Reactor Grade Research Articles

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.

Uniqueness, Reproducibility and Discrimination of Nuclear Warhead Gamma Signatures

Nuclear verification includes confirmation measures of nuclear weapons and fissile material, which often rely on the detection of radiation signatures. It is an open research question whether these signatures are unique. Or in reverse, could a malicious actor imitate these signatures and compromise verification of disarmament processes by using hoax objects instead of real warheads? This article presents several warhead emission models and a simulation framework to assess the uniqueness of their passive gamma radiation signatures. A red team approach includes detailed simulations of three proposed confirmation systems: The Information Barrier eXperimental, and the Information Barrier eXperimental II, both developed at Princeton University, and the Trusted Radiation Identification System, developed at the Sandia National Laboratories. For these simulations, various hoaxes are analyzed: warheads using reactor-grade plutonium instead of weapon-grade plutonium, combinations of different gamma emitting isotopes, and combinations of neutron sources with radioactive isotopes. Our findings reveal that numerous radioactive signatures are not unique and may be susceptible to imitation. Moreover, confirmation systems based on passive gamma radiation signatures using low bin numbers demonstrate a limited ability to accurately handle warheads of varying ages.

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.

Assessing the benefit of thorium fuel in a once through molten salt reactor

This study comprehensively evaluates thorium-based fuel utilisation in the Molten Salt Reactor (MSR) with a once-through fuel cycle. It examines the whether the use of thorium is beneficial in extending the fuel cycle length of an MSR, fuel utilisation efficiency, reactor safety, and reduction of long-lived radioactive waste. To observe those parameters, this research compares the use of thorium and uranium as fertile fuel in a single-fluid, once through MSR using Uranium-233 (U-233), Uranium-235 (U-235), and Reactor Grade Plutonium (RGPu) as well as the fissile driver. MCNP radiation transport code with ENDF/B-VII.0 continuous neutron library was used to analyse the neutronic parameters and fuel burnup in a multi-refuelling scheme for a total burnup time of 8 years. The findings indicate that the utilisation of thorium in conjunction with U-233 and U-235 significantly enhances fuel consumption and improve reactor safety. Such benefits, however, did not appear in RGPu. Moreover, thorium effectively reduces the production of long-lived radioactive waste, a crucial issue in nuclear waste management. Considering technical and environmental aspects, this study offers novel insights into using thorium as a more sustainable nuclear alternative under a once through fuel cycle, within a condition of multi-refuelling cycle which is only possible in an MSR.

Multivariate test applied to single Pu isotopic composition measurement

The determination of the plutonium isotopic composition by gamma spectrometry is one of the tasks performed by nuclear inspectors. It is typically measured using planar germanium detectors. The spectra are analysed with deconvolution codes, which calculate all isotopic abundances. The experimental results are then compared with the declared values and the acceptance criterion is often based on one unique isotope, 240Pu, to distinguish between reactor and weapon grade Pu.If the measurement is a part of a measurement series, it is possible to calculate covariance terms and carry out a more comprehensive comparison using Hotelling's test as recently presented in Berndt and Mortreau (2023) [1]. In this study, it was shown that the T2 test often rejects simulated false cases which are accepted if only the 240Pu abundance is considered. Hence, the T2 test should be preferred to the 240Pu criterion.In case of a single measurement, which is typical for inspections, the co-variance cannot be determined and the comprehensive calculation of the T2 test as performed in Berndt and Mortreau (2023) [1] cannot be applied. The purpose of the present paper is to consider approximations of the variance-covariance matrix to allow the application of the T2 test to isolated measurements. The results were then applied to more than 400 single measurements of Pu samples covering many different experimental conditions.

Investigating the possibility of using a mixture of thorium with different fissile materials as a fuel in TRISO particles for the PBMR-400 reactor

The increasing demand for nuclear reactors as a source of clean energy forced nuclear researchers to search for ways to enhance the safety operation of nuclear reactors. This paper discusses the viability of using different thorium compounds ((233U, Th)O2, (235U, Th)O2, (RGPu, Th)O2) as an accident tolerant fuel/advanced technology fuel (ATF) for Pebble Bed Modular Reactor 400 (PBMR-400). A comprehensive analysis of thorium-based fuel performance was carried out and compared with UO2 to develop a novel nuclear fuel with the capability to tolerate severe accident conditions. The neutronic characteristics such as the infinity multiplication factor, initial heavy metal concentration, RGPu concentration and fission products concentration were analyzed. In addition to, the main safety parameters have been calculated to distinguish the effect of the proposed fuels on them. The thermal spectrum and thermal output power have been visualized to give a good insight view inside the core about the distribution of the heat. In general, the neutronic analysis demonstrated good behavior of the examined thorium-based fuel in the PBMR-400 and especially in the case of (RGPu, Th)O2. PBMR-400 showed excellent performance in incinerating a large amount of reactor-grade plutonium when it was used in a kernel fuel bed.

