Liquid fuel, thermal neutron spectrum reactors

Yoshioka, Ritsuo; Kinoshita, Motoyasu

doi:10.1016/b978-0-08-101126-3.00011-7

Cited by 6 publications

(4 citation statements)

References 33 publications

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“…Both the accumulation of 233 U and the reduction of 235 U in the fuel salt are adverse to graphite‐moderator temperature coefficient, especially for a small SVF situation. Therefore, the TRC with different SVFs is also simulated up to 10 years (see Figure ) to prevent the FPs concentration in the fuel salt from exceeding its solubility limit (about 1%) during the whole operation . When the SVF is less than 20%, the TRC at 10 years is much greater than that at the beginning of life.…”

Section: Resultsmentioning

confidence: 99%

“…Therefore, the TRC with different SVFs is also simulated up to 10 years (see Figure 5) to prevent the FPs concentration in the fuel salt from exceeding its solubility limit (about 1%) during the whole operation. 45 When the SVF is less than 20%, the TRC at 10 years is much greater than that at the beginning of life. The TRC at 10 years for the SVF between 8% and 14% tends to be almost positive, which indicates that the SVF < 8% or SVF > 14% should be selected to ensure the safe operation of SMMSR.…”

Section: Once-through Fuel Cycle Modementioning

confidence: 99%

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Thorium utilization in a small modular molten salt reactor with progressive fuel cycle modes

Zou

et al. 2019

Int J Energy Res

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Summary The thorium‐uranium (Th‐U) fuel cycle is considered as a potential approach to ensure a long‐term supply of nuclear fuel. Small modular molten salt reactor (SMMSR) is regarded as one of the candidate reactors for Th utilization, since it inherits the merits of both MSR and small modular reactor. The Th utilization in a 220‐MWe SMMSR with the once‐through fuel cycle mode is investigated first. Then, the SMMSR with batch and online fuel processing modes is investigated second for comparison, considering the progressive development of fuel reprocessing technology. To keep a negative temperature reactivity feedback coefficient (TRC), a configuration for fuel salt volume fraction (SVF) equal to 15%, with a mixed fuel of low enriched uranium (LEU) and thorium at an operation time of 5 years is recommended for the once‐through mode, corresponding to the Th energy contribution (ThEC) of 37.6% and natural U and Th utilization efficiency (UE) of 0.51%. Considering the solubility limit of heavy nuclide (HN) proportion (below 18.0 mol%) in the fuel salt, the total operation time of the SMMSR shall be less than 50 years for the batch reprocessing mode with a 5‐year reprocessing interval time. In this case, the ThEC and UE can be improved to about 47.4% and 0.99%, respectively. Finally, the Th utilization and fuel sustainability are analyzed at a lifetime of 50 years for the online reprocessing fuel cycle mode, including both the only online fission products (FPs) removing scheme and the fuel transition scheme from LEU to 233U. For the former scheme, the ThEC and UE can be further improved to 58.6% and 1.52%, respectively. For the latter scheme, 233Pa is extracted continuously from the core to breed and store 233U. If a total reactor lifetime of 50 years is assumed, the operation time using LEU as starting and feeding fuel for 6 years is required, and the bred 233U during this 6‐year operation can start and maintain the reactor criticality for the remaining 44 years. In this case, the ThEC is improved significantly to 89.1% corresponding to a UE of 2.74%.

