Minor Actinides Research Articles

Theoretical Insights on the Complexation of Americium(III) and Europium(III) with Diglycolamide- and Dimethylacetamide-Functionalized Calix[4]arenes

Separation of minor actinides from lanthanides is one of the biggest challenges in spent fuel reprocessing due to the similar physicochemical properties of trivalent lanthanides (Ln(III)) and actinides (An(III)). Therefore, developing ligands with excellent extraction and separation performance is essential at present. As an excellent pre-organization platform, calixarene has received more attention on Ln(III)/An(III) separation. In this work, we systematically explored the complexation behaviors of the diglycolamide (DGA)/dimethylacetamide (DMA)-functionalized calix[4]arene extractants for Eu(III) and Am(III) using relativistic density functional theory (DFT). These calix[4]arene-derived ligands were obtained by functionalization with two or four binding units at the narrow edge of the calix[4]arene platform. All bonding nature analyses suggested that the Eu-L complexes possess stronger interaction compared to Am-L analogues, resulting in the higher extraction capacity of the these calix[4]arene ligands toward Eu(III). Thermodynamic analysis demonstrates that these pre-organized ligands on the calix[4]arene platform with four binding units yield better extraction abilities than the single ligands. Although DMA-functionalized ligands show stronger complexation stability for metal ions, in acidic solutions, the calix[4]arene ligands with DGA binding units have better extraction performance for Eu(III) and Am(III) due to the basicity of the DMA ligand. This work enabled us to gain a deeper understanding of the bonding properties between supramolecular ligands and lanthanides/actinides and afford useful insights into designing efficient supramolecular ligands for separating Ln(III)/An(III).

Efficient separation of americium by a mixed solvent of two extractants, a diamideamine and a nitrilotriacetamide.

The Japan Atomic Energy Agency (JAEA) has proposed the Solvent Extraction from Liquid waste using Extractants of CHON-type for Transmutation (SELECT) process by solvent extraction as a new separation technology to recover minor actinides (MA) from high-level liquid waste (HLLW) produced by spent fuel reprocessing. The MA separation in the SELECT process comprises the recovery of MA and rare earths (RE) from HLLW, MA/RE separation, and Am/Cm separation. Three highly practical extractants are used in the MA separation. Furthermore, this flow configuration facilitates the preparation of nitric acid concentrations in the aqueous phase. However, the separation factor between Cm and Nd in the MA/RE separation is small (SFCm/Nd = 2.5), requiring many extraction stages for continuous extraction in a mixer settler. Therefore, this study investigated the separation of only Am from an aqueous nitric acid solution containing MA (Am and Cm) and RE using an organic phase mixed with two extractants alkyl diamideamine with 2-ethylhexyl alkyl chains (ADAAM(EH)) and hexa-n-octylnitrilotriacetamide (HONTA) used in the SELECT process. Under high-concentration nitric acid conditions, Am and La, Ce, Pr, Nd (light lanthanides) were extracted in the ADAAM(EH) + HONTA mixed solvent, whereas Cm, medium, and heavy lanthanides, and Y were partitioned in the aqueous phase. Subsequently, only light lanthanides could be back extracted from the ADAAM(EH) + HONTA mixture solvent containing Am and light lanthanides in low nitric acid concentrations. Furthermore, Am could be easily stripped with 0.2M or 5M nitric acid. This method does not require the mutual separation of Cm and Nd, which have low separation factors. Am can be efficiently separated by one extraction and two back extractions, reducing the number of steps in the SELECT process.

Glass, ceramic, and glass-crystalline matrices for HLW immobilisation

We briefly overview the utilisation of glasses, ceramics and glass crystalline materials (GCMs) composed of both vitreous and crystalline phases focusing on nuclear waste immobilisation and potential use of some advanced waste forms for incorporation of the rare earth elements (REE) and minor actinide (MA: Am, Cm) fraction of high-level waste (HLW).