Zirconium sponge production: an integrated approach for chemical characterization of process intermediates using ICP-OES

Abstract The present paper discusses about the method developed for complete chemical characterization of Zirconium Washed and Dried Frit (WDF) and Reactor Grade ZrO2 powder which includes nonmetallic element sulphur using Inductively Coupled Plasma Optical Emission Spectrometer (ICP-OES). Studies have been carried-out on several important aspects of method development such as effect of matrix on determination of analytes, removal of matrix using different solvents, selection of interference free & sensitive wavelengths, calibration strategies like internal standard & standard addition, optimization of several instrumental parameters which includes RF power, plasma gas flow, nebulizer gas flow, nitrogen purging etc., and are discussed. A % RSD of less than 1 % for Zr and up to 3 % for other elements has been achieved in this method. The developed method has been validated using standard recovery of spiked real time samples with known amount of reference materials. Integrated approach adopted in the development of this method has resulted in reduction of analytical waste generated and also enabled to give quick analytical feedback to production plant for downstream processing.

Analysis of variation minor actinide pin configurations Np-237, AM-241, and Cm-244 in UN-PuN fueled pressurized water reactor

Actinide minor is a reactor waste with high toxicity and a long half-life. Minor actinides can be reduced by reusing them as fuel mixtures in reactors. This research uses PWR reactors with the primary fuel UN-PuN or Uranium Plutonium Nitride with a burning time of 5 years. The fuel consists of enriched Uranium, reactor-grade Plutonium from LWR waste, and minor actinides including Neptunium-237, Americium-241, and Curium-244. The purpose of this study was to find a design that is effective in reducing minor actinide waste. There are six designs or cases used in the addition of minor actinides. Each case has six minor actinide pins in each assembly. The addition of minor actinides is arranged in heterogeneous cores. The analysis was carried out by observing the values of k-eff, excess reactivity, and mass of minor actinides obtained from simulations using OpenMC code 0.13.2 and the ENDF/B-VIII library. The homogeneous core obtained an excess reactivity of 9.7 % with a percentage of plutonium of 8 %. The results of the homogeneous core are used as a reference for preparing a heterogeneous core. The heterogeneous core obtained an excess reactivity of 9.9 % with a percentage of plutonium F1: 5.5 %, F2: 8 %, and F3: 10.5 %. Np-237 can be reduced by 53 kg, and Am-241 can be reduced by 61 kg with minor actinide pins in case 1. Cm-244 can be reduced by 363 kilograms with minor actinide pins in case 6. Excess reactivity in the addition of Np-237 and Am-241 decreased to 5.3 %, while the accumulation of Cm-244 increased to 12.1 %.

Feasibility of Using Mixed-Oxide Fuel for a Pressurized Water Reactor Instead of Traditional UO2 Fuel Material

This study examines the feasibility of utilizing mixed-oxide fuel [(U0.9, rgPu0.1) O2] instead of traditional UO2 in nuclear reactors. The utilization of (U0.9, rgPu0.1) O2 is particularly significant as it represents an effective approach to nuclear fuel recycling by combining reactor-grade plutonium extracted from partially used nuclear fuel and depleted uranium obtained through the enrichment process. The fundamental neutronic characteristics, such as the radial power distribution, were investigated using the MCNPX 2.7 algorithm to identify the specific channel for subsequent thermal-hydraulic (TH) analysis. The TH analysis was conducted using COMSOL-Multiphysics, allowing for the estimation of the fuel rod’s axial and radial temperature profiles, as well as the determination of the departure from the nucleate boiling ratio. Furthermore, the coupling between heat transfer and solid structure (SS) was achieved using the Multiphysics tool in COMSOL-Multiphysics. This coupling facilitated the simulation of key SS parameters, including von Mises stress, volumetric strain, and displacement, while considering the influence of heat transfer. The results demonstrate significant improvements and enhanced safety margins when utilizing (U0.9, rgPu0.1) O2.

