This study examines the implications of tripling global nuclear capacity by 2050 on the nuclear fuel cycle, based on national projections and COP28 climate commitments. Regionally disaggregated electricity scenarios were generated and used as inputs for the ANICCA simulation code, applying Monte Carlo methods to assess uncertainty in fuel cycle metrics. Three strategies were analyzed: open cycle, partially closed cycle (Pu mono-recycling in LWRs), and advanced closed cycle (Pu and MA multi-recycling in LFRs). Results show that the open cycle could require about 15 million tons of natural uranium by 2100, surpassing identified reserves. Pu mono-recycling reduces uranium and enrichment needs by similar to 9% and achieves Pu balance post-2050. The advanced cycle cuts minor actinide accumulation by similar to 50%, easing long-term repository burdens. These results highlight the need to explore advanced fuel cycles and expand infrastructure for reprocessing, MOX fabrication, and waste management to meet sustainability goals under high nuclear deployment scenarios.

This paper presents the work performed in the subgroup 16 of the Working Party for Nuclear Criticality Safety (WPNCS) of the OECD Nuclear Energy Agency. The main goal was to define two decay heat benchmarks for Spent Nuclear Fuel (one pincell and one assembly), perform calculations and compare and analyze the results in light of existing calorimetric measurements. The selected case is the PWR UO2 assembly 0E2, irradiated at the Ringhals-3 reactor and measured at the Clab facility in Sweden. In total, 21 institutes worldwide participated to the exercise, leading to 55 calculated results (named C). It was found that the measured decay heat values (E) can be satisfactorily reproduced with two-dimensional assembly calculations, leading to an average C/E value of 0.99, with an uncertainty (or one standard deviation) of ±0.01.

Nuclear data uncertainties taken from the general-purpose evaluated libraries JEFF-3.3, ENDF/B-VIII.0 and JENDL-4.0u are propagated through a depletion model of the ARIANE GU3 sample using the SANDY stochastic sampling code combined with the Monte Carlo burnup code SERPENT-2. This approach enabled an accurate characterization of the uncertainty in many nuclide concentrations, for which measurements exist from post-irradiation experiments. Stochastic sampling methods for uncertainty propagation in Monte Carlo burnup calculations are notoriously computationally expensive. To address this, the contribution of nuclear data uncertainties to the model response was assessed independently of Monte Carlo uncertainties using a methodology based on conditional estimators. Interestingly, unlike best-estimate values, uncertainty estimates were found to be rather independent of model simplifications. This was demonstrated by comparing uncertainty results for the GU3 fuel assembly model and for a simplified pincell model. The possibility to transpose uncertainties between such models suggests that high assay data accuracy is not strictly necessary for uncertainty analyses. Finally, the variance decomposition analysis revealed gaps in the uncertainty datasets of major nuclear data libraries, leading to an underestimation of total uncertainties in burnup calculations.

Lead-cooled Fast Reactors (LFRs) with mixed-oxide (MOX) fuel are promising candidates in the Generation IV (Gen IV) small modular reactor (SMR) landscape due to their capacity for actinides transmutation, passive safety features, and minimized waste radiotoxicity. For secure management, storage, safeguards, rigorous characterization is necessary. This database was developed to support the optioneering and design of MOX-based lead-cooled fast reactors. This data article introduces a comprehensive dataset of isotopic mass densities, spanning 152 nuclides present in irradiated LFR-MOX fuel, additionally providing insights into fuel characteristics such as activity, decay heat rates, photon emission rates, spontaneous fission rates, and radiotoxicity values across various decay steps.Using the Serpent2 Monte Carlo code for fuel depletion calculations, and processed with SerpentTools, the dataset captures inventory data as a function of reactor power, fuel burnup, plutonium vector in the fresh MOX, and decay time at the end of irradiation, enabling analyses of SNF properties. The dataset is stored in Parquet format, including one primary depletion file and 13 decay files.

