Thorium-based Molten Salt Reactor Research Articles

Core Optimization for Extending the Graphite Irradiation Lifespan in a Small Modular Thorium-Based Molten Salt Reactor

The lifespan of core graphite under neutron irradiation in a commercial molten salt reactor (MSR) has an important influence on its economy. Flattening the fast neutron flux (≥0.05 MeV) distribution in the core is the main method to extend the graphite irradiation lifespan. In this paper, the effects of the key parameters of MSRs on fast neutron flux distribution, including volume fraction (VF) of fuel salt, pitch of hexagonal fuel assembly, core zoning, and layout of control rod assemblies, were studied. The fast neutron flux distribution in a regular hexagon fuel assembly was first analyzed by varying VF and pitch. It was demonstrated that changing VF is more effective in reducing the fast neutron flux in both global and local graphite blocks. Flattening the fast neutron flux distribution of a commercial MSR core was then carried out by zoning the core into two regions under different VFs. Considering both the fast neutron flux distribution and burnup depth, an optimized core was obtained. The fast neutron flux distribution of the optimized core was further flattened by the rational arrangement of control rod channels. The calculation results show that the final optimized core could reduce the maximum fast neutron flux of the graphite blocks by about 30% and result in a more negative temperature reactivity coefficient, while slightly decreasing the burnup and maintaining a fully acceptable core temperature distribution.

Highly selective removal of thorium from acidic solutions achieved on phosphonic acid functionalized core–shell magnetic nanocomposite: Understanding adsorption and binding mechanisms

Thorium has received widespread attention as potential nuclear fuel alternative for the next-generation thorium-based molten salt reactors. With the growth of nuclear energy, it is needful to exploit magnetic nanomaterials to remove thorium from aqueous solutions, owing to their superparamagnetism and large specific surface energy. Nevertheless, most of magnetic nanomaterials often suffered from low thorium selectivity and poor structural ability in acidic solutions. Herein, a novel core–shell magnetic nanocomposite Fe3O4@SiO2/P (AA-MMA-VPA) possessing phosphonic acid was developed by a controlled radical polymerization of vinylphosphonic acid (VPA) in the presence of Fe3O4 encapsulated by SiO2 (Fe3O4@SiO2). Due to the presence of the high content of phosphonic acid groups and the absence of any universal groups, Fe3O4@SiO2/P (AA-MMA-VPA) displayed exceptional selectivity for thorium in the presence of 11 co-existing cations. The separation factor STh/M value for Fe3O4@SiO2/P(AA-MMA-VPA) concerning all other competing cations surpassed 74.4, and the largest value for STh/M even amounted to 343.5. Further, Fe3O4@SiO2/P (AA-MMA-VPA) possessed a maximum adsorption capacity (qmax) of 485.4 mg g−1 at pH of 3.0, higher than that of other magnetic sorbents. By reason of its core–shell structure, the sorbent possessed an outstanding structure stability even after immersed into HNO3 solution at pH of 0.5 for 72 h. The sorbent could as well be magnetically recouped, and recycled at least six times without evident diminution in sorption amount. The sterling removal performance of the sorbent was assigned to the interaction of thorium with phosphonic acid, corroborated by FT-IR, XPS and DFT method. Thus, this study affords a new viewpoint for exploiting magnetic nanomaterials with sterling acidic resistance for the strongly selective entrapment of thorium from acidic media.

A simulation study of tritium distribution in a 10WM(e) thorium-based molten salt reactor

The TMSR, with its thorium fuel cycle and molten salt technology, is an increasingly popular alternative to traditional nuclear reactors. In this study, an enhanced computational model was used to assess the distribution of tritium in SINAP's proposed 10 MW(e) TMSR system. The findings reveal that, in the absence of tritium removal measures, approximately one-third of the produced tritium could contaminate the environment. Moreover, the saturation of graphite adsorption can result in even greater tritium permeation over time. However, by implementing a method akin to the MSRE spray gas removal, approximately two-thirds of the tritium can be eliminated, reducing permeation to around 10 % and facilitating control. Furthermore, reducing the tritium yield from the reactor can further diminish the overall tritium amount and the proportion of permeation. For instance, substituting FNaBe salt for FLiBe salt as the fuel carrier yields lower tritium, thereby rendering control unnecessary or significantly reduced.

