Minor Actinides Research Articles

Calculations of the principal neutronic characteristics of a hypothetical VVER assembly with minor actinides incorporated into PuO2-ThO2 fuel in a duplex configuration

Calculations of the principal neutronic characteristics of a hypothetical VVER assembly with minor actinides incorporated into PuO2-ThO2 fuel in a duplex configuration

Developing A Digital Twin of a TRUEX Extraction Process that Enables Non-proliferation and Safeguards Monitoring through Optical Spectroscopy and Machine Learning

Background The recycling of spent fuel can increase the longevity and decrease the waste footprint associated with the nuclear fuel cycle. However, nuclear fuel reprocessing introduces the risk of proliferation of fissile material. For example, the recycling of uranium (U) and plutonium (Pu) is generally accomplished via a solvent extraction process. The separation efficiency of the process is dependent on a variety of parameters such as acid concentration, organic to aqueous (O/A) volume ratio, number of separation stages, etc. A change in any of these parameters could result in different extraction conditions of fissile material and lead to a process upset with potential safeguards and proliferation concerns. Thus, the ability to monitor the solvent extraction process in real-time would enable the rapid identification and diagnosis of process upsets. Methods In this work, we discuss the development of a digital twin of a surrogate small-scale nuclear fuel reprocessing solvent extraction process utilizing online optical spectroscopic monitoring, machine learning (ML), and chemical modeling. A cascade of centrifugal contactors was employed with the TRUEX (TransUranic Extraction) solvent and surrogate minor actinides. Spectroscopic sensors were placed at strategic locations within the contactor bank and were used to predict TRUEX surrogate analyte concentrations at those locations. Results A solvent extraction digital twin consisting of the data acquisition, chemical model, and anomaly detection, provided expected concentrations at each contactor during normal operations. The digital twin demonstrated the ability to accurately simulate the extraction behavior under ideal conditions. Furthermore, the digital twin’s ML anomaly detection algorithm was shown to be successful in identifying several non-ideal process upset conditions.

Examination of the use of thorium-based fuel for burning minor actinides in European sodium cooled fast reactor

Abstract To reduce the potential and volume of radiotoxicity of radioactive wastes, minor actinides (MAs) elements recovered from discharged fuel are to be recycled. In this context, the main aim of this work is to examine the possibility of using thorium-based fuel for burning extra MAs in the fourth generation European Sodium cooled Fast Reactor (ESFR). For that MCNPX transport code was used to design a representative fuel assembly of this concept. The reference uranium-based fuel was uniformly replaced by thorium mixed with different fractions of MAs and burned for 2050 effective full power days. The influence of MAs on the evolution of reactivity and fuel transmutation is being investigated to ensure the performance of the fuel. Also, different neutronic and safety parameters such as neutron yield, neutron spectrum, neutron flux, effective delayed neutron fraction, Doppler constant, sodium void reactivity coefficient, were computed and their change with MAs content was investigated. Results indicate that recycling of MAs in ESFR using thorium-based fuel has a positive effect on reactivity and transmutation evolution but on other hand has a negative effect on safety parameters.

Real-time monitoring and protection strategies for dense granular flow spallation target in Accelerator-Driven System

An innovative concept of a high-power, gravity-driven, dense granular spallation target has been integrated into the prototype design of the China initiative Accelerator-Driven System (CiADS), which is envisioned as a promising nuclear demonstration facility aimed at converting long-lived fission products and minor actinides into short-lived isotopes. This spallation target system, bridging the proton beam accelerator and the subcritical reactor, plays a crucial role in ensuring safety control and facilitating dynamic operational strategies. In this study, we have developed a system design for an online detector array, incorporating several neutron fission chambers and thermocouples installed in the gap between the spallation target and the subcritical reactor to monitor operations. Furthermore, we have devised real-time monitoring and protection methods to tackle the challenges of measuring beam position and the blockage condition associated with granular flow, considering the distribution of neutron flux and heat deposition under abnormal irradiation conditions. Monte Carlo simulations have demonstrated the applicability and feasibility of the control system for the dense granular spallation target in the accelerator-driven system. The numerical results confirm that the proposed control system meets the requirements for stable measurement with acceptable accuracy.

