MA vectors are taken from reference [9]. As for PuF 3 +(MA)F 3 , the Pu:MA ratio is set at 9:1.…”

Section: General Descriptionmentioning

confidence: 99%

“…The latter is often considered as a potential proliferation threat, despite plutonium isotopes in LWR spent fuel is degraded so much that it is unsuitable for military diversion. Burning plutonium in MSR can help eliminating the perceived threat [9].…”

Section: Introduction *mentioning

confidence: 99%

“…This is important since, in term of fuel utilisation, both 235 U and 239 Pu are inferior than that of 233 U in thermal spectrum. It is also suggested that, to prevent proliferation potential even further, that plutonium isotopes are not to be separated from its accompanying minor actinides (MA) [9,12]. Thus, it is also possible that the plutonium used as startup fuel of MSR contains MA.…”

Section: Introduction *mentioning

confidence: 99%

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Investigation on Inherent Safety of One Fluid-Molten Salt Reactor (Of-Msr) With Various Starting Fuel

Dwijayanto

Hermawan

2020

Tri Dasa Mega

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Molten salt reactor (MSR) is often associated with thorium fuel cycle, thanks to its excellent neutron economy and online reprocessing capability. However, since 233U, the fissile used in pure thorium fuel cycle, is not commercially available, the MSR must be started with other fissile nuclides. Different fissile yields different inherent safety characteristics, and thus must be assessed accordingly. This paper investigates the inherent safety aspects of one fluid MSR (OF-MSR) using various fissile fuel, namely low-enriched uranium (LEU), reactor grade plutonium (RGPu), and reactor grade plutonium + minor actinides (PuMA). The calculation was performed using MCNPX2.6.0 programme with ENDF/B-VII library. Parameters assessed are temperature coefficient of reactivity (TCR) and void coefficient of reactivity (VCR). The result shows that TCR for LEU, RGPu, and PuMA are -3.13 pcm, -2.02 pcm and -1.79 pcm, respectively. Meanwhile, the VCR is negative only for LEU, whilst RGPu and PuMA suffer from positive void reactivity. Therefore, for the OF-MSR design used in this study, LEU is the only safe option as OF-MSR starting fuel.Keywords: MSR, Temperature coefficient of reactivity, Void coefficient of reactivity, Low enriched uranium, Reactor grade plutonium, Minor actinides

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“…LEU enrichment is set at 20%, to minimise 238 U content in the fuel. Plutonium and MA vectors are taken from reference [9]. As for PuF 3 +(MA)F 3 , the Pu:MA ratio is set at 9:1.…”

Section: General Descriptionmentioning

confidence: 99%

Section: Introduction *mentioning

confidence: 99%

Section: Introduction *mentioning

confidence: 99%

See 1 more Smart Citation

Investigation on Inherent Safety of One Fluid-Molten Salt Reactor (Of-Msr) With Various Starting Fuel

Dwijayanto

Hermawan

2020

Tri Dasa Mega

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“…Many studies have considered transuranic (TRU) waste incineration using existing technologies (Galahom, 2018) (Sahin, et al, 2012a) (Sahin, et al, 2012b), while others have considered advanced options, including but not limited to: Fast Reactors (FRs) (Wallenius & Bortot, 2018) (You & Hong, 2016) (Wider, et al, 2014) (Tucek, et al, 2008) (Bays, 2007); Molten Salt Reactors (Waris, et al, 2015) (Heuer, et al, 2014); Small Modular Reactors (Zou, et al, 2018); Pebble Bed High Temperature Reactors (Acir & Coskun, 2015); High Temperature Gas-cooled Reactors (Lennox, et al, 2007); and Accelerator Driven Systems (Al Qaaod, et al, 2018). However, it has been determined that, for the purposes of UK , 1998).…”

Section: Viable Technologies For Pu Dispositionmentioning

confidence: 99%

The effect of Am241 on UK plutonium recycle options in thorium-plutonium fuelled LWRs – Part I: PWRs

Morrison

Parks

2020

Annals of Nuclear Energy

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UK plutonium is expected to be managed using uranium-plutonium (U-Pu) mixed oxide (MOX) fuels in Light Water Reactors (LWRs). However, studies have shown that thorium-plutonium (Th-Pu) may be preferential. Research has mostly focussed on recycle of reactor grade Pu with limited minor actinide (MA) content. This study will determine if large quantities of americium (Am) in UK Pu may be restrictive to recycle schemes by determining the effect this has on reactivity feedback coefficients, fissile loading and incineration potential. Addition of Am is shown to result in predictable trends in reactivity feedback coefficients and spatial separation of Am and Pu is found to offer potential advantages over uniformly loaded fuel in terms of maximising fissile loading and incineration. Separation may also offer benefits in terms of targeting Am destruction, particularly if multiple recycle schemes are pursued, as this would maximise the fissile loading requirements while keeping reactivity feedback coefficients negative.

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“…One of the advanced nuclear reactors that implemented the natural circulation system to circulate its fuel as a passive safety design is the Molten Salt Reactor 14,15 . As reported in papers, 16‐19 the fuel of Molten Salt Reactor is molten salt from a mixture of LiF‐BeF 2 ‐ThF 4 ‐(PuMA)F 4 or LiF‐BeF 2 ‐ThF 4 ‐ 233 UF 4 . Where, Thorium fuel ( 232 Th) provides an advantage for safety factors in reactor operations related to the void condition 20‐23 …”

Section: Introductionmentioning

confidence: 99%

Experimental and simulation approach of the loop geometry effect on the natural circulation system of the advanced nuclear reactor

2020

Self Cite

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Summary Lately, natural circulation has become an exciting topic because it has been proposed and applied in some advanced nuclear reactors as a passive safety system. The aim of the present study is to examine the effect of the loop geometry on the natural circulation based cooling system of a given power source. The experimental approach was achieved by the construction of two natural circulation loops with vertical heater and vertical cooler. Both circulation loops were built with a same vertical length of 100 cm and different horizontal lengths, each of 50 and 100 cm, respectively. The heater was realized with a stainless pipe wrapped by a Nichrome wire, whereas the cooling system was designed and built as a pipe in 30 cm long square tube. Thus, the cooling water will circulate around the pipe loop inside the rectangular cooling chamber. The temperature data acquisition was achieved via an Arduino‐based system, controlling four K‐type thermocouple sensors. The numerical simulation part in the present work, was carried out by the COMSOL Multiphysics software, using the dedicated heat transfer and fluid dynamic module, namely the computational fluid dynamic module. The experimental data were considered to parameterize the COMSOL input model. The experimental results showed a clear difference between two passive cooling loops according to the horizontal length. The doublings of the horizontal and vertical lengths of the loop have a direct effect on the fluid flow rate.

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Comparative Studies on Plutonium and Minor Actinides Utilization in Small Molten Salt Reactors with Various Powers and Core Sizes

Cited by 27 publications

References 9 publications

Investigation on Inherent Safety of One Fluid-Molten Salt Reactor (Of-Msr) With Various Starting Fuel

Investigation on Inherent Safety of One Fluid-Molten Salt Reactor (Of-Msr) With Various Starting Fuel

The effect of Am241 on UK plutonium recycle options in thorium-plutonium fuelled LWRs – Part I: PWRs

Experimental and simulation approach of the loop geometry effect on the natural circulation system of the advanced nuclear reactor

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