Selective separation of thorium and uranyl in phases of different polarity using novel benzoxazole-based ligands: A DFT study

The separation of thorium and uranium is an important guarantee for the fuel cycle of thorium-based molten salt reactors. Benzoxazole-based ligands have a wide range of applications and are also potential ligands for actinide separation. In this work, density functional theory (DFT) calculations were used to study the intrinsic properties of two ligands, 2,6-bis(benzo[d]oxazole-2-yl)pyridine (ligand a) and 1-(6-(4-acetylbenzo[d]oxazol-2-yl)pyridin-2-yl)ethan-1-one (ligand b), revealing the structural characteristics and complexation behavior of their complexes with Th4+/UO22+. Quantum theory of atoms in molecules (QTAIM) and extended transition state - natural orbitals for chemical valence (ETS-NOCV) analyses show that the ligands form classical coordination bonds with Th4+/UO22+, with M−Noz bonds stronger than M−Npy bonds and partial electron acceptance by carbonyl O. The thermodynamic properties were used to evaluate the separation performance of these ligands for Th4+/UO22+ in four solvents (water, 1-hexanol, n-octanol and n-dodecane), with La more suitable for back extraction process and Lb having stronger binding with Th4+/UO22+. This work provides valuable theoretical insights into Th4+/UO22+ separation based on benzoxazole.

An Al2O3/Ni–Al tritium permeation barrier coating for the potential application in thorium-based molten salt reactor

The Al2O3/Ni–Al composite coating was formed on the outside surface of GH3535 superalloy by pack aluminizing and subsequent in-situ oxidation, and deuterium permeation tests were carried out to assess the tritium permeation barrier performance of the coating. Scanning Electron Microscopy (SEM) showed that the surface of Al2O3 films was dense, homogeneous, and crack-free. Transmission Electron Microscopy (TEM) photos manifested that the interface between γ-Al2O3 film and Ni–Al interlayer was distinct without micropores, and the thickness of Al2O3 film with the structure of γ-phase is approximately 80 nm. The Arrhenius equation for permeability of GH3535 was obtained as Φ=1.37·10−7·exp(−57.92(KJ mol−1)/RT) by deuterium permeation tests. The deuterium permeation reduction factor of Al2O3/Ni–Al coating kept 2–3 orders of magnitudes at a temperature from 400°C to 700°C. The D-permeation resistant performance of this coating still has the potential to be further improved.

Preparation of α-Al2O3/NiAl multilayer coatings on GH3535 superalloy surface by pack cementation and subsequent in-situ oxidation

Tritium would be produced in the Thorium-based Molten Salt Reactor and then quickly penetrate through the structural material of GH3535 superalloy into environment at elevated temperatures. To suppress the permeation of tritium, the α-Al2O3/NiAl multilayer coatings were prepared on GH3535 by pack aluminizing (PCA) and subsequent low vacuum oxidation as tritium permeation barriers. Herein, the effects of PCA process parameters on the composition and structure of the Ni–Al coatings were characterized, simultaneously, the influence of oxidation temperature on the composition and structure of Al2O3 films were analyzed in detail. Experimental results demonstrated that the main phase of the Ni–Al layers is Ni2Al3 (or NiAl) doped with Al86Cr14. Significant amounts of Mo-aluminides were found on the surface of the samples aluminized at 1050 °C. The Ni–Al layers growth kinetic equation under 650–850 °C was empirically obtained as h = 7.07•104W1/2t1/2exp[-7.53•104/RT]. After low vacuum heat-treatment, the α-alumina films with a thickness of approximately 1 μm were formed at 1050 °C for 2 h, densely and non-porously. Meanwhile, the brittle Ni2Al3 in the transition layer was transformed into the tough NiAl, providing a self-healing capability as well as alleviating the thermal mismatch between the coating and the substrate.

Experimental research on convective heat transfer characteristics of molten salt in a pebble bed channel with internal heat source

Convective heat transfer between the molten salt coolant and the fuel pebbles in the reactor core is one of the key thermal hydraulic phenomena for Thorium Molten Salt Reactors. In this study, a pebble bed experiment facility using induction heating to provide internal heat source is designed. The heat transfer characteristics of molten salt in the pebble bed channels are studied on the Heat Transfer Salt (HTS) loop. The effects of inlet temperature, molten salt flow rate and induction heating power are investigated on convective heat transfer characteristics. A new correlation is proposed for predicting the convective heat transfer coefficients of molten salt with Prandtl number at 11.27–14.51 and Reynolds number at 2800–6600. The results will provide an important reference for the core design of thorium based molten salt reactor.

