Incineration Of Minor Actinides Research Articles

Enhanced transmutation of minor actinides by incorporation of multi-beam design in Accelerator-Driven subcritical system

This study aims to identify the optimal beam design for incinerating the long-lived minor actinides in Accelerator-Driven System (ADS). All the analyses were performed using the MCNP6.1 code with cross-section library ENDF/B-VII. In this study, a depletion method for ADS was conducted and analyses were exercised to investigate the burnup performances for the various arrangements of a three-proton beam design. The influences of the various arrangements of the three-proton beam design were used as a criterion for comparison. It was found that the dual proton beam arrangement demonstrated the best depletion result for the incineration of minor actinides.

Transmutation of Plutonium and Minor Actinide in PWR Thorium-Transuranic Fuel Assembly

Long-lived radioactive waste is often considered a concerning issue on utilizing nuclear power. This waste is in the form of plutonium and minor actinides (MA), formed due to successive neutron capture of 238U. Both are not particularly hazardous radiologically, but their long half-life caused an issue in the public acceptance of radioactive waste disposal. Thereby, this issue must be resolved, either politically or technically. One of the technical solutions to address the issue of long-lived radioactive waste is the incineration of transuranic (TRU) elements in a pressurized water reactor (PWR) fuel assembly. Mixing TRU with thorium in a PWR fuel assembly can theoretically reduce TRU stockpile more effectively than uranium. This paper discusses plutonium and MA transmutation in a Thorium-Transuranic (Th-TRU) PWR fuel assembly using MCNP6 code and ENDF/B-VII library. A Westinghouse 17×17 PWR fuel assembly was chosen to determine the feasibility of TRU incineration. Three assemblies of 50:50, 55:45, and 60:40 mixture of UO2 and Th-TRU fuel rods are compared to one reference UO2 assembly from the criticality and burn-up point of view. From the calculations, the plutonium incineration rate was observed to be 9.47 %, 10.91 %, and 12.2 %, while MA incineration rates were found to be 11.5 %, 11.93 %, and 12.3 % for Th50-TRU50, Th55-TRU45, and Th60-TRU40, respectively. This observation indicates that a higher thorium fraction in the fuel assembly can increase the TRU transmutation rate. Therefore, a high thorium fraction is recommended to increase the transmutation rate of the TRU.

Transmutation of Plutonium and Minor Actinide in PWR Thorium-Transuranic Fuel Assembly

Long-lived radioactive waste is often considered a concerning issue on utilizing nuclear power. This waste is in the form of plutonium and minor actinides (MA), formed due to successive neutron capture of 238U. Both are not particularly hazardous radiologically, but their long half-life caused an issue in the public acceptance of radioactive waste disposal. Thereby, this issue must be resolved, either politically or technically. One of the technical solutions to address the issue of long-lived radioactive waste is the incineration of transuranic (TRU) elements in a pressurized water reactor (PWR) fuel assembly. Mixing TRU with thorium in a PWR fuel assembly can theoretically reduce TRU stockpile more effectively than uranium. This paper discusses plutonium and MA transmutation in a Thorium-Transuranic (Th-TRU) PWR fuel assembly using MCNP6 code and ENDF/B-VII library. A Westinghouse 17×17 PWR fuel assembly was chosen to determine the feasibility of TRU incineration. Three assemblies of 50:50, 55:45, and 60:40 mixture of UO2 and Th-TRU fuel rods are compared to one reference UO2 assembly from the criticality and burn-up point of view. From the calculations, the plutonium incineration rate was observed to be 9.47 %, 10.91 %, and 12.2 %, while MA incineration rates were found to be 11.5 %, 11.93 %, and 12.3 % for Th50-TRU50, Th55-TRU45, and Th60-TRU40, respectively. This observation indicates that a higher thorium fraction in the fuel assembly can increase the TRU transmutation rate. Therefore, a high thorium fraction is recommended to increase the transmutation rate of the TRU.