The Effect of Adding Minor Actinide Fuel Rods on GFR Reactor in Radiopharmaceutical Waste Production Using OpenMC Program

GFR is a generation IV reactor based on helium gas refrigeration capable of working at very high temperatures. The fast spectrum in this reactor makes it possible to use nitride-based fuel, namely Uranium Plutonium Nitride (UN-PuN). Adding minor actinide (MA) material to the primary fuel, UN-PuN can maximize reactor performance to near critical from the beginning to the end of burn-up. This study aims to analyze the effect of adding MA fuel rods to the heterogeneous core of 5 fuel variations (F1, F2, F3, F4, F5) on the probability of radiopharmaceutical waste production. The method in this research is to place MA fuel rods in this study using four designs based on the highest neutron flux value in one fuel assembly. The results of the neutron flux calculation show that the reactor’s active core’s central region (F1, F2, F3) needs to be added to MA fuel rods so that the resulting flux is more evenly distributed. The calculation of reactor criticality shows that Np fuel rod design 4 and Am fuel rod design 1 have the best keff value (keff ≈ 1) among other designs. The burn-up of MA fuel rods produces a minimal probability of producing Tc99m, Sr89, Y90, Rh105, Ag111, I231, and Sm15 radiopharmaceutical waste, even less than 1 kg.

Extraction behaviors of minor actinides and rare earth elements with NTA amide extractants

ABSTRACT Nitrilotriacetamide (NTA amide) has been developed as an extractant for the mutual separation of minor actinides (MA; Am and Cm) and rare earth elements (REs) by solvent extraction. NTA amide is a tertiary amine in which the three carboxyl groups of nitrilotriacetic acid are amidated. NTA amides have six hydrophobic long-chain alkyl groups on the side chains, and they have enhanced solubility in organic solvents, thereby allowing for the use of nonpolar diluents, such as n-dodecane. This study examined the extraction behavior of MA and REs using four NTA amides with six side chains comprising n-octyl groups, 2-ethylhexyl groups, and a mixture of these groups. The results showed that the linear-type hexaoctyl NTA amide (HONTA) had the highest extraction ability for MA. Furthermore, the extraction behavior of MA and REs was different under low and high nitrate conditions, and that of scandium (Sc) was found to be considerably different depending on the side chains.

Data and modeling sensitivity analysis for molten salt fast reactor benchmark – Static calculations

The Molten Salt Reactor (MSR) idea is increasingly being recognized in the nuclear field due to its potential safety, sustainability, and economic efficiency advantages. The Molten Salt Fast Reactor (MSFR) benchmark, introduced in 2019, highlighted variations in results tied to different neutron cross-section libraries. This study investigates the impact of utilizing the ENDF/B-VIII.0 and JEFF-3.3 cross-section libraries for MSFR benchmark assessment compared to the ENDF/B-VII.1 database. Monte Carlo based open source code OpenMC is used for the analyses. Rigorous sensitivity analyses assess the influence of individual components, including the cross-section database, resonance elastic scattering, and Thermal Scattering Law (TSL). Beyond the criticality assessments, parameters such as delayed neutron fraction, temperature coefficient of reactivity, and neutron spectrum are compared for different cross-section libraries. Our analyses reveal that incorporating new evaluations for 233U (n,γ) and fission cross-sections in ENDF/B-VIII.0 significantly alters criticality results, i.e., more than 1700 pcm difference is seen between libraries. Similarly, critical concentration using ENDF/B-VII.1 and JEFF-3.3 is over-predicted by approximately 3%. The variations in Thermal Scattering Law (TSL) files do not yield substantial differences in outcomes due to the fast spectrum of the reactor. In some cases, the treatment of resonance elastic scattering leads to reactivity differences greater than 50 pcm. The benchmark compares 233U-started and Minor Actinide (MA)-started core. From the reactor physics point of view, the MA-started core leads to a 29% higher (n, γ) reaction rate than the 233U-started core. A 3–4% smaller value of thermal reactivity coefficient is obtained using the ENDF/B-VIII.0 library compared to the ENDF/B-VII.1 value. Using the ENDF/B-VIII.0 for the MSFR benchmark signifies using newer and better data for the GEN-IV reactors neutron physics calculations.