Monte Carlo sampling is frequently employed for uncertainty quantification in depletion calculations. Several assumptions are needed to perform this analysis. In this work, an assessment of these assumptions is proposed via sample convergence studies and perturbation of the sampling distribution. The Uncertainty Analysis in Best-Estimate Modeling (UAM) Pincell Hot Full Power and the Turkey Point reference cases were considered for this purpose. The U-235 thermal independent fission yield uncertainties evaluated in JEFF-3.3 and JEFF-4.0 were propagated to the nuclide vector and to the system multiplication factor. Using JEFF-4.0 data, a 75% reduction in the uncertainty of selected nuclide concentrations and an 80% reduction in the multiplication factor uncertainty were observed, showcasing the effect of full covariance evaluations. The presented results also prove that the uncertainty in the considered observables shows marginal dependence on the sampling distribution.

The Rod-Extremity and Gadolinia AnaLysis (REGAL) Program is a joint international effort to expand the nuclide inventory experimental data for irradiated nuclear fuel, with a specific focus on addressing two challenging needs associated with the characterization of modern, high duty, nuclear fuel. The first challenge is filling the gaps in experimental nuclide inventory data for gadolinia (UO2–Gd2O3) fuel rods. The huge absorption cross sections of Gd-155 and Gd-157 in the Gd dopant in these rods lead to atypical spatial self-shielding patterns and have an impact on the neutronic environment within the fuel assembly compared to regular UO2 fuel rods. The second challenge is investigating the impact of burnup gradients at rod extremities on fuel composition and neutron leakage, to provide relevant experimental data for assessing computational capabilities to model such impact. A benchmark has been defined as a first step in the development of best-estimate models in the preliminary phase of the experimental data evaluation. Comparison of experimental results obtained in Phase I of the program for two measured pressurized water reactor (PWR) samples, one UO2 and one UO2–Gd2O3 sample, with calculated results obtained with different computational tools based on the defined benchmark are presented and discussed.

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.

Within the frame of the EU H2020 program SANDA project, sensitivity and uncertainty analyses have been performed for the ESFR, ASTRID and ALFRED reactor concepts and the multi-purpose flexible irradiation facility MYRRHA. Relevant reactor parameters, namely the effective multiplication factor, the effective delayed neutron fraction, the Doppler reactivity coefficient, the void worth and the worth of control rods have been investigated. SCALE, Serpent 2 and MCNP6.2+SUMMON codes together with the covariance data of the JEFF-3.3 nuclear data library have been used for this study. A ranking of the most important isotopes and reactions impacting the mentioned parameters has been derived for all these systems. Uncertainties have been quantified and have been found to be higher than the target accuracies proposed. Therefore, recommendations for improvement of nuclear data are given.

The neutronics design of new reactors has been historically relying on zero power research reactors for the simulation of the core irradiation conditions as well as for nuclear data and code validation. In this framework, a study on the possible MYRRHA representativity improvement coming from the loading of MOX fuel in VENUS-F zero power reactor is under investigation. This work analyses the effect of the nuclear data used in the sensitivity and representativity calculations, highlighting the need for a comprehensive set of nuclides and reactions to be further studied.

This paper summarized the efforts performed to understand decay heat estimation from existing spent nuclear fuel (SNF), under the auspices of the Working Party on Nuclear Criticality Safety (WPNCS) of the OECD Nuclear Energy Agency. Needs for precise estimations are related to safety, cost, and optimization of SNF handling, storage, and repository. The physical origins of decay heat (a more correct denomination would be decay power) are then introduced, to identify its main contributors (fission products and actinides) and time-dependent evolution. Due to limited absolute prediction capabilities, experimental information is crucial; measurement facilities and methods are then presented, highlighting both their relevance and our need for maintaining the unique current full-scale facility and developing new ones. The third part of this report is dedicated to the computational aspect of the decay heat estimation: calculation methods, codes, and validation. Different approaches and implementations currently exist for these three aspects, directly impacting our capabilities to predict decay heat and to inform decision-makers. Finally, recommendations from the expert community are proposed, potentially guiding future experimental and computational developments. One of the most important outcomes of this work is the consensus among participants on the need to reduce biases and uncertainties for the estimated SNF decay heat. If it is agreed that uncertainties (being one standard deviation) are on average small (less than a few percent), they still substantially impact various applications when one needs to consider up to three standard deviations, thus covering more than 95% of cases. The second main finding is the need of new decay heat measurements and validation for cases corresponding to more modern fuel characteristics: higher initial enrichment, higher average burnup, as well as shorter and longer cooling time. Similar needs exist for fuel types without public experimental data, such as MOX, VVER, or CANDU fuels. A third outcome is related to SNF assemblies for which no direct validation can be performed, representing the vast majority of cases (due to the large number of SNF assemblies currently stored, or too short or too long cooling periods of interest). A few solutions are possible, depending on the application. For the final repository, systematic measurements of quantities related to decay heat can be performed, such as neutron or gamma emission. This would provide indications of the SNF decay heat at the time of encapsulation. For other applications (short- or long-term cooling), the community would benefit from applying consistent and accepted recommendations on calculation methods, for both decay heat and uncertainties. This would improve the understanding of the results and make comparisons easier.