Enhanced Am/Eu separation ability of disulfonated diamide N-heterocyclic ligands by adjusting N-, O-donor affinity: A theoretical comparative study

Separation of minor actinides (MAs) from lanthanides (Ln) is a vital and challenging stage in advanced nuclear waste treatment. Development of water-soluble back-extraction ligand is regarded as a feasible alternative method for high efficiency An(III)/Ln(III) separation. Therefore, theoretical investigating on the deep mechanism behind the selective back-extraction of An(III) is very important. In this work, Am- and Eu-complexes with three water-soluble disulfonated N-heterocyclic ligands, disulfonated N,N’-diphenyl-2,9-diamide-1,10-phenanthroline (DS-Ph-DaPhen, L1), disulfonated N,N’-diphenyl-2,9-diamide-2,2′-Dipyridine (DS-Ph-DADipy, L2), and disulfonated N,N’-diphenyl-2,9-diamide 5H-cyclopenta[2,1-b:3,4-b’]-Dipyridine (DS-Ph-Da-5H-Cyclo-PentaDipy, L3) were systematically investigated using quasi-relativistic density functional theory (DFT) method. Electrostaic potential (ESP) and restrained electroStatic potential (RESP) atomic charge on donor atoms indicate that N atoms’ affinity toward metal ions in L2 and L3 was increased by replaceing the phenanthroline ring with dipyridine ring and 5H-Cyclo-PentaDipyridine ring, while the affinity of O atoms seem to decrease from L1 to L2 and L3. Through Wiberg bond indices and quantum theory of atoms in molecule (QTAIM) analyses, three ligands’ Am(III) preference may be attributed to the weak but different extend partial covalency in Am-N and Eu-N bonds. Based on the NBO population and PDOS analyses, slightly more overlap of Am-5f and N/O-2p orbitals than those between Eu-4f and N/O-2p orbitals. Based on the theoretical calculation, the covalency difference between Am-N and Eu-N bonds seems to have inverse correlation with the intensity of M−N bonds. However, all the theoretical results suggest that, by increasing N’s affinity and weakening O’ affinity at the same time, ligand’s Am(III)/Eu(III) separation ability can be improved. This work offers some useful information for changing ligands’ An(III)/Ln(III) back-extraction and separation abilities by adjusting the affinity of donor N and O atoms in its central N-heterocyclic skeleton, and may shed light for future designing of more efficient An(III)/Ln(III) back-extraction ligands.

Role of sulphur and oxygen in phosphine ligands in the extraction process of Am3+ over Eu3+ ions in water medium – A theoretical study

Spent nuclear fuel management, particularly the separation of minor actinides, is critical to the continued growth of the nuclear industry and to address the global energy crisis. However, due to the identical chemical properties of trivalent An3+ and Ln3+, it is difficult to distinguish them. Therefore, theoretical research on the mechanism of selective extraction of An3+ is needed. Two phosphine oxide ligands and two phosphine sulphide ligands with similar structures but different bridging frameworks namely phenanthroline and biphenyl, were theoretically explored and compared using relativistic density functional theory in the solvent phase. Here, the M-O bonds are stronger than the M-S bonds. QTAIM shows that the M-O/N bonds of the phosphine oxide ligands have a small amount of covalent character. Gibbs free energy shows that the formation of the complexes is spontaneous while energy decomposition analysis deciphers the binding of M(NO3)3 to these ligands. All the ligands studied had greater affinity for Am3+ than for Eu3+, which contributes to the selectivity of Am3+ over Eu3+ in the water medium. This study provides some important information for the An/Ln separation of phosphine oxide and phosphine sulphide ligands, which may help in the development of more effective phosphine based ligands for the separation of An3+/Ln3+ ions.