Enhancing fuel breed-burn performance in a sodium-cooled fast reactor using a novel reactivity control method

This study aims to enhance the fuel cycle and fissile breeding performance of a sodium-cooled fast breeder reactor (FBR) by utilizing minor actinides (MAs) as a means of reactivity control alongside partially-inserted control rods. Choosing the PFBR-500 as the reference design, four core models, designated as Cases A, B, C, and D, utilizing various proportions of minor actinides (MAs) were built and simulated using OpenMC. The MA concentrations were optimized to compensate for the withdrawal of control rods and ensure the same initial reactivity for all cores. Burnup analysis over 365 EFPDs revealed a significant increase in cycle length and burnup for the modified cores along with a modest rise in breeding ratio. Notably, Case C, employing 3.45 wt.% MAs in the 88 inner-core fuel subassemblies achieved an extra 62.25 EFPDs cycle length, a 33.74% rise in single-cycle burnup, and a 3.86% increase in breeding gain compared to the reference. Loading MAs into the inner-core region proved to be more effective in enhancing both fertile-to-fissile conversion (thus extending the cycle length and fuel burnup) and transmutation than utilizing MAs throughout the core due to greater neutron flux at the core center. While Case D utilizing 2.2 wt.% MAs both in the inner and outer core fuel subassemblies had the highest overall MA loading, it demonstrated lower increments in cycle length, burnup, and breeding gain compared to Case C. Case C also exhibited the highest overall destruction rate (approximately 24%/y) for Np and Am isotopes, and successfully transmuted around 24.08 kg 237Np, 9.33 kg 241Am and 3.82 kg 243Am over the course of a single year. The addition of MAs also achieved a slight flattening of the axial and radial flux profile and a decrease in the flux peaking factor. However, it slightly lowered the beta-effective, Doppler constant, and control rod assembly worth, and shifted the coolant void reactivity worth to the positive side.

Searching for an optimal fuel for a supercritical water reactor (SCWR) to dispose of minor actinides resulting from the nuclear waste

This work investigates the optimal treatment of Minor Actinides (MAs) produced in the spent fuel of nuclear reactors operating around the world. This process is very important in addressing the challenges associated with nuclear waste and reducing environmental impact. A three-dimensional model of a supercritical water reactor (SCWR) has been designed using MCNPX to make a comprehensive neutronic analysis. MAs with a concentration of 1 % have been added to UO2, (Th, U)O2, (Th, 233U)O2 and (Th, rgPu)O2. These fuels have been uploaded in the SCWR assembly and burned in a separate fuel cycle. The infinity multiplication factor (kinf) of the suggested has been investigated with and without adding the minor actinides to analyze the effect of MAs on the reactor reactivity. The fuel constituents, plutonium concentration, MAs concentration and transmutation rate have been tracked with fuel burnup. The reactivity temperature coefficients have been calculated for the suggested cases to ensure the validity of the suggested fuels. The power peaking factor (PPF) and radial power distribution have been calculated for the suggested fuels. The neutronic analysis confirms the suitability of the suggested fuel in burning a significant amount of MAs.