Selective Capture Mechanism of Radioactive Thorium from Highly Acidic Solution by a Layered Metal Sulfide.

Thorium as a potential nuclear fuel for the next-generation thorium-based molten salt reactors holds significant environmental and economic promise over the current uranium-based nuclear reactors. However, because thorium (Th4+) usually coexists with other rare earth elements, alkali or alkaline earth metals in minerals, or highly acidic radioactive waste, seeking acid-resistant sorbents with excellent selectivity, high capacity, and fast removal rate for Th4+ is still a challenging task. In this work, we investigated a robust layered metal sulfide (KInSn2S6, KMS-5) for Th4+ removal from strong acidic solutions. We report that KMS-5 could capture Th4+ from a 0.1 M HNO3 solution with extremely high efficiency (∼99.9%), fast sorption kinetics (equilibrium time < 10 min), and large distribution coefficient (up to 1.5 × 106 mL/g). Furthermore, KMS-5 exhibited excellent sorption selectivity towards Th4+ in the presence of large amounts of competitive metal ions like Eu3+, Na+, and Ca2+. This extraordinary capture property for Th4+ is attributed to the facile ion exchange of Th4+ with K+ in the interlayers and subsequent formation of a stable coordination complex via Th-S bonds. These results indicate that KMS-5 is a promising functional sorbent for the effective capture of Th4+ from highly acidic solutions.

Estimation of fission yield and cross-section for thorium-based molten salt reactor and accelerator driven system

Thorium is the primary candidate fuel in Thorium-based Molten Salt Reactor (TMSR) and Accelerator Driven System (ADS). However, to utilize Thorium in the reactor and accelerator systems, nuclear data in the literature are not sufficient for neutron-induced fission reaction process. Therefore, we estimated neutron fission cross-section [Formula: see text] (mb), independent [Formula: see text] and primary fission fragment yield [Formula: see text], mass-dependent average neutron multiplicity [Formula: see text], energy-dependent average neutron multiplicity [Formula: see text] and neutron emission multiplicity distribution [Formula: see text] of [Formula: see text] for different neutron incident energy values, 1 MeV, 1.5 MeV, 3 MeV, 5.5 MeV, 7.7 MeV, 10 MeV, 14.8 MeV and 20 MeV, where the energy region is important for the reactor and accelerator systems used in the industry. The obtained results are compared with experimental data obtained from the literature and are discussed for each energy value.

Safe, clean, proliferation resistant and cost-effective Thorium-based Molten Salt Reactors for sustainable development

ABSTRACT Sustainable development requires sustainable energy sources. Nuclear energy is proposed, but it is perceived as problematic in terms of proliferation, waste, safety and costs. In this paper, these issues are analyzed, and it is demonstrated that this perception is not rooted in the reality of modern nuclear technologies. In fact, the paper concludes that Thorium-based Molten Salt Reactors (TMSR) technology is clean, safe, proliferation resistant and cost effective, and even better than traditional nuclear technologies. Given that thorium is a plentiful resource, this technology can propel humanity forward for the next 1000 years or more. What is lacking is an understanding of TMSR by those who allocate funding for research. Hence, this paper is also a call for action.

НОВАЯ СИСТЕМА УПРАВЛЕНИЯ И ЗАЩИТЫ РЕАКТОРА SD-TMSR

The current work introduces a reliable safety system based on control rods in addition to the online feed system reactivity control in the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR). The reactivity of the SD-TMSR core is controlled through two systems of control assemblies: (1) the Control Safety Devices (CSD) and (2) the Diverse Safety Devices (DSD). In the present work, the control rods are natural B4C and B4C-90 (with 90% weight content of the main absorbing 10B isotope). Since the numbers and distribution of control assemblies in SD-TMSR have not been studied previously, we proposed a unique distribution as a starting point for this analysis. The distribution of these 25 fuel assemblies with control rods in the SD-TMSR core and their numbering scheme were presented in this paper. Additionally, excess reactivity, control rod worth, and shutdown margin were calculated using Monte Carlo code Serpent2. Analysis results showed that the proposed placement of the control rods will make it possible to compensate for excess reactivity during fuel burnout and emergency shutdown of the reactor.