The cutting path planning of main components in China initiative accelerator driven subcritical system based on the quantum evolutionary algorithm

The CiADS is a promising nuclear demonstration facility for the purpose of transmuting long-life fission products and minor actinides to the short-life isotopes. The main components in CiADS are the large volume equipment with complex structure and high radioactivity, which are also the important source items for the decommissioning plan design that should be cut into smaller blocks for the proper management and the safety storage. However, it has been believed as the extremely difficult and challenging task to construct the cutting paths on the main components in order to obtain the optimized blocks with acceptable range of radioactivity, since the cutting paths are hard to describe with explicit expressions in the complex model structures, and the manual selection of cutting blocks with suitable radioactivity is the error prone and labor intensity task. Considering this situation, the cutting path planning method has been developed in this study. During the decommissioning process, the complex models of main components have been converted to the simplified geometries with voxelized description, and then the quantum evolutionary algorithm has been used to automatically find the appropriate cutting blocks by means of the voxelized description for the purpose of meeting the decommissioned requirements. Moreover, the reconstruction procedure has been used to record the topological information of cutting paths on the main components with the visualization function in the graphical user interface. Finally, the verification and validation of the cutting path planning method have been tested in the decommissioning scenario, and all the numerical results show that the optimized method can effectively fulfill the cutting path planning tasks of the CiADS facility.

Developing a Reactivity-Equivalent Physical Transformation to Simulate an Axially Heterogeneous Boiling Water Reactor

Hitachi is advancing their designs for a conceptual reactor called the resource-renewable boiling water reactor (RBWR), a concept reactor similar to the advanced boiling water reactor with a harder neutron spectrum. This design aims to minimise construction costs and waste production as well as to utilise separated plutonium and minor actinide fuel. However, the axial heterogeneity of the design poses calculation difficulties. The aim of this work is to use a known method, reactivity-equivalent physical transformation (RPT), for calculating fuel with double heterogeneity and apply it to a BWR-type fuel pin. This could reduce the calculation time needed for optimisation of the design of the RBWR. The objective of the study is to use SCALE 6.2 to produce an equivalent axial pin model by comparison with the burnup and neutron spectra of a radial model of the fuel. This model can then be used for 2D burnup calculations, and in future work will be used for the generation of two-group and multigroup cross-sections for further deterministic calculations as part of a two-step approach for analysis of the RBWR. The RPT method has been extensively tested on spherical fuel, and SCALE is a standard industry code. The initial radial model is a hexagonal assembly with 20% enriched UO2 fuel in a zircaloy cladding, surrounded by light water moderator. The derived axial model has a water density distribution taken from Hitachi’s RBWR designs. Criticality over 70 GWd/tU burnup is estimated using the model. The application of the RPT to the BWR pin was shown to be possible, but to have limitations with the introduction of additional radial complexity. For a single pin, excellent agreement between the radial and axial models could be found across a range of water densities, but in the case of an assembly level calculation distinct equivalence models were required for each water density. In addition, the produced RPT model is validated using SCALE’s 3D Monte Carlo module, KENO.

Neutronic investigation of prospective dual-cooled micro-heterogeneous duplex fuel for small modular long-life reactors: Assembly level design and analysis

Nuclear fuel is a critical component of any nuclear reactor, influencing both its safety and economic viability. Continuous fuel quality improvements resulted in a significant reduction in failure rate, and assembly design changes contributed to a significant increase in power density and performance during irradiation. In terms of fuel development, a new fuel type called “dual-cooled annular fuel” has recently been proposed to be used in nuclear reactors. The dual-cooled annular fuel design benefits from heat transfer to the coolant at both the outer and inner channels. Another emerging trend in nuclear fuel development is the use of thorium-based fuel in the form of duplex fuel design. As a result, the current work is an attempt to incorporate potential dual-cooled annular duplex fuels into the assembly of an innovative Small Modular Reactor (SMR). This paper, therefore, compares the neutronic properties of annular duplex fuel to those of annular UO2 fuel. Considering the dual-cooled feature, this study presents both duplex configurations (i.e. ThO2– UO2 and UO2–ThO2). The neutronic properties include fissile loading determination, multiplication factors, discharge burnup, and isotopic concentration. This analysis also determined the reactivity coefficients, power peaking factors, energy-dependent neutron flux, and an analysis of possible burnable absorbers and configurations. The results of this study show that both annular duplex fuels can achieve 6% higher discharge burnup than annular UO2 fuel, despite nearly identical reactivity values at the beginning of the cycle. The production of plutonium and other minor actinides has been greatly reduced. Annular duplex fuels have lower power peaking factors, which are beneficial for reactor operation, and the UO2–ThO2 configuration performs better. The calculated reactivity feedback coefficients show negative patterns and are within the annular UO2 fuel's range. The proposed annular duplex fuel designs have additional intriguing neutronic properties that open up possibilities for future research.

