Nuclear Waste Transmutation Research Articles

Preliminary study of transuranic transmutation in a small modular chloride salt fast reactor

The management of radioactive waste poses a significant challenge to the sustainable development of nuclear energy. Efficient transmutation of nuclear wastes is crucial to minimize their accumulation. A small modular chloride salt fast reactor (sm-MCFR) capable of transmuting transuranic elements (TRU) is proposed in this paper, combining the advantages of the small modular reactor (SMR) and the molten salt reactor (MSR). The sm-MCFR is characterized by a high fuel loading and a compact core structure that can be quickly deployed around large commercial reactors to achieve TRU transmutation. To evaluate the TRU burnup capability of the sm-MCFR, several fuel salts and reprocessing modes were analyzed using the internally developed TRITON MODEC Coupled Burnup Code (TMCBurnup) tool. NaCl-MgCl3 with 98 % enrichment in 37Cl is chosen as carrier salt for the sm-MCFR, which can achieve 76.7 % TRU transmutation rate in average and 355 kg·GW−1·a−1 TRU transmutation quality at a continuous reprocessing rate of 10 L/d for 50 operation years. The optimized sm-MCFR reduced the radioactive toxicity of TRU by 84 %, thereby simplifying waste reprocessing. In addition, the sm-MCFR has a negative temperature feedback coefficient of −7.195 pcm/K, favoring safe reactor operation.

Evaluating Material Attractiveness of Minor Actinide Nuclear Fuel Intended for a Waste Transmutation Lead-Cooled Fast Reactor

Long-lived high-level waste from commercial nuclear power reactors is a problem that concerns stakeholders and scientists working in the back end of the nuclear fuel cycle. Nuclear waste transmutation is under investigation to tackle this problem, transforming nuclides that represent a long-term source of radioactivity, radiotoxicity, and heat into short-lived or stable nuclides. However, the transmutation process will require that several long-lived isotopes be separated from the spent nuclear fuel, which raises proliferation concerns. In this paper, we perform an investigation of the attractiveness characteristics related to the material used in a lead-cooled fast reactor system concept designed to burn minor actinides before and after irradiation. The materials evaluated are separated uranium, neptunium, plutonium, americium, and curium. We also evaluated grouped product materials, neptunium + americium and neptunium + plutonium. Additionally, we present potential safeguards and physical protection implications for the proposed materials. The main conclusion of this paper is that the separated neptunium and plutonium generated by the fast reactor are materials that deserve attention mainly related to physical protection measures.

The influence of source locations of different multi-beam concept on the power density distribution and nuclear waste transmutation performance for ADS–NWT

The multi-beam concept for the accelerator driven subcritical reactor (ADS) has the advantages in reducing the requirements for the accelerator and can flatten the power distribution of the core. However, the layout of the multi-beam should meet the goal of balancing the relationship between the total power density, the neutron leakage, and the nuclear waste transmutation performance. For the further study of the multi-beam concept for the ADS, this paper focuses on the influence of the source locations of different multi-beam concept on the power density distribution, the neutron leakage, and the nuclear waste transmutation performance for the ADS for a nuclear waste transmutation reactor (ADS–NWT). The different variations on the radial power density distribution and the nuclear waste transmutation performance for the three-, four-, six-, and seven-beam concept for ADS–NWT were analyzed. The results show that the related neutronics performance for the multi-beam concept for ADS-NWT is not only affected by the external source locations and the core geometry, but also affected by the keff and the interference effect between the spallation targets. Moreover, the fission reaction ratio and the fission reaction / (n, gamma) absorbtion ratio for minor actinides (MA) and plutonium (Pu) for the few beams concept are both higher than that for the many beams concept. Based on the analysis of the influence of source locations on the neutron spectrum, the total neutron flux, and the nuclear waste transmutation performance for MA and Pu, the related conclusions about how to arrange the best beam position from the perspective of nuclear waste transmutation and production capacity are provided.

Theoretical investigations into the bonding and separation properties of non-rigid, partially rigid, and rigid ligands derived from Et-Tol-PTA with trivalent lanthanides and actinides.