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Section: Resultsmentioning

confidence: 99%

Section: Once-through Fuel Cycle Modementioning

confidence: 99%

Thorium utilization in a small modular molten salt reactor with progressive fuel cycle modes

Zou

et al. 2019

Int J Energy Res

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“…Consequently, some fraction of β eff is "lost" from the core [26,27]. The real β eff value is lower from the value calculated by MCNP, which can range from 10% in MSR-FUJI to 40% in MSBR case [28]. Calculation of lost β eff is beyond the scope of this study so that the β eff value calculated is left as it is.…”

Section: Methodsmentioning

confidence: 90%

Neutronic Analysis of LEU-started Molten Chloride Fast Reactor without Fuel Reprocessing

Dwijayanto

Harto

2021

CERiMRE

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One of the rarely explored molten salt reactor (MSR) designs is the molten chloride fast reactor (MCFR). This MSR design employs chloride salt instead of fluoride and operated in a fast spectrum. MCFR brings all the advantages of an MSR including breeding whilst being able to burn plutonium and minor actinides efficiently. Since not many countries have access to civilian plutonium, MCFR can also be started using low-enriched uranium (LEU). This study is an initial neutronic analysis of an MCFR using LEU as its startup fuel. Parameters analyzed are conversion ratio (CR) and its neutronic safety, namely effective delayed neutron fraction (βeff), temperature coefficient of reactivity (TCR), and void coefficient of reactivity (VCR). The core is divided into Core Zone and Blanket Zone. The fuel composition of NaCl-UCl3 with a molar fraction ratio of 60:40 and 50:50 is used in Core Zone and Blanket Zone, respectively. The neutronic calculation is performed using MCNP6 code with ENDF/B-VII library. For reference geometry, CR is valued at 0.9298, βeff at 0.00731, TCR at -19.8 pcm/°C, and average VCR at -154.31 pcm/void%. Thereby, the MCFR fulfills inherent safety criteria. Although its value is remarkably high, CR can be further optimized by modifying the separator and reflector material.

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“…Verification of the DYMOS code against experimental data of MSRE, which was operated at Oak Ridge National Laboratory (ORNL) in the 1960s adopting fluoride fuel salt with graphite moderator, is already reported by the authors [23,24,25]. In a summary, DYMOS There is a benchmark report for MSRE zeropower experiments, where various models are compared with MSRE data [26].…”

Section: Verification For Msre Experimentsmentioning

confidence: 99%

Proposal of Application of a Simple Analysis Code DYMOS for Accident Analyses of Molten Salt Reactors

Shimazu,

Yoshioka,

Ogasawara

2023

JENRR

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The Molten Salt Reactor (MSR) possesses unique characteristics, including the circulation of liquid fuel salt in-core and ex-core, and the absence of cladding tubes of fuel rods. To perform safety analyses for accidents and transients, the selection of appropriate models is crucial. Proposed models vary from the coupling of a three-dimensional (3-D) nuclear reactor model and a detailed thermal/hydraulic (T/H) model to a simple point reactor model coupled with a lumped parameter model of the T/H system. The authors have been investigating the development of a simple and accurate model that can be utilized during the design stage and licensing evaluation, while also providing transparency, which means that models can be easily understandable by experts and reproduced in licensing evaluation. Given the distinctive features of the MSR, the peak heat flux or fuel cladding peak temperature is not a requirement. Instead, the most decisive parameter for safety evaluations is the fuel salt temperature in the fuel salt boundary. As demonstrated in this paper, the outlet of the MSR core consistently displays the highest temperature.Based on the afore-mentioned prospects for both nuclear and T/H systems, the authors have developed a simple safety and transient analysis code for MSR (DYMOS). The DYMOS code has been verified for MSRE experiments, as described in this paper. However, the previous verification was limited to small experimental MSRs, and there is a lack of verification for large reactor systems. This paper shows that the DYMOS code is applicable to these larger reactors. In other words, the main objective of this article is not to claim the originality of the model of DYMOS code, but to propose applicability of such simple code to large reactor systems.

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Liquid fuel, thermal neutron spectrum reactors

Cited by 6 publications

References 33 publications

Thorium utilization in a small modular molten salt reactor with progressive fuel cycle modes

Thorium utilization in a small modular molten salt reactor with progressive fuel cycle modes

Neutronic Analysis of LEU-started Molten Chloride Fast Reactor without Fuel Reprocessing

Proposal of Application of a Simple Analysis Code DYMOS for Accident Analyses of Molten Salt Reactors

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