Neutronic performance evaluation of Plutonium Recycling in two core sizes for a 250 MWt Molten salt reactor

MSR has a unique feature in the fuel system, using liquid salt as fuel, and has flexibility on material utilization in the fuel, such as Plutonium. Recycling Plutonium into fuel is a good option to increase the non-proliferation and fuel sustainability aspects. This paper investigates the effect of different reactor core sizes and different Plutonium grades in neutronic performances on 250 MWt MSR. The calculation is performed using PIJ and CITATION modules in the SRAC program and JENDL 4.0 as a nuclear data library. The reactor core size is varied into two different sizes (type-A and type-B), where the type-A reactor has double the size of type-B. The variation of Plutonium grades is employed in this study for several fuel cases. The fraction of Plutonium in the salt (PuF4) is optimized based on the obtained reactor criticality for 2000 days of operation time. The results indicate that reactor type-B has a higher PuF4 consumption, and the percentage of plutonium reduction becomes higher than reactor type-A. Additionally, the comparison of neutronic performance on the varied fuel cases is also analyzed in this paper.

Evaluation of neutronic performance for the VVER-1000 reactor core with regenerated uranium-plutonium fuel

The possibility has been considered for the VVER-1000 reactor fuel loading to be formed based on regenerated fuel with the use of the spent fuel accumulated in reactors of the same type. A study was undertaken to investigate the change in the isotopic composition of the plutonium discharged from a thermal reactor in the course of its multiple reprocessing and recycle in a thermal neutron reactor. To obtain an equilibrium isotopic composition of the reactor-grade plutonium, 3D neutronic calculations were performed for the stationary fuel cycles of a VVER-1000 serial reactor with conventional oxide fuel and oxide fuel based on regenerated uranium and based on an undivided mixture of uranium and plutonium oxides from SNF. The neutronic performance of reactor cores was compared for the above mentioned fuel types in the course of the fuel company, including the following: in-core radial power density shaping, values of reactivity coefficients for various thermal parameters, reactivity control system efficiency, etc.

A novel approach for managing the excess reactivity at the beginning of the fuel cycle of VVER-1200

This work discusses various fuel pellet designs loaded with different types and concentrations of burnable absorbers to suppress the excess reactivity at the early stage of the fuel life in the VVER-1200 fuel assembly. MCNPX, V. 2.7 with the CINDER 90 fuel depletion code was utilized to analyze the neutronic characteristics of the proposed cases. Concentric shells (CS) of integral burnable absorbers are investigated in the fuel tube. Gadolinium and Erbium at different concentrations are investigated in the CS at different operation cycles to manage the excess reactivity. Gadolinium and boron as traditional burnable absorbers (BAs) are investigated in the standard fuel assembly to compare their results with the suggested cases. The infinity multiplication factor (kinf), the concentrations of the fuel composition, reactor-grade plutonium (rgPu), minor actinides (MAs) and some important fission products poisoned were studied with burnup. The radial power and thermal neutron flux distribution through one-six of the VVER-1200 assembly have been analyzed to distinguish the optimum BA and design for flattening the power profile. The neutronic analysis indicates that using Er2O3 in a concentric shell of a fuel pellet is an interesting method for reactivity management.

Behaviors of actinides in chromatographic separation by using TBP resin in nitric acid solution and hydrochloric acid solution

ABSTRACT We are developing a simple actinide separation method for a pre-treatment process of actinide analysis by mass spectrometry such as thermal ionization mass spectrometry or inductively coupled plasma mass spectrometry. The actinide separation was conducted by using Tributyl phosphate (TBP) impregnated resin commercially supplied by Triskem International. TBP is commonly used in the PUREX system as a solvent for the recovery and recycling of uranium and plutonium from spent nuclear fuels through liquid–liquid separation in the HNO3 system. In this study, TBP-impregnated resin is employed as a solid extractant for the mutual separation of uranium, plutonium, thorium, and trivalent actinides through solid–liquid separation by combination with nitric acid (HNO3) and hydrochloride acid (HCl), with divalent iron ion as a reductant for valence control of plutonium. We confirmed that uranium was extracted with a recovery rate of more than 96%.