The separation of trivalent actinide elements from lanthanide elements represents one of the most formidable challenges within the context of nuclear waste partitioning and transmutation (P&T) processes. Consequently, we embarked on a systematic investigation aimed at elucidating the bonding properties and thermodynamic behavior of a N-ethyl-N-tolyl-2-amide-1,10-phenanthroline (Et-Tol-PTA) ligand in conjunction with trivalent actinide and lanthanide elements. This investigation involved the utilization of various density functional theory (DFT) methods and a comparative analysis between small-core pseudopotential basis sets and all-electron basis sets. It was found that well-performing results were achieved using the PBE0 functional in both bond length and thermodynamic energy calculations, with minimal impact being exerted by the basis set on the results. Furthermore, an exploration was carried out into the bonding and thermodynamic properties of trivalent actinides and lanthanides with ligands derived from Et-Tol-PTA, encompassing non-rigid (La), partially rigid (Lb, Lc), and rigid (Ld) ligands. Thermodynamically, advantages in the separation of Am(III)/Eu(III) were exhibited by Lb and Lc ligands, while excellent performance in the separation of Am(III)/Cm(III) was demonstrated by the La ligand. Analyses conducted using quantum theory of atoms in molecules (QTAIM), reduced density gradient (RDG), and natural bond orbital (NBO) methodologies revealed the presence of partial covalent character in the bonds between oxygen (O) and metal (M), as well as between nitrogen (N) and metal (M), with a higher degree of covalent character being observed in O-Am and N-Am bonds compared to O-Cm/Eu and N-Cm/Eu interactions.

New Waste Management Options Created by iMAGINE through Direct Operation on Spent Nuclear Fuel Feed

The demand for improving the nuclear waste management has since long been identified as one of the major hurdles for widespread use of nuclear energy. Nuclear waste management, through partitioning and transmutation (P&T), has been researched since the 1990s with partitioning being a prerequisite for the process. Recently, an innovative approach of reactors directly operating on spent, or nowadays often called used nuclear fuel, iMAGINE has been proposed which could deliver on the aims of P&T as a side effect to more efficient and sustainable nuclear energy production in the future. A HELIOS model of the core has been used to analyze the long-term operation of a molten salt reactor including the investigation of the minor actinide accumulation over the entire burnup period. The results shown here confirm that long-term reactor operation is possible, even with higher amounts of vitrified waste loaded. Thus, it is possible to achieve the aims of P&T without prior partitioning, but it is certainly less efficient since the high concentration of minor actinides (MAs), required for efficient burning, is impossible to obtain in a short operational time. On this basis, the proposed nuclear waste management approach will be a long-term effort when it is accomplished without partitioning/separation technologies. However, none of the analyses contradicts this effort. The key points are: (a) when the technology for treating the waste is possible and reliable, the time horizon will not be a major concern; (b) the waste management is now intrinsically linked with energy production instead of requiring dedicated costly facilities, delivering a promising economic basis; (c) the waste management is now associated with long-term energy production and massively improved resource utilization. The study of feedback effects has shown that the modeled system has a strong negative feedback effect of ~−6 pcm/K, and even with spent nuclear fuel feed reduces to ~−3.8 pcm/K, ensuring the basis for a safe operation. In summary, it has been demonstrated that the objectives of P&T are achievable without prior partitioning, an approach which was never even discussed in the past. These ground-breaking results and the new insights will allow or even require rethinking the nuclear waste management of the future.

Clustering fusion algorithm for selection of historical beam steady-state transmission data in CAFe facility

Accelerator-driven systems (ADS) are promising technologies for nuclear waste transmutation and energy production. The China ADS Front-end Superconducting Demo Linac (CAFe) is a prototype of the China Initiative Accelerator Driven System (CiADS), which aims to verify the feasibility of key technologies of CiADS. In this article, a novel method for historical data screening of the beam transport in the medium energy beam transport (MEBT) section of CAFe is presented. A clustering fusion algorithm based on unsupervised learning and beam transmission characteristics is designed to extract a large number of samples with the beam in steady-state transmission from historical beam data. A deep neural network model was constructed to fit the beam transport characteristics and verify the reliability of the screened data samples. The method can improve the efficiency and accuracy of data analysis and provide valuable insights for the optimization and control of beam transport in CAFe.