MYRRHA is a flexible experimental facility being designed at the SCK CEN, in Mol, Belgium. Cooled by lead-bismuth, it is conceived to operate both in sub-critical mode, as an accelerator driven system, and in critical mode, as a fast reactor. In order to comply with MYRRHA reactor design requirements, uncertainties due to nuclear data must be quantified. Significant gaps between the uncertainties and the target accuracies have been systematically shown in the past. In this paper, first, a Sensitivity and Uncertainty analysis with JEFF-3.3 nuclear data library of the effective neutron multiplication factor k eff of the latest MYRRHA reactor design - v1.8 - is presented. Then, since target accuracy for k eff of 300 pcm is exceeded, a Target Accuracy Requirement assessment is performed in order to find out the required accuracy on cross section data to meet the requested target accuracy. To reach the requested target accuracy, a reduction of the uncertainty in the fission and capture cross sections of 240 Pu JEFF-3.3 evaluation is needed.

This database contains the isotopic mass density and the contribution to activity, decay heat, photon emission, spontaneous fission rate, (α,n) emission rates and radiotoxicity of 150 nuclides that are present in nuclear fuel irradiated in PWRs. These nuclides are of paramount importance for nuclear waste characterization and fuel cycle analysis. These values were obtained by depletion calculations based on a 3D pin-cell geometry model and performed with the Monte Carlo reactor physics burnup calculation code Serpent2, with state-of-the-art nuclear data libraries and relevant methods. The calculations cover a wide range of burnup levels for conventional PWRs and take into account both UOX and MOX fuel. A broad span for initial enrichment for UOX (from 1.5% to 6.0%), and for both the initial plutonium content (from 4.0% to 12.0% and the plutonium isotopic composition of MOX has been considered. This database has been made publicly available due to its relevance in the fields of waste and fuel characterization, nuclear safeguards and radiation protection, and it will allow other potential users to avoid the time-consuming calculations required to obtain the aforementioned data. Additionally, it constitutes an interesting dataset for model training in machine learning applications related to nuclear science and engineering.

A method to determine the neutron production rate of a spent nuclear fuel segment sample by means of non-destructive assay conducted under standard controlled-area conditions is described and demonstrated. A neutron well counter designed for routine nuclear safeguards applications is applied. The method relies on a transfer procedure that is adapted to the hot cell facilities at the Laboratory for High and Medium level Activity of SCK CEN in Belgium. Experiments with 252 Cf(sf) sources, certified for their neutron emission rate, were carried out at the Joint Research Centre to determine the characteristics of the detection device. Measurements of a segment of a spent nuclear fuel rod were carried out at SCK CEN resulting in an absolute and non-destructive measurement of the neutron production rate avoiding any reference to a representative spent nuclear fuel sample to calibrate the device. Results of these measurements were used to study the performance of depletion codes, i.e., ALEPH2, SCALE, and Serpent2. The study includes a code-to-code and code-to-experiment comparison using different nuclear data libraries.