Minor Actinides Transmutation and 233U Breeding in a Closed Th-U Cycle Based on Molten Chloride Salt Fast Reactor

Long-lived minor actinides (MAs) are one of the primary contributors to the long-term radiological hazards of nuclear waste, and the buildup of MAs is hampering the development of nuclear power. The transmutation of MAs in reactors is regarded as a potential way to replace direct disposal to reduce the impact of MA on the environment and improve the utilization of fuel. Due to its superior features, such as outstanding neutron economy, no fuel assembly fabrication, high neutron flux, and especially online refueling and reprocessing, the molten chloride salt fast reactor (MCFR) is regarded as one of the potential reactors for MA incineration. In this work, MA transmutation capability and 233U breeding performance for an optimized MCFR have been evaluated in different scenarios. The results show that the MA transmutation capability and 233U breeding performance with online transuranic elements (TRU) and 232Th feeding scenario are improved significantly compared with the case in online 233U and 232Th feeding, when the initial MA loading is 5 mol%, the total mass of MA transmutation and MA incineration is 7160 kg and 1759 kg during the whole 100 years operation under online TRU and 232Th feeding scenario, and the corresponding average annual net production of 233U is 450 kg, however, the MA transmutation amount, MA incineration amount and average annual net production of 233U for online 233U and 232Th feeding scenario is 5298 kg, 1315 kg, and 249 kg, respectively. In addition, the research also shows that the increase in initial loading of MA has no obvious effect on the improvement of the 233U breeding performance but can improve the transmutation efficiency of MA under online TRU and 232Th feeding scenarios. Furthermore, if 233U is continuously extracted online from the core during the operation, the 233U breeding performance will be significantly improved, but it will deteriorate the safety performance, such as the fuel temperature coefficient of reactivity (TCR) and the effective delayed neutron fraction (EDNF), more importantly, it will also put forward higher requirements for the immature online reprocessing technology.

Calculations of neutron fluxes and isotope conversion rates in a thorium-fuelled MYRRHA reactor, using GEANT4 and MCNPX

Neutronics calculations have been performed of the MYRRHA ADS Reactor with a thorium-based fuel mixture, using the simulation programs MCNPX (Waters, 2002) and Geant4 (Agostinelli, 2003). Thorium is often considered for ADS systems, and this is the first evaluation of the possibilities for thorium based fuels using a reactor design which has been developed in detail. It also extends the application of the widely-used Geant4 program to the geometry of MYRRHA and to thorium. An asymptotic 232Th/233Umixture is considered, together with the standard MOX fuel and a possible 232Th/MOX starter. Neutron fluxes and spectra are calculated at several regions in the core: fuel cells, IPS cells and the two (Mo and Ac) isotope production cells. These are then used for simple calculations of the fuel evolution and of the potential for the incineration of minor actinide waste. Results from the two programs agree and support each other and show that the thorium fuel is viable, and has good evolution/breeding properties, and that minor actinide incineration, though it will not take place on a significant scale, will be demonstrable.

Preliminary Study on Minor Actinide Incineration in RSG-GAS without Isotope Separation

Minor actinides (MA) resulted from nuclear power plants is often considered as nuisance in spent fuel management due to its considerably long half-life. One of available strategies to deal with MA is to incinerate it, in order to reduce its radioactivity. This paper presents a study on MA incineration in RSG-GAS research reactor. Unlike previous study, this work did not separate the MA into individual isotopes, but incinerated as a whole. ORIGEN2.1 code is employed to calculate MA incineration within RSG-GAS core. MA composition used in this study consists of Np, Am, and Cm isotopes. The Central Irradiation Position (CIP) of RSG-GAS is loaded by 6 kg of MA and irradiated for two years. The result shows that about 1 kg of MA were incinerated after two years of irradiation, or 18,87% of the initial concentration. However, the increase of Cm-242 isotope, along with newly-formed Pu isotopes, were found to be significantly increasing short-term radioactivity compared to un-irradiated MA. Thus, two years-worth of MA incineration cannot be considered as effective, and other strategies must be pursued.

Study on the fusion reactor performance with different materials and nuclear waste actinides

In this study, both pure fusion blanket and fusion-fission (hybrid) reactor blanket performance were investigated and discussed separately in two phases. In the first phase, a Monte Carlo radiation damage analysis has been performed for stainless steel (SS304, SS316, and oxide dispersion strengthened (ODS)), molybdenum, vanadium, and tungsten as the first wall (FW) materials, in combination with selected tritium breeders. The main technical parameters for fusion reactors, such as tritium breeding ratio, fusion energy multiplication factor (M), displacement per atom (DPA), and gas production (He, H) have been evaluated. All numerical calculations have been carried out in spherical geometry with MCNP6 code package using continuous energy cross-sections from the ENDF/B-VIII.0 library, except DPA calculations. Instead of the ENDF/B-VIII.0 library, the 30-group CLAW-IV library was employed for DPA calculations. Structural material selection for the FW respect to radiation damage limits and reactor performance for energy production and tritium has been concluded. Conventional thermal reactors, such as light water reactors andCanada Deuterium Uranium (CANDU) reactors are producing substantial quantities of transuranic elements, which represent serious nuisance and permanent hazard potential. On the other hand, they become fissionable material under high energetic fusion neutron irradiation and multiply the fusion energy. In the second phase, the investigations are extended to the incineration of minor actinides (MA) in the fusion-fission (hybrid) mode. The transmutation history of MA nuclear waste is included. MA are added into the first zone of the coolant in TRi-structural ISOtropic particle TRISO particles with a volume fraction of 6%. The transformation scenario for all MA by SS 304 steel FW is practically the same as with the ODS FW.