Influencing Bonding Interactions of the Neptunyl (V, VI) Cations with Electron-Donating and -Withdrawing Groups.

Neptunium makes up the largest percentage of minor actinides found in spent nuclear fuel, yet separations of this element have proven difficult due to its rich redox chemistry. Developing new reprocessing techniques should rely on understanding how to control the Np oxidation state and its interactions with different ligands. Designing new ligands for separations requires understanding how to properly tune a system toward a desired trait through functionalization. Emerging technologies for minor actinide separations focus on ligands containing carboxylate or pyridine functional groups, which are desirable due to their high degree of functionalization. Here, we use DFT calculations to study the interactions of carboxylate and polypyridine ligands with the neptunyl cation [Np(V/VI)O2]+/2+. A systematic study is performed by varying the electronic properties of the carboxylate and polypyridine ligands through the inclusion of different electron-withdrawing and electron-donating R groups. We focus on how these groups can affect geometric properties, electronic structure, and bonding characterization as a function of the metal oxidation state and ligand character and discuss how these factors can play a role in neptunium ligand design principles.

A safety enhanced sodium-cooled MOX fueled fast reactor core concept

ABSTRACT We have developed a 750MWe sodium-cooled MOX fueled fast reactor core with enhanced safety against typical ATWS. Design goals were set as follows: 1) negative void reactivity for the loss of flow (ULOF); and 2) burnup reactivity less than 1 $ for the transient over power (UTOP). To achieve the goals the following were considered: 1) adoption of the axially heterogeneous core with a sodium plenum and a gas expansion module to reduce void reactivity; and 2) addition of minor actinides (MAs) to the internal blanket to reduce burnup reactivity. Configurations of the core were optimized as follows. First, void reactivity was reduced by a parameter survey for the inner core height, B-10 content of the upper shield, and standby position of the backup control rods. Second, burnup reactivity was reduced by a parameter survey for the MA content in the internal blanket, etc. We designed the safety enhanced fast reactor core that did not undergo sodium boiling if the ULOF occurred and we maintained fuel integrity for the UTOP by the transient analysis. By installing a self-actuated shutdown system in the backup control rods, the improved operational reliability of this shutdown system and the improved safety can be expected.

Improvement in the elution performance of an N,N,N’,N-tetraoctyl diglycolamide impregnated extraction chromatography adsorbent using neodymium via micro-particle-induced X-ray emission analysis

An adsorbent used for the recovery of trivalent minor actinides (MA(III); Am and Cm) from high level liquid waste generated from reprocessing of spent nuclear fuel was subjected to micro-particle induced X-ray emission (micro-PIXE) analysis to improve its elution performance. The experimental adsorbent comprised SiO2 particles, a polymer coating, and an N,N,N′,N′-tetraoctyl diglycolamide (TODGA). The particles to be analyzed were subjected to Nd adsorption and an elution operation, but Nd in the adsorbent was found to be uniformly distributed. This might have been caused by individual differences in the amount of impregnated TODGA. The remaining Nd species were not localized to a specific part of the adsorbent after the adsorption operation. Some Nd elements were retained in the adsorbent after elution, probably because of the poor diffusion of the mobile phase inside the adsorbent. An adsorbent having a different microstructure from the first was then evaluated, and rapid elution was observed on new adsorbent along micro-PIXE analysis.

