Analysis of Th-U breeding capability for an accelerator-driven subcritical molten salt reactor

Summary Ex‐core transition to thorium fuel cycle is an effective way to address the issue of the lacking of the 233U in nature. It consists to extract online the thorium bred 233Pa from the core and store it outside. The obtained 233Pa stockpile will decay into the 233U (T1/2 = 27 days), and when enough 233U is cumulated, a new 233U‐based reactor can be loaded. However, this approach will inevitably provoke the heavy nuclides (HN) concentration in the core during the extraction operation, which would affect the chemical stability of fuel salt and subsequently endanger the safety of the core operation. In this study, an improved ex‐core transition with unchanged HN concentration is proposed for a 500‐MWth small modular heavy‐water moderated molten salt reactor (SM‐HWMSR). To achieve this, two operating stages are needed. In the first stage, transuranic (TRU) from the spent fuel of LWRs is used to feed the low enrichment uranium (LEU, 19.75 wt% 235U/U) started core with 300 operating days to ensure negative temperature reactivity coefficient (TRC). In the second stage, the thorium bred 233U from the first stage is mixed with TRU to refuel the core. The proposed ex‐core transition scheme aims to burn TRU, meanwhile breed 233U to start new MSRs. The fuel cycle performance over 60‐years' operation is analyzed. The results demonstrate that it is possible to realize ex‐core transition with unchanged concentration of HN in the fuel salt. As the fraction of 233U in the mixed fuel is set to be 15 mol%, it only takes 3 years to start a new reactor with the stored 233U. The radiotoxicity of TRU‐based fuel is decreased after 60 years' burning in the SM‐HWMSR core. Novelty Statement “Ex‐core transition to thorium cycle in an SM‐HWMSR with unchanged concentration of heavy metal nuclides in the fuel salt” was investigated by Yapeng Zhang, Jianhui Wu, Jingen Chen, and Xiangzhou Cai. The results demonstrate that it is possible to realize ex‐core transition with unchanged concentration of HN in the fuel salt. As the fraction of 233U in the mixed fuel is set to be 15 mol%, it only takes 3 years to start a new reactor by the stored 233U.

“…The CR is an important parameter and generally used for measuring the Th‐U conversion capability. The definition of conventional CR 26‐28 can be expressed as:

CR = \frac{R_{c} ((), {}^{232}{Th} + {}^{234}U + {}^{238}U + {}^{238}{Pu} + {}^{240}{Pu} - {}^{233}{Pa})}{R_{a} ((), {}^{233}U + {}^{235}U + {}^{239}{Pu} + {}^{241}{Pu})},

Section: Resultsmentioning

confidence: 99%

Ex‐core transition to thorium cycle in a small modular heavy‐water moderated molten salt reactor with unchanged concentration of heavy metal nuclides in the fuel salt

Zhang

Chen

et al. 2021

“…The other Cs isotopes for 135 Cs transmutation would also absorb a large amount of neutrons in the deletion process, which deteriorates the neutron economy of the core. Therefore, the Th‐U breeding performance for the online returning 135 Cs transmutation scenario with different mass fractions of 135 Cs is also evaluated by net 233 U production, 30,31 which is defined as

^{233} U ((), production) =^{233} U ((), residue) +^{233} Pa ((), extract) -^{233} U ((), inject),

where 233 U (inject) refers to the total injected 233 U mass into the core, which consists of the initially loaded and online‐injected 233 U masses; 233 Pa (extract) refers to the extracted 233 Pa mass from the core; and 233 U (residue) refers to the residual 233 U mass in the core. A positive value of the net 233 U production means that the reactor can product extra 233 U, while a negative value indicates that extra 233 U produced from other reactors must be injected into the core to maintain the reactor criticality.…”

Section: Resultsmentioning

confidence: 99%

Transmutation of ¹³⁵ Cs in a single‐fluid double‐zone thorium molten salt reactor