Reactor with metallic fuel and lead-208 coolant

The paper considers the concept of a fast lead cooled 25MW reactor for a variety of applications, including incineration of minor actinides, production of medical radioisotopes, testing of radiation-damaged nuclear technology materials, etc. A specific feature of the proposed reactor is rather a high neutron flux of 2.6·1015 n/(cm2·s) at the core center, high average neutron energy of 0.95 MeV at the core center, and a large fraction (40%) of hard neutrons (En > 0.8 MeV). The extremely high estimated reactor parameters are achieved thanks to the small core dimensions (DxH ≈ 0.50×0.42 m2), innovative metallic fuel of the Pu-Am-Np-Zr alloy, and the 208Pb enriched lead coolant. A relatively high probability of 241Am fission (about 50%) is achieved in the reactor core’s hard spectrum, this making it possible to incinerate up to 4 kg of 241Am during one reactor campaign of 1000 effective days.

Proton-induced spallation of mercury and xenon targets for neutron flux produced between CEM03.01 and INCL4-ABLA spallation packages

Accelerator-Driven Subcritical (ADS) reactors are one of the pathways considered for the incineration of minor actinides. They use a proton beam accelerator up to a GeV energy, bombarding a spallation target, usually made of a material of high atomic number (W, Pb, Ta or U). The neutrons generated during the spallation reactions are then multiplied. The main goal of this study is to make a comparison between two models of physical packages via the Monte Carlo code MCNP6. The first package is the INCL4-ABLA and the second one is the CEM03.01, and we have deduced the best package according to the neutron production rate. In this study, we are interested in two cylindrical targets of natural mercury and xenon 20cm in height and 20 in diameter and a proton beam varies from 0.1 to 1.5 GeV.

Proton-induced spallation of mercury and xenon targets for neutron flux produced between CEM03.01 and INCL4-ABLA spallation packages

Accelerator-Driven Subcritical (ADS) reactors are one of the pathways considered for the incineration of minor actinides. They use a proton beam accelerator up to a GeV energy, bombarding a spallation target, usually made of a material of high atomic number (W, Pb, Ta or U). The neutrons generated during the spallation reactions are then multiplied. The main goal of this study is to make a comparison between two models of physical packages via the Monte Carlo code MCNP6. The first package is the INCL4-ABLA and the second one is the CEM03.01, and we have deduced the best package according to the neutron production rate. In this study, we are interested in two cylindrical targets of natural mercury and xenon 20cm in height and 20 in diameter and a proton beam varies from 0.1 to 1.5 GeV.

Minor actinide incineration and Th-U breeding in a small FLiNaK Molten Salt Fast Reactor

Minor actinide (MA) accumulated in spent fuel is the primary contributor to the long-term radiological hazards of high-level nuclear waste. Due to its outstanding features such as fast neutron spectrum, no fuel assembly fabrication, on-line refueling and reprocessing, Molten Salt Fast Reactor (MSFR) is regarded as one of the candidate reactors for MA incineration. In the present work, we evaluate the MA incineration capability for a 500MWth FLiNaK MSFR by considering on-line MA feeding into the fuel salt and/or 233U extraction from the fertile salt, where the initial FLiNaK fuel salt is composed of 70mol% (LiF, NaF and KF) and 30mol% (ThF4, 233UF4 and MAF3). The simulated results show that, during a 50-year operation, the MA incineration efficiency with on-line MA feeding and 233U extraction scenario is significantly higher compared with the case without either MA feeding or 233U extraction. Based on the scenario of on-line MA feeding and 233U extraction, the initial MA loading is further adjusted to keep a small excess reactivity during the whole operation. And the MA incineration amount, the depletion ratio between MA and total heavy nuclides and the MA incineration efficiency at 50-year operating time are 7301kg, 77.4% and 33.3kg/TWh, respectively, with a doubling time of 39-year for 233U breeding. We also analyze and discuss the evolutions of the fuel temperature feedback coefficient, the effective delayed neutron fraction and the intensity of spontaneous fission neutron during the whole operating time.