Use of remix spent mixed fuel plutonium in the BN-1200 reactor

The VVER-1000 thermal neutron reactor can operate on mixed uranium-plutonium fuel with a content of reactor-grade plutonium up to 5% with a 100% loaded core. In this case, plutonium burns up to 56% of odd isotopes. The energy potential of such plutonium is very low, and its further use in thermal reactors is impractical. However, such plutonium can be used in fast neutron reactors. The paper presents the results of investigating the possibility for such isotopic plutonium composition to be used in the BN-1200 thermal neutron reactor and its value be increased for the plutonium recycle in the reactor. For this purpose, a precision model of the BN-1200 reactor has been developed using the Serpent Monte Carlo code. The model has been verified against the reference values of the nuclear fuel burnup and breeding ratios. The study has shown that such plutonium can be used in the BN-1200 reactor MOX fuel. Maintaining the operating cycle length requires the plutonium fraction in the MOX fuel to be increased up to 2%. In the BN-1200 reactor, the isotopic composition has been found to improve for the further recycle of plutonium in the thermal reactor, i.e. odd plutonium isotopes increase. The fewer odd plutonium isotopes at the beginning of the BN-1200 operating cycle, the greater their increase. It can be seen as the result of the calculation that plutonium from VVER-1000 spent mixed fuel must be loaded into the BN-1200 reactor at least twice to increase the fraction of odd isotopes to the level of reactor-grade plutonium.

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.

Influence of high temperature GPC/SEC detection modes on the accurate molar mass estimation of slurry phase multimodal HDPE reactor materials

In the present study, trimodal high-density polyethylene (tri-HDPE) slurry phase samples from the polymerization reactors were subjected to high temperature gel permeation chromatography/size exclusion chromatography (HT-GPC/SEC) analysis. The molecular weight (Mw) and its distributions were evaluated using commercially available HT-GPC instruments equipped with different detection modes (DRI, IR5 MCT, and light scattering). The SEC studies reveal the fact that there is a considerable challenge in generating a repeatable and reproducible data for the reactor grade HDPE samples which could be attributed to different detection modes employed and instability of HDPE reactor grades which are susceptible to undergo thermo-oxidative degradation during HT-GPC analysis. Among various instrumentation and detection modes investigated, the HT-GPC with IR5 MCT detection showed reliable results (%Error <15) when compared with absolute Mw data obtained from MALS detection. Based on the findings, proposed suitable SEC mode to characterize the HDPE reactor grade polymer samples.

Experimental Test of a Process Upset in the EURO-GANEX Process and Spectroscopic Study of the Product

ABSTRACT EURO-GANEX is an innovative hydrometallurgical process for the group separation of transuranic actinides from future spent nuclear fuels. A flowsheet test of a process upset (or ‘maloperation’) has been carried out involving the reduction in the concentration of the scrub acid feed to the extract-scrub contactors. The experimental test was completed in two stages. First, the extract-scrub contactors were run under normal flowsheet conditions before initiating the process upset by reducing the scrub acidity from 0.5 to 0.05 mol/L HNO3. At low acidity, a change in the online UV-Vis spectra of the solvent phase indicated the formation of hydrolyzed plutonium species or plutonium colloid formation. Surprisingly, however, this did not lead to the recycle and accumulation of plutonium in the extract-scrub contactors as expected indicating that the EURO-GANEX process is quite robust towards this process upset. Further spectroscopic investigations of the organic-phase species were also performed to characterize the conditions under which the hydrolyzed plutonium species form in the EURO-GANEX solvent as well as separate TODGA and DMDOHEMA phases. The spectroscopic studies supported the view that Pu(IV) hydrolyses at low acidity but the hydrolysis is limited by the organic ligands and is consequently reversible on raising the acidity again. The hydrolysis was also observed in separate TODGA and DMDOHEMA phases.