Speculative High-Level Nuclear Waste Processing Methods

With the exception of recent progress in Sweden and Finland, traditional and planned high-level nuclear waste processing disposal methods have yet to be successfully implemented on a commercial scale. Failure to solve the high-level waste issue is an impediment to the expansion of nuclear power production particularly in North America and Europe. Given these issues, this paper presents three speculative methods and an accelerator/reactor approach to process high-level nuclear waste. The first speculative approach explores the possibility of incorporating an underground nuclear detonation as a means to process the waste. A second method uses antiprotons and antineutrons to transmute the high-level waste. The third utilizes plasma to obliterate the isotopes comprising the radioactive waste. A somewhat less speculative approach involving nuclear waste transmutation using a reactor or accelerator is also presented. However, this approach has several issues to overcome and is not a likely a near-term solution to the waste disposal/processing issue.

Design and experimental validation of differential pumped vacuum system for HINEG windowless gas target

High Intensity Neutron Generator (HINEG) is a facility aiming for neutronics design validation, materials and components irradiation test, nuclear waste transmutation and nuclear technology applications, etc. A high-pressure windowless gas target system is adopted for HINEG to achieve a higher neutron yield. Therefore, a three-stage differential vacuum system has been developed to couple the pressure transition of 6 orders of magnitude to the 300 kV/50 mA high-intensity beam transport in a series of large-diameter pipes. A suitable pumping system has been developed to increases the target pressure which is used to generate D-T neutron yield of 2.1 × 1013 n/s. A windowless pressure transition from high gas pressure (1 × 103 Pa) in the gas target chamber to high vacuum (3 × 10−3 Pa) in the accelerator is reached. In this paper, a new differential pumped vacuum system is designed for HIENG. Innovatively, the newly designed differential vacuum system can not only meet the transport requirements of the high-current low-energy deuteron beam, but also couple the pressure transition of 6 orders of magnitude, which can solve the contradiction of the large-diameter pipe required for high-intensity beam transport and the small diameter pipe needed for the high-pressure difference.The experimental validation and design optimization of differential pumped vacuum system is introduced. The experiment results indicate the validity of the theoretical calculation. Compared to molecular pumps, roots pumps might be more suitable for windowless differential vacuum system coupled to large-diameter apertures and pipes. The design could be used for reference in other windowless gas target systems coupled to accelerator with high-current low-energy ion beam.

Study of External Source Effects for Different Multibeam Concept for Accelerator-Driven Subcritical Reactor for Nuclear Waste Transmutation (ADS-NWT)

The multibeam concept for the accelerator-driven subcritical reactor (ADS) has advantages in the power distribution of the core, and it could reduce the requirement of proton beam current intensity for each beam. In comparison with the single-beam concept, the multibeam concept could also reduce the thermal load of the beam window significantly. This paper focuses on the study of external source effects for different multibeam concepts for an ADS for nuclear waste transmutation (ADS-NWT). The different multibeam concepts include the three-beam, four-beam, six-beam, and seven-beam concepts for the ADS-NWT. By using the calculation tools FLUKA and SuperMC with the nuclear data library ENDF/B-VII.1, the variations of the keff and total power, as the function of the position of the spallation targets, are provided for each multibeam concept. The results show that the keff and total power were affected by an interference effect between the spallation targets. For the transport of fission neutrons in the core, the maximum radius of the interference effect between the spallation targets was 40 cm. Considering the transport of spallation neutrons in the ADS-NWT, the maximum radius of the interference effect between the spallation targets was 60 cm. The spallation targets were moved from the inner circle to the outer circle of the fuel zone, and the different variations in keff and total power trend for the three-beam, four-beam, six-beam, and seven-beam concepts for the ADS-NWT were obtained.

Laser-based photonuclear production of medical isotopes and nuclear waste transmutation

The results of complex simulations using PIC-GEANT4 (particle-in-cell and Monte-Carlo) codes based on the generation of a high-energy electron bunch by a short laser pulse propagating in a relativistic self-trapping regime in a near-critical plasma has been applied to assess the possibility of medical isotope production and nuclear waste transmutation. It has been demonstrated that a 10 Hz 30 fs 4 J laser pulse is well suited to the production of therapeutic amounts of several standard medical radionuclides (111In, 123I, 103Pd, 62Cu, 64Cu). The use of direct electron irradiation has an advantage over the use of bremsstrahlung gamma radiation from the converter due to the simplification of the production scheme without loss of radionuclide yield. The study of the transmutation of long-lived fusion products showed low efficiency and the need for preliminary isotope separation. Achieving as little as 10% reduction in the activity of a 10 g sample requires the continuous operation of the next-generation laser system at a high repetition rate (1 MHz–100 kHz) for (one to ten) years.