Computational simulation of minor actinide burning in a BN-600 reactor with fuel without uranium and plutonium

The paper presents the results of studies on the burning of minor actinides (MA) extracted from SNF of thermal reactors in a BN-600 reactor, which uses the complete set of MAs instead of traditional nuclear fuel types: uranium and/or plutonium. The advantages of such approach to MA burning are that long-lived waste is recycled and energy is produced that can be used, e.g., to generate electricity. Besides, where, e.g., a reactor with uranium or MOX fuel is used for transmutation, apart from burning “foreign” minor actinides, it will additionally generate “its own” MAs. The studies have shown that such reactor can be efficient only if based on fast neutrons, which is due to the specific properties of the minor actinide neutron capture and fission cross-sections as compared with traditional fuel nuclides. The calculation results have shown rather a high rate of MA transmutation and burning in a reactor fueled with minor actinides.

Effects of Evaluated Nuclear Data Libraries on the Calculation Results for Fuel Burnup with Minor Actinides in a VVER Reactor

Effects of Evaluated Nuclear Data Libraries on the Calculation Results for Fuel Burnup with Minor Actinides in a VVER Reactor

Burn-Up Analysis of Plutonium and Minor Actinide Recycling on Gas Cooled Fast Reactor (GCFR) Using SRAC COREBN

Research on the analysis of recycling, plutonium, and minor actinides in the Gas Cooled Fast Reactor (GFCR) using SRAC COREBN has been done. This research uses a fuel mixture of uranium, plutonium, and minor actinides. The analysis was conducted with computational simulation using the COREBN code, an additional code to SRAC. The purpose of this research is to find out the influence of the addition of plutonium and actinide on the composition of nuclear fuel at the end of reactor operation and the limitation of the value of the multiplication factor (keff at the end of the reactor burn up period). Found in this research is the final value, the multiplication factor (keff) of 1.19964, and the conversion ratio value of 0.766813 in the burn-up period of 1515 days as well as the maximum power density value of 125.85 watts/cm³, the relative power density value at the radius y of 1,230263 and radius x equal to 1.19737 the atomic density value experienced a change in the number of nuclides in the types of nuclides U235, U238, Pu239, Pu241, Np237, and Am243 at the end of the burn-up period.

Potential methods to reduce the thermal impact of vitrified waste from spent mixed oxide fuel on geological repositories

Compared to the vitrified, high-level radioactive waste (HLW) from spent UO2 fuel (UO2-HLW glass), that of mixed oxide (MOX) fuel (MOX-HLW glass) exhibits high heat generation over a long period of time. Thus, to reduce the associated waste volume and repository footprint, it is essential to prevent bentonite buffer alteration by mitigating heat generation. In this study, we evaluated five potential methods of decreasing the bentonite buffer temperature in a geological repository for MOX-HLW glass. Of these five methods, the separation of minor actinides (MA) from high-level liquid waste (HLLW) was the most effective. This is because the main nuclide involved in heat generation in MOX-HLW glass is Am-241. An MA separation rate of 81% was required to maintain the bentonite buffer temperature below the Japanese upper disposal limit of 100 °C. Thus, it is important to measure the amount of Am-241 when disposing of MOX-HLW glass. Assuming that only Am-241 was disposed of, the maximum Am-241 waste loading under a bentonite buffer temperature of 100 °C was 0.33 wt%, equating to the disposal of only 1.32 kg of Am-241 per 400 kg of MOX-HLW glass. Thus, there is a need for new methods of treating and disposing of Am-241.Graphical abstractRelationship between bentonite buffer temperature and (a) minor actinide (MA) separation ratio from high-level liquid waste (HLLW), (b) mixed oxide (MOX) high-level radioactive waste (HLW) glass and UO2 HLLW mixing ratio, (c) waste loading in MOX-HLW glass, (d) MOX-HLW glass storage time, and (e) canister footprint, representing five methods of mitigating heat generation from disposed MOX-HLW glass.