Temperature dependences of phase composition, densification, and thermal/chemical stability of Sr0.5Zr2(PO4)3-Nd0.5Sm0.5PO4 composite ceramics for nuclear waste form

A kind of Sr0.5Zr2(PO4)3-Nd0.5Sm0.5PO4 (hereinafter SrZP-Nd0.5Sm0.5PO4) composite ceramics was prepared through a one-step microwave sintering technique, where Sr, Nd, Sm were simulated for fission products and minor actinides (MA). The crystal structure and morphology of constituent phases were systematically characterized by XRD, TEM, and SEM techniques. The temperature dependent phase analysis results of SrZP-Nd0.5Sm0.5PO4 composite ceramics illustrate that with evaluating temperature monazite structure formed primarily, and the SrZP phase formed subsequently. With the temperature rising further, the monazite phase finally transformed into Nd0.5Sm0.5PO4 monazite solid solution. XRD refinement and TEM measurements are used to verify the crystal structures of the as-prepared SrZP and Nd0.5Sm0.5PO4 phases in the composite ceramics. The dependence of densification behaviors on temperature shows that the distinct densification of the SrZP-Nd0.5Sm0.5PO4 composite ceramics occurs in the temperature range of 950 °C to 1000 °C, and the relative density reaches 97.7 % at 1050 °C. TG-DSC tests show that the SrZP-Nd0.5Sm0.5PO4 composite ceramics exhibit excellent thermal stability at high temperatures. And SrZP-Nd0.5Sm0.5PO4 composite ceramics possess outstanding chemical stability, whose normalized leaching rate of Sr, Nd, and Sm is 10−4 g·m−2·d−1, 10−7 g·m−2·d−1, and 10−7 g·m−2·d−1 order of magnitude, respectively.

Searching for incinerating the accumulated plutonium around the world by mixing it with thorium and using this mixture as a nuclear fuel in the CANDU-6

This work investigates the possibility of getting rid of the accumulated plutonium around the world by mixing it with fertile materials and using it as a nuclear fuel for CANDU-6. Using plutonium isotopes with thorium as a nuclear fuel for CANDU-6 is a means of preventing the production of other Pu isotopes. MCNPX has been used to design a three dimensional model of the CANDU-6 bundle. Three fuel types including (238U, rgPu)O2, (232Th, rgPu)O2 and (232Th, wgPu)O2 have been examined as nuclear fuel in the designed model and their results were compared with UO2 as a standard fuel. The fuel burnup parameters such as kinf, fissile inventory ratio, plutonium concentration and minor actinides concentration have been analyzed for the suggested fuels. Some of the most related safety parameters such as effective delayed neutrons (βeff), moderator temperature coefficient, coolant temperature coefficient and Doppler constant have been studied. The excess thermal neutrons in the CANDU-6 maximized the benefit of using plutonium-based fuels, where a significant amount of plutonium has been burned during the fuel cycle. From the neutronic and safety point of view, thorium fuel mixed with reactor grade plutonium has proven to be the most promising candidate among the investigated fuels.

Effects of high valency and polarizability ion (W6+) on the phase evolution of CaZr1–2xLn2xTi2–xWxO7 zirconolite-based solid solution (Ln = Nd, Sm, Gd, Ho, Yb)

Zirconolite-based structures have served as robust matrices for the incorporation of minor actinide-rich high-level waste (HLW). The co-doping effects are critical for exploring the phase evolution of zirconolite-based structures, thereby providing a guide for HLW immobilization. Nevertheless, few studies have studied the synergistic effect of high valent ion W6+ and Lanthanides (Ln, surrogates to minor actinides) co-doping. This study proposed a systematical study on the Ln-W (Ln = Nd, Sm, Gd, Ho, Yb) co-doping in zirconolite-derived structure using X-ray diffraction (XRD) and scanning electron microscope with energy dispersive X-ray spectroscopy (SEM-EDX). The results showed that near-single-phase zirconolite-2M was recorded for samples with low level doping of Ln-W (Nd–W, Sm–W and Gd–W with x = 0.05; Ho–W and Yb–W with x = 0.05–0.1), and its structure underwent a moderate expansion along the c-axis, as revealed by a Pawley refinement method. One key finding is that near-single-phase zirconolite-2M was recorded for samples with low level doping of Ln-W (Nd–W, Sm–W and Gd–W with x = 0.05; Ho–W and Yb–W with x = 0.05–0.1), and its structure underwent a moderate expansion along the c-axis, as revealed by a Pawley refinement method. Notably, in addition to a transformation from zirconolite-2M to pyrochlore, a pseudo-yellow phase CaWO4 was also noted with increasing co-doping contents. This work demonstrates that the high charge compensator W6+ and lanthanides can be simultaneously incorporated into zirconolite structure, and a different phase transformation was revealed compared to that in low valent element doping system, which gives a new insight into the optimization of HLW immobilization route.