A methodology for evaluating the transmutation efficiency of long-lived minor actinides

Up to now no definite internationally recognized quantitative criterion of minor actinides (MAs) transmutation efficiency was worked out, although this would be highly desirable. The absolute and relative total mass reduction of MAs are completely inadequate because they ignore the accumulation of higher radiotoxic long-lived MAs from the transmuted nuclide. In the current work, we introduce a new criterion for transmutation efficiency of MAs in nuclear reactors and demonstrate its efficiency by comparing two molten salt reactors; the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR) and the Small Molten Salt Fast Reactor (SMSFR). Our proposed criterion takes into account the mass of all useful actinides, short-lived MAs, and short-lived fission products (FPs). In contrast, the mass parameters calculate the reduction in the MAs mass regardless of the produced nuclides. We introduce a new approach to load MAs into both reactors. The proposed approach merges the advantages of both homogeneous and heterogeneous approaches. The overall change in the actinides and FPs mass during the irradiation has been calculated using direct SERPENT-2 calculations. The results show that the transmutation efficiency of 241Am (the prime isotope for the transmutation) in the SD-TMSR is much higher than in the SMSFR. After 1500 days of radiation, the transmutation efficiency reaches 82.6% for SD-TMSR, however, for SMSFR it reaches 52.5%.

Transmutation of americium and neptunium in the single-fluid double-zone thorium-based molten salt reactor

In the current work, we introduce a new approach compared to solid-fuel reactors to load the minor actinides (MAs) into the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR). The proposed approach merges the advantages of both homogeneous and heterogeneous approaches. Among MAs nuclides, 241Am and 237Np are selected for transmutation due to their long half-life. We simulate two separate tanks; Pu + U tank and FPs tank. In this study, a tank is a right cylinder with a volume of 1.0 m3. The Pu + U tank is used to store Pu and U isotopes extracted from the central channel of the SD-TMSR. However, the FPs tank is used to store all fission products (FPs) produced from the transmutation process in the central channel. The overall change in the actinides and FPs mass during the irradiation has been calculated using direct SERPENT-2 calculations. The results show that the transmutation ratio of 241Am and 237Np reaches 98.5% and 93.2%, respectively after 1500 days of irradiation. We notice that the major isotope in the Pu + U tank is 238Pu. Under 241Am irradiation, our proposed approach offers ≈0.3kg of 238Pu after 1 year of operation. However, under 237Np irradiation, 2.5 times more 238Pu can be extracted after the same period of operation. The produced 238Pu can be used in the radioisotope thermoelectric generators (RTG, RITEG) and radioisotope heater units.

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

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.

Design of Test Platform for Thorium-based Molten Salt Reactor Protection System

This paper summarizes the characteristics and test requirements of the thorium-based molten salt reactor protection system, and proposes a test platform design for the fourth-generation thorium-based molten salt reactor (TMSR) protection system. It mainly includes system design, hardware selection, software design and system interface. Based on the Producer-Consumer software architecture built by virtual instrument technology, the test platform uses the NI PXI hardware platform to realize channel testing, functional testing, online debugging and other functions. This design can fully test and verify the protection system without the process system, and can be extended to test other reactor-type protection systems or monitoring systems.

Distribution and behaviour of 233Pa in 2LiF-BeF2 molten salt.

Distribution and behavior of 233Pa, essential in the thorium–uranium nuclear fuel cycle, were studied in 2LiF–BeF2 (66 : 34 mole%, FLiBe) molten salt by γ-ray spectrometry. The experiments showed that 233Pa deposited slightly on the surface of graphite crucible. The addition of Hastelloy and metallic lithium decreased the 233Pa specific activity in the salt by 1 to 2 orders of magnitude rapidly. Analysis indicated that reductive deposition of 233Pa was responsible for the rapid decrease of 233Pa specific activity in the salt. Additional experiments strongly supported the mechanism of reductive deposition of 233Pa induced by Hastelloy and metallic lithium. In view of the large deposition of 233Pa on Hastelloy, the possible influence of fissile nuclide 233U produced from 233Pa decay on the operation of thorium-based molten salt reactor was discussed.

Preliminary design of control rods in the single-fluid double-zone thorium molten salt reactor (SD-TMSR)

Recent studies on Molten Salt Reactors (MSRs) showed that the excess reactivity at the beginning of the operation is large for many fueling strategies and must be compensated by a reactivity control system. The current work introduces a reliable safety system based on control rods in addition to the online feed system reactivity control in the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR). Three different initial fissile loadings are considered: 233U, reactor-grade Pu, and transuranic (TRU) elements as a startup fuel. We applied six different absorbing materials to investigate the main operational and safety parameters in the SD-TMSR: natural B4C, enriched B4C with 90% 10B, HfB2, HfH1.62, Eu2O3, and Gd2O3. The present work focuses on control rod design, integral and differential control rod worth, shutdown margin, and shadowing effects at steady-state. We employed the SERPENT-2 Monte-Carlo code to calculate the reactivity worth and analyze the performance of the reactivity control system. We showed that 233U and reactor-grade Pu startup cores maintain adequate shutdown margin with all considered absorbers. Finally, this paper proposes a design of control rod clusters that compensate the excess reactivity of the SD-TMSR loaded with different initial fissile material.