Development of 1A-class linear accelerator for nuclear transmutation of nuclear wastes

Development of 1A-class linear accelerator for nuclear transmutation of nuclear wastes

Selective extraction of Am(III) from Cm(III) and Eu(III) using a novel phenanthrolinamide ligand: Thermodynamics, species, and structure

The separation of tridentate actinides from lanthanides and that between Am(III) and Cm(III) is one of the most challenging tasks in the process of partitioning–transmutation of nuclear waste. Herein, we report a novel tridentate N–N–O hybrid ligand, i.e., N-ethyl-N-tolyl-2-amide-1,10-phenanthroline (ETPhenAm), for the selective extraction toward Am(III)/Cm(III) over Eu(III) from HNO3 solutions. ETPhenAm in 3-nitrotrifluorotoluene has excellent Am(III) extraction ability and Am(III)/Eu(III) separation performance. Notably, the good selectivity of ETPhenAm toward Am(III) over Cm(III) suggests that ETPhenAm exhibits good ionic radius selectivity in the interaction with the intraseries cations such as Am(III) and Cm(III). Extractive complexes of formula M(NO3)3L and M(NO3)3L2 (M = Am, Cm, and Eu) are identified in the extraction system. HNO3 concentration in the aqueous phase heavily affects the composition of the extractive species in the organic phase because of the strong basicity of ETPhenAm. Extraction observations are validated by determining the protonation constants of the ligand and the reaction equilibrium constants for the generation of complexes M(NO3)3L and M(NO3)3L2 (M = Nd and Eu) in CH3OH/10%(v)H2O solution. The structures of complexes M(NO3)3L and M(NO3)3L2 are confirmed through X-ray diffraction of compounds [La(NO3)3(ETPhenAm)2] (I), [Nd(NO3)2(ETPhenAm)2]NO3 (II), and [Eu(NO3)2(ETPhenAm)2]NO3∙[Eu(NO3)3(ETPhenAm)(CH3OH)]∙2CH3OH (III).

Thermo-mechanical behaviour of the MYRRHA core support structure

This work is part of the research on the design of MYRRHA primary system. MYRRHA, a multipurpose research facility, being developed at SCK-CEN, is based on the Accelerator Driven System (ADS) concept. This research facility intends to show the applicability of the ADS principle, and aims to demonstrate, among other objectives, the feasibility of efficient transmutation of high-level nuclear waste. The present study investigates, mainly, the thermo-mechanical behaviour of the core support bundle, fuel assemblies and core devices, in particular, the deformation in terms of the displacement and gaps distribution. To reach this goal, which is arduous, an efficient modeling methodology based on the finite element method has been devised. An effective solution strategy has been formulated to carry out a highly challenging analysis due to severe numerical issues and instabilities, making the solutions particularly difficult to reach for some cases. Analysis cases are differentiated by some friction coefficient values; and the presence or not of a spring pre-compression force. Despite serious convergence issues, converged solutions have been achieved for all the cases studied. From the results analysis, a remarkable feature observed is the stability of the solutions obtained. And, on the whole, the value of the friction coefficient used does not change significantly the displacement, the gaps patterns, and the stresses. Similarly, account of the pre-compression force does not affect much the global behaviour. The kinematic quantities found should be useful for some preliminary reactivity predictions of the reactor.