Optimization of a mirror-based neutron source using differential evolution algorithm

This study is dedicated to the assessment of capabilities of gas-dynamic trap (GDT) and gas-dynamic multiple-mirror trap (GDMT) as potential neutron sources for subcritical hybrids. In mathematical terms the problem of the study has been formulated as determining the global maximum of fusion gain (Qpl), the latter represented as a function of trap parameters. A differential evolution method has been applied to perform the search. Considered in all calculations has been a configuration of the neutron source with 20 m long distance between the mirrors and 100 MW heating power. It is important to mention that the numerical study has also taken into account a number of constraints on plasma characteristics so as to provide physical credibility of searched-for trap configurations. According to the results obtained the traps considered have demonstrated fusion gain up to 0.2, depending on the constraints applied. This enables them to be used either as neutron sources within subcritical reactors for minor actinides incineration or as material-testing facilities.

Prospects of subcritical molten salt reactor for minor actinides incineration in closed fuel cycle

Abstract A subcritical molten salt reactor is proposed for minor actinides (separated from spent fuel VVER-1000 light water reactor) incineration and for 233U conversion from 232Th. Here the subcritical molten salt reactor with fuel composition of heavy nuclide fluorides in molten LiF–NaF–KF salt and with external neutron source, based on 1 GeV proton accelerator and molten salt cooled tungsten target is considered. The paper presents the results of parametrical analysis of equilibrium nuclide composition of molten salt reactor with minor actinides feed in dependence of core dimensions, average neutron flux and external neutron source intensity. Reactor design is defined; requirements to external neutron source are posed; heavy nuclides equilibrium and fuel cycle main parameters are calculated.

Hierarchically porous carbon–zirconium carbide spheres as potentially reusable transmutation targets

The preparation of hierarchically porous phase-pure carbon–zirconium carbide spheres with surface areas upwards of 70 m2/g and diameters in the 1–2 mm range has been achieved. The zirconium carbide beads were prepared through carbothermal reduction of polyacrylonitrile-zirconium composites prepared via three different routes including infiltration of a zirconium precursor into preformed polyacrylonitrile (PAN) beads and two one-pot co-precipitation methods. Depending on the route used to prepare the composites, relatively high surface area phase-pure zirconium carbides with the radial macroporous internal structure of the PAN template could be prepared. The adsorption of the elements U, Mo, Cs and Sr by the zirconium carbide beads was studied as a function of pH in the range 1–13 and target element concentrations of 0.025 mmol/L. The as-prepared hierarchical zirconium carbide beads demonstrated almost quantitative uptake of Mo below pH 4 with uptake decreasing to close to zero and above pH 10.5. On the other hand, U uptake was negligible below pH 4 and became quantitative in the range 4–12. Cesium was not extracted to any significant extent in the pH range studied and Sr only to a limited extent above pH 10. The engineered hierarchical porosities, neutron transparency, thermal stability and anticipated chemical stability of the present zirconium carbide spheres makes them promising candidates as reusable transmutation targets for Minor Actinide incineration.

Plutonium and Minor Actinides incineration options using innovative Na-cooled fast reactors: Impacting on phasing-out and on-going fuel cycles

The flexibility of innovative Na-cooled fast reactors for burning Pu and/or Minor Actinides (MA) is investigated with respect to different fuel cycle strategies. Under phasing-out conditions, the burner systems are used for reducing to a minimum level the accumulated TRansUranic (TRU) inventory, whereas when continuous use of nuclear energy is envisaged (on-going case), burner systems may be dedicated to MA management only.As an example of a phasing-out case, the accumulated German TRU inventory (at 2022) is assumed to be transmuted in a chosen time period of 150 years. For this purpose, two different burner fast reactors concepts, developed at KIT, are deployed in a Partitioning and Transmutation based fuel cycle. The effects are analyzed in order to confirm the behavior expected by the neutronics studies and to provide a basis for further optimization of the scenarios with respect to a number of reactors, deployment paces and fuel compositions.Additionally the performance of the MA burner is assessed to provide an effective MA mass stabilization in case of a continuous use of nuclear energy. Preliminary results are compared with those of past studies based on the European Sodium-cooled Fast Reactor.

Metallic inert matrix fuel concept for minor actinides incineration to achieve ultra-high burn-up

The advantages of using Inert Matrix Fuel (IMF) in a design of an isolated arrangement of fuel are considered, with emphasis on, low temperatures in the fuel center, achievement of high burn-ups, and an environment friendly process for the fuel element fabrication. Changes in the currently existing concept of IMF usage are suggested, involving novel IMF design in the nuclear fuel cycle.