Multilateral evaluation of the effects of utilizing thorium oxide in the Bushehr VVER-1000 reactor

Abstract The use of thorium oxide in thermal reactors is currently being explored due to its promising outcomes. One primary concern is how to reduce the pollution of core components and nuclear waste. The inclusion of 232Th in the reactor leads to the production of 233U and decreases the formation of minor actinides. On the other hand, 238U increases the production of 239Pu, a toxic and strategically significant isotope, along with minor actinides. Therefore, the potential of thorium to reduce the production of 239Pu, one of the most harmful isotopes, can be assessed in the Bushehr VVER-1000 reactor. This study involved replacing some uranium in the Bushehr VVER-1000 reactor with 232Th. The research focused on examining the environmental impact of nuclear waste, including activity, chain reactions, and isotope levels, over a two-year period. The impact of the new fuel substitution was evaluated in the Bushehr reactor, which has a power of 3,000 MWth, in three scenarios: thorium addition to 1.6 % enriched assemblies, thorium addition to 2.4 % enriched assemblies, and thorium addition to 3.6 % enriched assemblies. These changes were analyzed in terms of nuclear waste contamination, plutonium production, fuel burn-up, and conversion ratio, and compared to the reactor using UO2 fuel. The simulation was conducted using the MCNPX 2.6.0 computational code and heterogeneous geometry. The results indicate that nuclear waste pollution decreased when thorium was added to the 1.6 % and 2.4 % enriched assemblies, but increased when added to the 3.6 % assemblies. Additionally, fuel burn-up increased with the addition of thorium in the 1.6 % and 2.4 % assemblies, but decreased in the 3.6 % assemblies. However, the conversion ratio increased in all cases. The fuel temperature coefficient (FTC), moderator temperature coefficient (MTC), and void coefficient (VC) were calculated and evaluated.

Phase transition in Eu2Zr2–xO7–2x (x = 0–1) solid solutions: A combined structural and spectroscopic study

AbstractZirconate pyrochlores have been extensively studied as candidate waste forms for the immobilization of actinide wastes due to their flexible crystal chemistry, excellent structural stability, and exceptional radiation tolerance. Herein, we report the synthesis of a series of compounds Eu2Zr2–xO7–2x (x = 0–1, step = 0.125) using a co‐precipitation route to form solid solutions between the two end members, Eu2Zr2O7 pyrochlore and Eu2ZrO5 defect fluorite. It is important to understand the pyrochlore to defect fluorite transformation as a function of Zr/Ln molar ratio. The structural investigation by x‐ray diffraction has shown the increasing cation disorder and corresponding structural transitions from the ordered pyrochlore to disordered pyrochlore, then to defect fluorite. While diffuse reflectance spectroscopy provides similar optical absorption features reflecting the presence of Eu(III) ion, Raman spectroscopy has revealed more details on the phase transition and short‐range structures. This work highlights the structural flexibility in zirconate materials as potential nuclear waste forms for the immobilization of actinide‐rich waste streams, especially suitable for the separated minor actinides.

A novel irradiation module for ANICCA fuel cycle code based on multi-task learning

This paper introduces an alternative approach to irradiation modelling within the context of nuclear fuel cycle codes like ANICCA, the fuel cycle code developed by the Belgian Nuclear Research Centre (SCK CEN). The focus lies on upgrading the irradiation module by substituting the CRAM-based depletion calculations for a more flexible and innovative approach based on Multi-Task Learning (MTL). Utilizing data generated with SERPENT2 Monte Carlo simulations, two MTL neural networks are developed to surrogate irradiation processes of Uranium Oxide (UOX) and Mixed Oxide (MOX) fuels, respectively. MTL enables simultaneous learning of the evolution of the inventories for different observables – i.e., transuranium elements, fission products, minor actinides and fertile materials – offering improved predictive capabilities compared to non-MTL neural networks, as demonstrated in the cross-validation tests.Hyperparameters of the models were found using Bayesian optimization, resulting in superior model performance compared to the old CRAM-based model. Verification against SERPENT2, used as a reference code for comparison, demonstrates the stability and accuracy of the MTL-based models, outperforming the original CRAM method for the majority of predicted isotopes.