Physics design of a coupled RFQ-DTL cavity for a compact neutron source

Neutrons have been widely used, such as neutron imaging, transmutation of nuclear waste, Neutron Capture Therapy (NCT), et al. Neutrons are generally produced by nuclear reactors or accelerator-driven neutron sources. Comparing with nuclear reactors, accelerator-driven neutron sources have many advantages, such as compact and cheap. We proposed a coupled RFQ-DTL cavity as the main component of a miniaturized neutron source. A high accelerating efficiency was achieved by placing Radio Frequency Quadrupole (RFQ) and Drift Tube Linac (DTL) in one cavity. Since the fringe field will increase the energy spread, we propose a method using the fringe field as a buncher to reduce the negative effect, and it will eliminate the use of external bunchers. The coupled RFQ-DTL cavity will accelerator a 25 mA proton beam from 35 keV to 2.5 MeV within 2.23 m. RFQ section and DTL section are 1.55 m and 0.68 m, respectively. Compared with a conventional RFQ, the length of coupled RFQ-DTL is shorten by 53%. The total transmission efficiency is 98.55%. A variable energy design for the beam output energy can be achieved by changing the insertion depth of the DTL tuner and controlling the feeding power of the cavity.

High power particle accelerator for driving the nuclear waste transmutation system at nuclear power plant

The existing technology for treatment of NPP nuclear waste is a deep geological disposal technology. But this technology is still has possibility of leakage due to heating caused by radioactivity process from long lived elements of nuclear waste. The development of high power particle accelerator technology which can produce high power proton beam to induce nuclear spallation reaction has introduced a new accelerator-based technology for treatment of nuclear waste from NPP named the Accelerator-driven Transmutation Waste (ATW). This paper describes result of study on status of accelerator based transmutation system for NPP nuclear waste. ATW is subcritical nuclear reactor combined with high power proton accelerator for transmutation of nuclear waste utilizing neutrons from spallation reaction due to collision of high power proton beam with heavy metal material as spallation target. The types of accelerators used for ATW are cyclotron, linac and synchrotron which able to produce proton beam of 10 up to 100 MW beam power, 10 up to 100 mA beam current and 0.6 up to 1.5 GeV beam energy. Demonstration scale ATW requires 1 up to 2 MW of proton beam power to generate 50 up to 100 MWth of thermal power in subcritical reactor. Whereas industrial ATW requires 10 up to 75 MW of proton beam power to generate hundred MWth of thermal power in subcritical reactor. Several countries have roadmap of ADS research and development which the main goal of nuclear waste transmutation using ADS called ATW. ATW also can be used for electrical energy production which is safer and more reliable.

Nuclear waste transmutation and tritium production in a fusion-driven hybrid reactor based on different reflectors

Abstract A comprehensive comparison neutronics calculation has been performed to investigate the effect of different reflectors on the tritium production and transmutation of minor actinides in a fusion hybrid reactor. Fusion-Driven Subcritical Spent Fuel Burner hybrid reactor (FDS-SFB), has been chosen as the computational model. All the designs and calculations have been carried out by using SuperMC Monte-Carlo code with ENDF/B-VII neutron cross-section library. Beryllium (Be), beryllium oxide (BeO), HT-9 stainless steel, tungsten carbide (WC), magnesium oxide (MgO), silicon carbide (SiC), and titanium carbide (TiC) have been selected as the candidate reflectors. Other neutronic parameters, such as the neutron flux spectrum as well as the effective multiplication factor (keff), have been calculated for each of the reflectors to understand the behavior of each reflector material. In the case of tritium production, as expected, Be and BeO had the best performance and if a non-beryllium reflector was considered, MgO could be the best. In the case of transmutation of TRUs, for 241Am, 243Am, and 242Pu, BeO had the best results. For 244Cm and 237Np, TiC and MgO were the best choices respectively. For the total transmutation, BeO, MgO, and WC had the best capability to transmute TRUs among the other reflectors. Considering both parameters, i.e., TRUs transmutation and Tritium production, BeO (among beryllium-reflectors) and MgO (among non-beryllium reflectors) could be an appropriate choice to fusion hybrid designs.

Preliminary results on the study of 237Np(n,f) at n_TOF/EAR2 in energies related to nuclear waste transmutation

The accurate knowledge of neutron-induced fission cross sections of isotopes involved in the nuclear fuel cycle is essential for the optimum design and safe operation of next generation nuclear systems. Such experimental data can additionally provide constraints for the adjustment of nuclear model parameters used in the evaluation process, resulting in a further understanding of the nuclear fission process. In this respect measurements of the 237Np(n,f) cross section have been performed at the n_TOF facility at CERN in the horizontal 185 m flight-path (EAR1) which were discrepant by 7% in the MeV region. The neutron-induced fission cross section of 237Np(n,f) was recently restudied at the EAR2 19.5 m vertical beam-line at CERN’s n_TOF facility, over a wide range of neutron energies, from 100 keV up to 15 MeV, using the time-of-flight technique and a modern set-up based on Micromegas detectors. This study was performed in an attempt to resolve the aforementioned discrepancies and to provide accurate data of a reaction that is frequently used as reference in measurements related to feasibility and design studies of advanced nuclear systems. Preliminary results with a high statistical accuracy that resolve the discrepancies will be presented along with a brief discussion concerning the facility and the analysis.