Impact of Resonance Treatment on Minor Actinide Incineration in a D–T Thorium Fusion Concept

Resonance treatments have an essential role to reliable neutronic calculations with different neutronic parameters. In this study presents the effect of resonance treatment and various tritium breeder materials on the incineration of the nitride fuels containing minor actinide mixed thoria in the Deuterium–Tritium fusion driven hybrid reactor as time dependent. Neutron transport calculations under resonance treatment and without resonance treatment are performed by using XSDRNPM/SCALE 5 codes. The impact of resonance treatments and various tritium breeder materials on tritium breeding, energy multiplication, total fission rate (∑f), cumulative fissile fuel enrichment, fissile fuel breeding, average burn up values are comparatively investigated. It is observed that the neutronic results affect from both resonance treatment and the tritium breeder materials as time dependent.

On a fast reactor cycle scheme that incorporates a thoria-based minor actinide-containing cermet fuel

A fast reactor cycle scheme that incorporates a thoria-based minor actinide-containing cermet fuel is given. The present cermet fuel consists of an oxide solid solution of Th and minor actinides and Mo-inert matrix. It has been proposed as a high-performance device that can enhance minor actinide incineration in a fast reactor cycle. It is used in an independent small sub-cycle, whereby dedicated cycle technologies are adopted. Two-step reprocessing process was proposed for the present cermet fuel; it consists of a pre-removal of Mo-inert matrix and an actinide recovery. A preliminary test for the pre-removal of Mo-inert matrix was carried out using a surrogate cermet fuel. Burnup characteristics of a fast reactor core loaded with the cermet fuel were investigated by using neutronic calculation codes. It was revealed that a heterogeneous composition of Mo-inert matrix in inner and outer cores may lead to an effective transmutation of minor actinides and a flattened power density. It was concluded that the present cermet fuel was potentially promising as a high-performance incineration device of minor actinides for fast reactors.

Fuel cycle analysis of GFR using advanced fuels

The potential benefits of a synergistic light-water reactor (LWR) and gas-cooled fast reactor (GFR) fuel cycle were evaluated for its impact on the front-end and back-end of the fuel cycle. Comparisons were made with conventional once-through cycle (OTC) and MOX fuel cycle. Variations in the synergistic LWR/GFR fuel cycles were based on the degree of recycle in the LWR including both plutonium and reprocessed uranium with concomitant impact on used LWR fuel inventory. This provided a wide range in composition of GFR feed from low to high plutonium content with impact on transmutation/incineration within the GFR fuel cycle. Self-recycle of all actinides was modeled for the GFR with analyses demonstrating that the GFR can be sustained on and consequently accept a wide range of feed materials. Analyses were done using Monteburns along with MCNP and Origen2.2 to model a 60-year period corresponding to the anticipated lifetime of supposed contemporary LWRs and GFRs. All cycles were evaluated based on actinide total mass and isotopic inventory, radiotoxicity, heatload, and resource requirements including natural uranium and SWU. For comparison, all fuel cycles were normalized based on 1 TWHe output. Improvements in fuel cycle performance are dictated by the production and incineration of minor actinides in the GFR and their continued recycle may not be feasible due to the buildup of troublesome isotopes such as 244Cm and 252Cf. But where uranium and plutonium continue to be recycled beyond the 60-year period, the LWR/GFR cycles demonstrated order of magnitude reductions in used fuel inventories, heatload, and radiotoxicity on a per TWHe basis over LWR only cycles. The full details of the advanced fuel cycle methodology and results are presented.

The GDT-based fusion neutron source as driver of a minor actinides burner

At present, accelerator driven systems seem to have a good chance to play an important role in a long-term sustainable utilization of fission reactor technology. Compared with reactors, such systems offer several advantages regarding the incineration of minor actinides. However, they demand intense neutron sources. The spallation neutron source is favoured for this purpose because of its high neutron emission intensity, which seems to be attainable. The Budker Institute of Nuclear Physics Novosibirsk is developing a 14 MeV fusion neutron source on the base of the gas dynamic trap (GDT), which is primarily destined for an irradiation facility for fusion material research. The potential of this neutron source as driver of a minor actinides burner was studied by means of neutron transport calculations and compared with a spallation source. To this end, a simple model of a burner was derived from an international numerical benchmark exercise that was conducted by the Nuclear Energy Agency of the OECD. The paper presents and discusses the main results of the study and draws the conclusion that both the source strength and the efficiency of the GDT-based neutron source must be substantially increased. Moreover, advices are derived, which show that by stretching the neutron production volume and by raising the electron temperature of the GDT plasma the desired improvements could be accomplished.