In this work, a study dedicated to the characterization of the neutronics aspect of the Spent Nuclear Fuel (SNF), as part of the European project EURAD (Work Package 8), is presented. Both measured nuclide concentrations from Post Irradiation Examination samples and decay heat from calorimetric measurements are compared to simulations performed by different partners of the project. Based on these detailed studies and data from the published literature, recommendations are proposed with respect to best practices for SNF modelling, as well as biases and uncertainties for a number of important nuclides and the SNF decay heat for a cooling period from 1 to 1000 years. Finally, specific needs are presented for the improvement of current code prediction capabilities.

Nuclear data uncertainty analysis on the spent nuclear fuel inventory was performed on the Takahama-3 NT3G23 assembly, where the sample SF95-4 was irradiated up to a burnup of approximately 36 GWd/ t according to the SFCOMPO benchmark. The cross-section covariance matrices stored in the ENDF/B-VIII.0, JEFF-3.3 and JENDL-4.0u evaluated nuclear data libraries were propagated with the stochastic sampling algorithms implemented in the SANDY code. A comparison of the concentration uncertainty differences obtained using data from the three libraries is reported. Similarities were found with the fuel composition uncertainty results obtained for the Calvert Cliffs MKP109 sample P SFCOMPO benchmark. Such a similarity was also found when comparing concentration uncertainties along the sample irradiation. Therefore, the main contributors to the concentration uncertainty of a number of nuclides were identified at different burnup levels in the two samples. To complement the similarity analysis, a correlation study of the concentration distributions predicted by the two models was performed. The reported results hint a dominance of the common uncertainty propagation mechanisms over the model differences in the determination of concentration uncertainty.