Order-disorder structural transition and thermal expansion properties in Ce-substituted Y2Zr2O7 system

In the pursuit of advancing inert matrix fuel (IMF) applications, zirconate pyrochlores have emerged as promising candidates for the incorporation of minor actinides and plutonium. This study employs CeO2 as a nonradioactive surrogate for PuO2 and focuses on the synthesis of cerium-substituted Y2Zr2O7 samples (Y2-xCexZr2O7, 0.0 ≤ x ≤ 2.0) through a solid-state route under both reducing and oxidizing conditions to mimic plutonium incorporation. The impact of cerium content and its valence state, which in turn is influenced by synthesis conditions, on crystal structure and phase stability has been investigated through X-ray diffraction and Raman spectroscopic studies. Samples synthesized under reduced conditions undergo a defect fluorite to pyrochlore phase transition through a biphasic mixture consisting of these two phases upon increasing Ce-substitution. Conversely, the high temperature oxidizing synthesis conditions yield a defect fluorite phase field over a wide composition range (0.0 ≤ x ≤ 1.6) and a tetragonal phase at x = 2.0. Intriguingly, pyrochlore-type phases obtained under reducing conditions transform into the metastable kappa-phases upon mild oxidation at a relatively low temperature (1073 K), as confirmed by Raman spectroscopic studies. A combination of thermo-gravimetric and Raman spectroscopic investigations confirms the complete reduction of Ce4+ to Ce3+ under reducing conditions. The lattice thermal expansion behavior has also been investigated for the defect fluorite phases, synthesized under oxidizing conditions, by means of high temperature XRD spanning the temperature range 298–1273 K. Notably, the thermal expansion of compositions exhibits an increasing trend with increasing cerium content.

Methodology for Multiparameter Evaluation of Barriers Against Proliferation of Minor Actinides

There exist elements apart from uranium and plutonium that could potentially be used to construct the core of a nuclear explosive device. These belong to the so-called minor actinides (MAs), which exist in nonnegligible amounts in spent nuclear fuel (SNF) and are in nearly all cases not covered by international safeguards. Future reprocessing of SNF could result in significant separation of these elements, potentially leading to new proliferation concerns. In this work, a methodology for a transparent assessment of the barriers against proliferation of MAs has been developed and applied to the case of neptunium, americium, and curium separated from spent fuel from pressurized water reactors. In this methodology, openly available data and Monte Carlo simulations have been used to assess the barriers posed by a number of parameters relevant to the production of a nuclear explosive device from SNF. The evaluation shows that the properties of neptunium present low barriers to proliferation and that it should be discussed within the context of future nonproliferation treaties and possibly be placed under international safeguards. The properties of americium and curium present higher barriers to proliferation, meaning that these elements require less focus in the nonproliferation context.

Neutronic analysis of a tokamak hybrid blanket cooled by thorium-molten salt fuel mixture