Recycling alternatives for minor actinide and plutonium in a boiling water reactor

Recycling is an alternative for reducing minor actinide and plutonium stockpiles from depleted fuel assemblies. In the current study, it is proposed the use of a MOX fuel assembly called MOXIM-MA. It comprises 4 rods with minor actinides, 14 UO2-gadolinium rods, 8 water rods, and 66 MOX fuel rods. This MOXIM-MA design is used in a 2027 MWth BWR, reducing 0.130 kg HM of minor actinide and 5.042 kg HM of Plutonium per fuel assembly at discharge after discharge 4-operating cycles of 18 months each. The reactor core energy produced does not suffer any penalty by using this fuel assembly. There is no need for additional changes to the original control safety system. The radiotoxicity results show almost one magnitude order increase compared with standard UO2 fuel assemblies for the same BWR reactor core. Still, this increase is the same magnitude experienced when MOX fuel assemblies without minor actinide are used. A twin reactor power plant, 2027 MWth per unit, with 20 and 25 years of operational life, is considered to propose a plutonium and minor actinide recycling strategy. Standard UO2 fuel assemblies, at discharge for this plant, produce 0.135 kg HM of MA and 1.707 kg HM of plutonium. By recycling them, at least 6 depleted fuel assemblies are needed for each MOXIM-MA fuel assembly processed. The suggested strategy starts by considering the cumulative amount of minor actinide and plutonium coming from the depleted fuel assemblies discharged during the plant's operating life at the moment of this analysis. The study shows that the MOXIM-MA fuel assemblies will be used in only one of the reactor units in a once-through scheme. It describes how the auto recycling of plutonium and MA could be done to supply the MOXIM-MA fuel assemblies during the rest of the plant's operation. Results promise to deliver the plant's whole operation life a decrease of 39.50% in plutonium inventories and 12.70% in MA inventories, achieving this goal without penalizing energy production.

Design of an electron cyclotron emission diagnostics suite for COMPASS Upgrade tokamak.

COMPASS Upgrade is a medium size and high field tokamak that is capable of addressing key challenges for reactor grade tokamaks, including power exhaust and advanced confinement scenarios. Electron cyclotron emission will be available among the first diagnostics to provide measurements of high spatial and temporal resolution of electron temperature profiles and electron temperature fluctuation profiles through a radial view. A separate oblique view at 12° from normal will be utilized to study non-thermal electrons. Both the radial and oblique views are envisioned to be located in a wide-angle midplane port, which has dimensions that enable simultaneous hosting of the front-end of their quasi-optical (QO) designs. Each QO design will have an in situ hot calibration source in the front-end to provide standalone and calibrated Te (R,t) measurements. The conceptual design for each QO system, the Gaussian beam analysis, and the details of the diagnostic channels are presented.

Temporal variability of plutonium in surface waters of the Sea of Japan

Long-term temporal variations of plutonium in Sea of Japan (SOJ) surface waters have been examined with the aim to better understand its behavior during several decades. The first observation is that 239,240Pu activity concentrations in surface waters of the SOJ during 1977–2019 were 6.5 ± 4.7 mBq m−3 in average, and 5.1 mBq m−3 as the median, whereas 137Cs and 90Sr activity concentrations decreased with time, except of the perturbation due to the 2011 Fukushima Dai-ichi Nuclear Power Plant (FDNPP) accident. Another observation is that sporadic high 239,240Pu activity concentrations occurred in the east Japan Basin, ranging from 1 to 39 mBq m−3. The spatial distribution of 239,240Pu activity concentrations in surface waters revealed that high 239,240Pu levels (>20 mBq m−3) occurred in 1994 in the northern SOJ, which was considered to be due to winter convection. To elucidate factors controlling the temporal variability of surface 239,240Pu levels in the SOJ, a relationship between surface 239,240Pu activity concentrations and vertical diffusion coefficients was examined. The results revealed that this relationship could be classified into two groups: one group did not show a change with increasing diffusion coefficient, while the other group showed a positive correlation. The vertical 239,240Pu distribution in SOJ waters suggests that the high surface 239,240Pu levels occurred due to the upwelling of cyclonic eddy. The rapid recycling of deeper plutonium occurred in the SOJ due to deep winter convection and upwelling associated with cyclonic eddy. The plutonium levels in the SOJ have been found to be sensitive to climate changes. Warming of the SOJ may cause a reduction of winter convection and eddy activity as a result of increasing sea surface temperature. This leads to a decline of recirculation of plutonium and other bioavailable elements from Japan Sea Proper Water (JSPW) to surface water layers. Plutonium would be, therefore, an important indicator of biogeochemical processes in the marine environment, helping to assess climate change impacts on marine ecological systems.