Thermal-and fast-spectrum molten salt reactors for minor actinides transmutation

The long-lived Minor Actinides (MAs): 237Np, 241Am, 243Am, 243Cm, 244Cm, and 245Cm are responsible for the effective dose and heat generation after direct disposal in deep geological structures. Thus, long-lived MAs represent a major burden of nuclear power. The long-lived MAs have not yet been utilized as nuclear fuel. Therefore, the transmutation of these MAs is proposed as an alternative to the direct final disposal. In the current work, we analyze and compare the MAs transmutation performance in the critical Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR) and Small Molten Salt Fast Reactor (SMSFR). We study the variation of the Keff and core reactivity with different MAs loadings, the neutron spectrum shift, the time evolution of MAs and major nuclides inventories, and the transmutation ratio (TR). The TR of the long-lived MAs is calculated by using the SERPENT-2 Monte-Carlo code. The total neutron flux in the SD-TMSR and SMSFR can reach 4.1 × 1014 and 1.8 × 1015 n.cm−2.s−1, respectively. The results show that the SD-TMSR consumes about 50% of the generated Pu isotopes in the fuel salt, however, the SMSFR consumes about 86.5% of the generated Pu isotopes. During burnup, we apply the online reprocessing and refueling, therefore, the core is maintained critical and the total fuel mass in the core and blanket is almost constant. The results demonstrate that both reactors effectively transmute 237Np, 241Am,243Am, and 243Cm, meanwhile, the SMSFR has a higher TR than the SD-TMSR. The TR of the total MAs reaches 54.84% and 87.97% in the SD-TMSR and SMSFR, respectively.

Strategies for thorium fuel cycle transition in the SD-TMSR

Liquid-fueled Molten Salt Reactor (MSR) systems represent advances in safety, economics, and sustainability. The MSR has been designed to operate with a Th/233U fuel cycle with 233U used as startup fissile material. Since 233U does not exist in nature, we must examine other available fissile materials to start up these reactor concepts. This work investigates the fuel cycle and neutronics performance of the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR) with different fissile material loadings at startup: High Assay Low Enriched Uranium (HALEU) (19.79%), Pu mixed with HALEU (19.79%), reactor-grade Pu (a mixture of Pu isotopes chemically extracted from Pressurized Water Reactor (PWR) spent nuclear fuel (SNF) with 33 GWd/tHM burnup), transuranic elements (TRU) from Light Water Reactor (LWR) SNF, and 233U. The MSR burnup routine provided by SERPENT-2 is used to simulate the online reprocessing and refueling in the SD-TMSR. The effective multiplication factor, fuel salt composition evolution, and net production of 233U are studied in the present work. Additionally, the neutron spectrum shift during the reactor operation is calculated. The results show that the continuous flow of reactor-grade Pu helps transition to the thorium fuel cycle within a relatively short time (≈4.5 years) compared to 26 years for 233U startup fuel. Finally, using TRU as the initial fuel materials offers the possibility of operating the SD-TMSR for an extended period of time (≈40 years) without any external feed of 233U.

Numerical simulation of stress behavior of dowel–brick structures in TMSR

As essential elements of the graphite reflector in thorium-based molten salt reactor, dowel–brick structures are used to withstand complex working loads in the reactor core and their failure may lead to serious damage of the graphite reactor core. It is crucial to investigate the stress behavior of dowel–brick structures for safe operation of the graphite reactor. In this study, three groups of finite element analyses and a strain test were carried out to investigate how the geometric parameters of the dowels affect the stress behavior of the dowel–brick structure. The numerical results indicate that the stress behavior of a dowel–brick structure is significantly affected by the diameter, length, and aspect ratio of the dowels. The maximum stress in the lower and upper bricks decreases with an increase in the dowel length. The location of maximum stress on both lower and upper bricks shifts from the root of the socket to the edge of that socket beside the contact region, as the length of the dowel increases. The shift of the maximum stress location occurs earlier for the upper bricks than for the lower bricks. The results of strain tests show good agreement with those of numerical analyses.