Understanding the Coordination Chemistry of AmIII/CmIII in the DOTA Cavity: Insights from Energetics and Electronic Structure Theory.

The minor actinides Am/Cm show multiple possibilities for coordination, providing great opportunities for their extraction and adsorption separation. Herein, we report complexation in an aqueous medium of AmIII/CmIII in the DOTA (H4DOTA = 1,4,7,10-tetraazacyclododecane-1,4,7,10-tetraacetic acid) cavity with axial ligands (OH-, F-, and H2O), based on the energetics and electronic structure properties using density functional theory (DFT). The formation and substitution reactions of OH--capped complexes are more likely to occur due to their enhanced hydration Gibbs free energies, followed by F-, and then H2O. Both the longer An-ODOTA bond lengths and the larger bite angle (∠O-An-O) in the OH--capped complexes reflect the enhanced coordination provided by the axial ligand, slightly less so for F-. Energy decomposition analysis based on the electronic structure supports the preference for OH--capped complexes with a near-perfect balance between attractive and repulsive contributions toward the interaction. Furthermore, molecular orbital analysis revealed that the frontier molecular orbitals of Am and Cm complexes are substantially different; that is, the highest occupied molecular orbital (HOMO) and lowest unoccupied molecular orbital (LUMO) compositions of the Am complexes are all contributed by 5f, while the HOMO and LUMO compositions of the Cm complexes are derived from 5f and 6d, respectively. Finally, the metal-exchange reactions demonstrate competitive complexation of DOTA toward AmIII over CmIII for the OH--capped system. These results imply the importance of coordination chemistry in actinide chemistry in general and specifically in AmIII/CmIII solution chemistry.

An assessment of scenarios for transuranic burning and 233U production in a small molten chloride fast reactor

Molten Chloride Fast Reactor (MCFR) attracts more and more attention owing to its significantly hard neutron spectrum and high solubility for actinide trichlorides, making it an appealing option for burning transuranic (TRU) elements discharged from pressurized water reactors (PWRs). Meanwhile, in recent years, the development of thorium fuel cycle has aroused considerable interest worldwide with the purpose of saving uranium reserves and further reducing the production of long-lived minor actinides. A small MCFR incorporating TRU burning and 233U production is proposed in this work in order to evaluate its incineration and production capabilities. By employing an in-house developed tool based on SCALE 6.1, the fuel cycle features of the small MCFR with three fuel loading scenarios, including pure TRU (scenario 1), TRU plus depleted uranium (scenario 2) and TRU plus thorium (scenario 3), are analyzed. The evolution behaviors of main nuclides, reactivity worth, neutron spectra and temperature reactivity coefficient (TRC) are discussed in detail. It is found that the MCFR can burn TRU effectively and the transmutation rates for the three scenarios are 930 kg/GWe/year, 375 kg/GWe/year and 768 kg/GWe/year, respectively, achieving a maximum capability for TRU burning of about 3.7 times the annual TRU production in a 1-GWe PWR with 50 MWd/kgU discharge burnup. Meanwhile, scenario 3 has the highest 233U production capability of about 110.3 kg per year, although it does not offer processing advantages over scenario 1 and scenario 2. The analysis of TRC indicates that the fuel salt TRC and the total core TRC are both strongly negative for all three scenarios owing to the thermal expansion of liquid fuel salt. Finally, a nuclear energy system consisting of both fast and thermal molten salt reactors is proposed and evaluated to establish a closed nuclear fuel cycle with very low radioactive wastes.