In this study, a novel approach has been investigated by using a mixture of thorium and molten salts as a dual-purpose coolant and medium for the production of fissile fuel in a Fusion Fission Hybrid Reactor (FFHR) for the reference geometry of ITER. The study highlighted the broader benefits of thorium fuel cycling, safety features, and reduced radioactive minor actinides generation. The use of a thorium-melted salt coolant for fissile fuel production in a fusion-fission hybrid reactor represented a promising path towards efficient and sustainable nuclear energy, with potential benefits in terms of safety features and reduced generation of radioactive minor actinides. In this study, SS 316 LN-IG was selected as the first wall material for the reactor, and a molten salt fuel mixture of LiF–ThF4 was used as the coolant, taking into account the eutectic points of the material, the nominal fusion power in the FFHR for the Tokamak design concept is considered to be 500 MW. The nuclear code MCNP6 was used with the nuclear data libraries ENDF/B-VIII and CLAW-IV for the neutron calculations. The time evolution of the isotopes in the reactor was calculated with the interface code MCNPAS. The study results are evaluated in terms of tritium breeding ratio, energy multiplication factor, radiation damage, fissile fuel production and fuel burn-up value.The 4-year operation history of total TBR value is calculated and always above 1.05 and increases with time.Th initially decreased from 631.3 tonnes to 587.2 tonnes, while 233U production during this period was 9.1 tonnes. According to these results, the first wall replacement period was calculated as 3.9 years.

Structural and thermo-physical properties of sol-gel derived yttria stabilized zirconia microspheres: A viable inert matrix for minor actinide incineration

To embark upon the challenges in the field of nuclear industry including proliferation and waste management, the utilization of U-free fuel matrix (Inert Matrix Fuel) are proposed to be utilized to burn and transmute various minor actinides including Pu, Am etc. Thus in this direction, yttria stabilized zirconia (YSZ) microsphere matrix co-doped with different mol% of Ce and Nd were synthesized using internal gelation method. Different metal concentrations [M] and hexamthylnetetramine+urea (HMTA-urea) concentrations ([HMTA-urea]/[M]= [R]) were used in order to prepare crack free YSZ microspheres. The conventional sol gel method used led to the formation of anisotropic tetragonal phase due to the leaching of constituent elements. Thus a modified scheme was developed as elaborated in detail in the manuscript which resulted into an isotropic cubic phase for all compositions. The thermo-physical properties like thermal expansion, heat capacity and thermal conductivity of the samples were evaluated in the temperature range 298–1473 K and are at par with the values reported for other matrices. The thermal conductivity data showed that the conduction in these materials mainly occurs through phonon conduction and the presence of enhanced oxygen vacancies in them due to Ce and Nd doping, diminishes the thermal conductivity significantly. Further, the local structure was probed using EXAFS which confirmed them to be homogenous and coordination number 8 for all the samples. Thus with the development of such kind of fuels, the risk of proliferation can be avoided with additional advantage of energy generation.

Potential of mixed oxide fuel for higher fissile conversion ratio in light water reactors

The prospect of adopting various mixed oxide fuels to increase the conversion ratio (CR) of conventional LWRs has been explored in this study. The Westinghouse AP1000 fuel assembly (FA) was taken as the reference design. While the geometry of the FA was kept the same, five assembly models were constructed utilizing different MOX fuels in the fuel rods. Depleted UO2 was mixed with PuO2 for the first model, the next two models featured a combination of UO2 with ThO2, the fourth one utilized a mixture of PuO2 and ThO2, and the final model employed a combination of UO2, PuO2, and ThO2. Depletion analysis of these fuel assembly models was then carried out using the lattice physics code Dragon 5. The burnup-dependent infinite multiplication factor and reactivity were determined and a significant improvement in cycle length was observed at the end of the burnup period because of fertile to fissile conversion. A higher fast fission rate was obtained for plutonium-based fuel models compared to the reference, whereas the thorium-based fuel models demonstrated higher thermal fission rates. The linear reactivity model (LRM) was utilized to calculate fuel cycle performance for a three-batch refueling scheme. The three fuel models utilizing PuO2 achieved improved discharge burnup and cycle length, and required lower charge fuel mass to obtain the same thermal power. All of the five proposed models attained higher conversion ratios compared to the reference, while the highest increment in CR was obtained for the last model utilizing all three oxides. The CR of conventional UO2 fuel in AP1000 was 0.64, whereas it was calculated to be 0.73 for Model 5. Key safety parameters including the fuel and moderator temperature coefficients of reactivity and the effective delayed neutron fraction of the models were also determined. Finally, the isotopic evolution of minor actinides and fission product poisons of the models were analyzed as a function of burnup and compared with the reference.