Chen

et al. 2020

Summary The single‐fluid double‐zone thorium molten salt reactor (SD‐TMSR) allows for online reprocessing fission products and adding nuclear fuel, which provides a high Th‐U breeding capability for thorium utilization. Although 135Xe in the core can be online extracted by the helium bubbling system during reactor operation to improve the reactor neutron economy, it would decay to the long‐lived 135Cs with a half‐life of 2.3 × 106 years, which is undesired from the viewpoint of the development of nuclear energy. The total production of 135Cs decayed from the extracted 135Xe is about 1.83 tons for the SD‐TMSR after 60 years operation. To reduce the significant accumulation of 135Cs, we propose a cooling transmutation method, by which the produced 135Cs is reinjected into the core for transmutation. First, a separation and multigroup cooling system in the SD‐TMSR is adopted to enhance the mass fraction of decayed 135Cs in the total Cs isotopes from 29% to 75%. Then, the separated Cs isotopes with optimized mass fraction of 135Cs are online injected into the core for transmutation. The total transmuted mass of 135Cs is about 0.61 tons after 60 years operation, corresponding to the transmuted fraction of about 37.42% and the transmutation rate of 10.17 kg/(GWe y). Moreover, the net 233U production is about 2.03 tons which is much larger than the initial 233U loading inventory of the SD‐TMSR, indicating that the Th‐U breeding mode for the transmutation of 135Cs in the SD‐TMSR can be achieved over 60 years. Novelty Statement “Transmutation of 135Cs in a single‐fluid double‐zone thorium molten salt reactor” was investigated by Kunfeng Ma, Chenggang Yu*, Jingen Chen*, and Xiangzhou Cai. A cooling transmutation method is performed to reduce the 135Cs decayed from 135Xe outside of core. It is found that a separation and multigroup cooling system improves the mass fraction of 135Cs in the total Cs isotopes from 29% to 75% and an online returning 135Cs transmutation scenario transmutes about 0.61 tons of 135Cs over 60 years.

“…Hence, the transmutation of MAs in thermal reactors is an effective means to eliminate the high radioactive hazards. Until now, numerous research works on MA transmutation based on different types of reactors have been done, demonstrating that thermal reactors, fast reactors, and accelerator‐driven systems are feasible for this purpose 6‐13 . However, the limited MA loading is a main constraint for obtaining high transmutation capability in solid‐fuel reactors because it must reconcile the enough reactivity for criticality without online refueling 14 …”

Section: Introductionmentioning

confidence: 99%

Parametric study on minor actinides transmutation in a graphite‐moderated thorium‐based molten salt reactors

Zou

et al. 2021

Summary Transmutation of minor actinides (MAs) is an effective solution to reduce the long‐term radiotoxicity of nuclear waste. Online refueling and no fuel rod fabrication in thorium‐based molten salt reactors (TMSR) are two remarkable advantages of MA transmutation. However, MA solubility limit of the fuel salt and the neutron spectrum in the core are two key parameters that have a strong impact on MA transmutation capability. In this paper, three types of carrier salt with different MA solubility limits and various salt volume fractions (SVFs) are introduced in a TMSR core to evaluate the MA transmutation characteristics. The results indicate that the Flibe salt with the best neutron economy can obtain the highest MA transmutation rate (TR), while the Flinak salt with the largest MA inventory can achieve the highest specific MA transmutation consumption (STC). STC and TR for the case with the Flibe salt and 5% SVF are about 200 kg/GWth.yr and 82%, respectively, which indicates that it can consume the annual MA yields from about 3.37 typical pressurized water reactors (PWRs). Meanwhile, the STC and TR for the case with the Flinak salt and 40% SVF are about 366 kg/GWth.yr and 75%, respectively, which indicates that it can consume the annual MA yields from about 50 typical PWRs. In addition, the total radiotoxicity of MAs after 50‐year operation in TMSR can be lowered by as high as about 50%. The significant production of Pu isotopes can be reused as nuclear fuel in fast reactors. In particular, Pu‐238, with the mass fraction of more than 60% in Pu isotopes, is a useful material for many nuclear applications. In conclusion, it is feasible to transmute MAs in TMSR and thus achieve the goals of reduction of MA's long‐term radioactive hazards.