Transmutation of MAs and LLFPs with a lead-cooled fast reactor

The management of nuclear wastes has long been a problem that hinders the sustainable and clean utilization of nuclear energy since the advent of nuclear power. These nuclear wastes include minor actinides (MAs: 237Np, 241Am, 243Am, 244Cm and 245Cm) and long-lived fission products (LLFPs: 79Se, 93Zr, 99Tc, 107Pd, 129I and 135Cs), and yet are hard to be handled. In this work, we propose a scheme that can transmute almost all the MAs and LLFPs with a lead-cooled fast reactor (LFR). In this scheme, the MAs and the LLFPs are loaded to the fuel assembly and the blanket assembly for transmutation, respectively. In order to study the effect of MAs loading on the operation of the core, the neutron flux distribution, spectra, and the keff are further compared with and without MAs loading. Then the LLFPs composition is optimized and the support ratio is obtained to be 1.22 for 237Np, 1.63 for 241Am, 1.27 for 243Am, 1.32 for 79Se, 1.53 for 99Tc, 1.02 for 107Pd, and 1.12 for 129I, respectively, indicating that a self-sustained transmutation can be achieved. Accordingly, the transmutation rate of these nuclides was 13.07%/y for 237Np, 15.18%/y for 241Am, 13.34%/y for 243Am, 0.58%/y for 79Se, 0.92%/y for 99Tc, 1.17%/y for 107Pd, 0.56%/y for 129I. Our results show that a lead-cooled fast reactor can be potentially used to manage nuclear wastes with high levels of long-lived radioactivity.

Strategies of eliminating nuclear waste using accelerator-driven system in the transition stage for sustainable and clean nuclear energy in China

Based on the current situations and development tendencies of nuclear energy in China, three scales of nuclear-energy development are predicted. This work explores the strategies of transmuting the Minor Actinides (MAs) in the spent nuclear fuel using Accelerator-Driven Subcritical systems (ADS) to achieve the transition to sustainable and clean nuclear fission energy. The mass-flow data of fuel cycle with time for ADS transmutation system is obtained through simulations, which provides the basic reference for the fuel fabrication, the reprocessing scale and the interim storage of discharged fuel in Chinese ADS systems arranged in different stages. The time required for eliminating over 95% MAs inventory from Pressurized Water Reactor (PWR) spent fuel is discussed. The mass and toxicity of final high-level waste in different stages are analyzed to demonstrate the excellent ability of ADS to eliminate the mass of MAs and the inventory of high-level waste.

Scoping Studies for a Lead-Lithium-Cooled, Minor-Actinide-Burning, Fission-Fusion Hybrid Reactor Design

A feasibility study of a subcritical fission-fusion hybrid reactor using lead-lithium eutectic as a coolant and minor actinides (MAs) as fuel is presented. Such a reactor could support the fission community by transmuting MAs and the fusion community by breeding tritium. The feasibility of such a reactor for the burnup of MAs is assessed in terms of burnup performance, tritium breeding, and safety characteristics. Tandem mirrors are a promising neutron source technology, and a deuterium-tritium tandem mirror is considered here for the neutron source with power Psource = 1.13 MW assumed for scoping purposes. Subcritical reactivities from keff = 0.9800 to keff = 0.9950 were considered, representing the initial reference for subcritical reactivity and the assessed upper limit, respectively. Stability analyses indicated the reactivity would be stable under perturbations of fuel, coolant, and inlet temperatures, with a positive reactivity insertion expected during reactor shutdown. This range corresponded to nuclear heating values of 150 to 650 MW and mass burn rates of 53 to 216 kg/year. The upper mass burn rate limit would require 1110 reactor years with a capacity factor of 0.9 to fission the global supply of MAs and could offset the annual U.S. MA production with eight reactors. Tritium breeding was assessed for keff = 0.9800 and 3.795% 6Li enrichment in the coolant, and a tritium breeding ratio of 1.602 0.017 was tallied, suggesting the reactor could, without elevated 6Li enrichment, produce tritium to both sustain operation and supply tritium for other fusion devices. Time-series modeling of fuel burnup was conducted for a four-batch loading scheme and three different fuel residence times at keff = 0.9800, which showed system performance would drop with burnup, and that the rate of this drop was lower for longer fuel residence times, motivating a means of reactivity control. Last, changes in fuel composition with burnup were assessed for relative concentrations of MAs, transmutation products, and fission products. The breeding of plutonium in the blanket was calculated and found to be of minimal proliferation concern.