Transmutation of 129I Containing Nuclear Waste by Proton Bombardment

As a long-lived nuclear waste, iodine (129I) which is primarily generated from nuclear fission of uranium or plutonium is considered harmful to human and environment. Therefore, proper steps are required to treat the radioactive isotope. In this work, we propose transmutation of I-129 isotope using cyclotron-based proton bombardment. The TALYS code was employed to calculate nuclear cross-sections of various nuclear reactions possible to transmute I-129 into shortlived radionuclides or stable isotopes. Twelve different nuclear reactions, namely (p,n), (p,2n), (p,3n), (p,np), (p,2np), (p,a), (p,d), (p,2p), (p,nα), (p,γ), (p,3He), and (p,t), which have significant cross-sections (greater than 1 mb) were analyzed. Based on the nuclear reactions, there were 6 possible stable isotopes produced from proton bombardment of 129I, such as 129Xe, 128Xe, 127I, 126Te, 125Te and 130Xe. Short-lived radionuclides such as 29mXe (half life = 8.88 days), 127Xe (half life = 36.35 days), 127mXe (half life = 62.9 seconds), 128I (half life = 24.99 minutes), 125mTe (half life = 57.40 days), 127Te (half life = 9.35 hours) and 127mTe (half life = 106.1 days) could also be possibly produced from the proton irradiation of 129I nuclear waste. To sum up, the longest radionuclide could be generated from proton bombardment of 129I nuclear waste is 127mTe (half life = 106.1 days), which is a lot shorter than 129I (half life = 1.57x107 years). This theoretical study indicates that transmutation of 129I nuclear waste by proton bombardment into short-lived radionuclides is greatly feasible. Current available cyclotrons in Indonesia may be employed to help transmute 129I into short-lived radionuclides or stable isotopes. This theoretical study can be used as a reference for future 129I nuclear waste if proton beams are employed in the transmutation.

First results of the 241Am(n,f) cross section measurement at the Experimental Area 2 of the n_TOF facility at CERN

Feasibility, design and sensitivity studies on innovative nuclear reactors that could address the issue of nuclear waste transmutation using fuels enriched in minor actinides, require high accuracy cross section data for a variety of neutron-induced reactions from thermal energies to several tens of MeV. The isotope 241Am (T1/2= 433 years) is present in high-level nuclear waste (HLW), representing about 1.8 % of the actinide mass in spent PWR UOx fuel. Its importance increases with cooling time due to additional production from the β-decay of 241Pu with a half-life of 14.3 years. The production rate of 241 Am in conventional reactors, including its further accumulation through the decay of 241Pu and its destruction through transmutation/incineration are very important parameters for the design of any recycling solution. In the present work, the 241 Am(n,f) reaction cross-section was measured using Micromegas detectors at the Experimental Area 2 of the n_TOF facility at CERN. For the measurement, the 235U(n,f) and 238U(n,f) reference reactions were used for the determination of the neutron flux. In the present work an overview of the experimental setup and the adopted data analysis techniques is given along with preliminary results.

Spallation reaction study for long-lived fission products in nuclear waste

Spallation reaction for the long-lived fission product 107Pd has been studied for the purpose of nuclear waste transmutation. The isotopic-distribution cross sections on both proton and deuteron were obtained at 118 MeV/nucleon in inverse kinematics at the RIKEN Radioactive Isotope Beam Factory. A large cross-section difference was found between the proton and deuteron results for the light-mass products. The data were compared with the SPACS semi-empirical parameterization and the PHITS calculations including both the intranuclear cascade and evaporation processes. In addition, the potential of spallation reaction for transmutation of 107Pd is discussed.