Production of 238 Pu proceeds via neutron irradiation of 237 Np, which is created as a by-product in nuclear fission reactors, with a typical production slightly less than 1000 g/tU. When reprocessing spent nuclear fuel as e.g. done in the ORANO plant at La Hague (France) via the plutonium uranium reduction extraction (PUREX) process, neptunium is partially co extracted with uranium from dissolved irradiated nuclear fuel but as of today, it is not further refined, but instead added to the high-level waste and vitrified. The PUREX process can in principle be modified for neptunium recovery, and reprocessing of civil spent fuel can thus provide an abundant source of 237 Np. Neutron irradiation of separated 237Np to produce 238pU is conceptually simple, but producing sizeable quantities of 238 Pu with acceptable isotopic purity, separating it from the host matrix in which it is generated, its further processing and encapsulation, poses formidable technological challenges. All plutonium isotopes of technological interest are extremely radiotoxic alpha emitters. The elevated specific activity of isotope 238 adds highly concentrated radiolysis issues in liquid phase processing, static charging problems during powder handling, and heat generation when forming solid samples to the normal challenges of handling highly radiotoxic materials. Also, the precursor material 237Np is a radiotoxic alpha emitter with a lesser specific activity compared to 238 Pu. Especially the very rich and weakly explored chemistry of Np is a challenge here. In the present contribution, options for the irradiation of 237Np containing transmutation targets in the BR-2 high flux reactor of SCK CEN (Mol, Belgium) are presented and boundary conditions for the production of such targets are discussed. The principal technology steps are: •Fabrication of 237Np targets for neutron irradiation •Production of 238 Pu by neutron irradiation of 237Np targets •Processing of irradiated targets and Pu/Np separation •Conversion of separated 238 Pu to solid PuO 2 pellets Several technology options exist in each of the steps listed above, and these are reviewed with specific attention paid to those options which have been brought to actual production stage in the past, or which are intensively pursued today. Two principal options stand out: mixed ceramic-metal (CERMET) based routes developed and implemented at the Savannah River Site between the early 1950's and late 1980's and ceramic based routes pursued a.o. by Oak Ridge National Laboratory that have recently gained more attention. Technology options chosen in the early days are of course not necessarily the ones which would today be preferred. The European nuclear technology also developed differently over the past decades than the US nuclear technology, particularly with respect to civil spent fuel reprocessing, Pu separation and (U,Pu)O 2 manufacturing for Light Water Reactor (LWR) application. The two front-end processes (CERMET and full-oxide) for Np target production have similarities with established industry-scale fuel manufacturing processes in Europe: CERMET processes are applied in Materials Test Reactor (MTR) fuel fabrication and full-oxide processes are the reference technology for power reactors. The full-oxide process is furthermore also implemented at industrial scale for mixed uranium-plutonium (MOX) oxide fuel manufacturing, which shares similar radiotoxicity concerns as for Np-targets. CERMET front-end processes are in Europe applied for uranium-based fuels. An assessment of 238 Pu production capabilities in the BR-2 reactor has shown that suitable core positions can be selected with sufficiently low by-production of unwanted 236 Pu. Further assessment will be needed to evaluate which is the flexibility regarding core positions that guarantee limited production of 236 Pu. Production rates of 238 Pu were evaluated for unperturbed flux conditions to deduce an upper boundary of possible production rates. Preliminary calculations under perturbed flux conditions were performed for a design option that yet has to be optimized to deduce a lower boundary. The unperturbed flux results showed a theoretical upper boundary for the transformation yield slightly below 20%, achieved after three cycles of 28 days each and 28 days downtime between each cycle. Two production campaigns (i.e. six cycles) can reasonably be foreseen per year. Prolonged irradiation reduces the Pu vector below acceptable quality. The perturbed flux calculations for an un-optimized target showed a lower boundary slightly above 5% under the same irradiation conditions. Actual production yields are expected to be closer to the lower boundary than the upper boundary. Assuming a loading of 3 kg of Np, one may thus expect 150 g 238 Pu for a single production campaign, or 300 g 238 Pu per year. Design optimizations are expected to improve this yield. Regarding the processing of irradiated targets, the principal concerns are waste-related. The dissolution stage for the NpO 2 Al CERMET targets is particularly problematic and from this perspective full-ceramic NpO 2 targets would be preferred. The anion exchange purification stage as historically applied at SRS could be replaced by solvent extraction with TBP, most probably in combination with additional purification of the products via ion exchange. It is believed that the TBP process will create less waste than an anion exchange process and therefore seems to be the preferred method. Experience with these processes is available in Europe. The decay of 238 Pu dominates the radiology of the plutonium generated from 237Np irradiation. Decay parameters of a typical Pu-vector issued by 237Np irradiation have been compared with those of LWR MOX for which much broader experience exists in Europe. Even compared to MOX, issued from high burnup UO2 and including the ingrowth of 241 Am equivalent to a time lapse of 2 1/2 years, the radiological parameters per unit mass of Pu issued from 237 Np irradiation are higher by about two orders of magnitude. This difference will have to be taken into account for the Pu conversion process and the PuO 2 production process.

In the framework of the Horizon 2020 project ESFR-SMART (2017-2021), the European sodium fast reactor (ESFR) core was updated through a safety-related modification and optimization of the core design from the earlier FP7 CP-ESFR project (2009-2013). This study is dedicated to neutronic analyses of the improved ESFR core design. The conducted work is reported in two parts. Part 1 deals with the evaluation of the safety-related neutronic parameters of the fresh beginning-of-life (BOG) core carried out by eight organizations using both continuous energy Monte Carlo and deterministic computer codes. In addition to the neutronics characterization of the core, a special emphasis was put on the calibration and verification of the computational tools involved in the analyses. Part II is devoted to once-through and realistic batchwise burnup calculations aiming at the establishing of the equilibrium core state, which will later serve as a basis for detailed safety analyses.

ANICCA is the nuclear fuel cycle code developed by the Belgian Nuclear Research Centre (SCK CEN). Nuclear Fuel Cycle codes are of special importance for the assessment of the scenarios and the study of nuclear reactor fleet deployment and decommissioning. In said studies, the flow and inventory of Spent Nuclear Fuel (SNF) are of paramount importance, which are calculated through the irradiation module. In this work, a new approach to the irradiation module is presented. The approach is based on two direct neural networks which predict the final isotopic inventory in the SNF by using the initial fuel composition and the discharge burnup as inputs. These neural networks have been trained in Keras by a database produced with SERPENT2 continuous energy Monte Carlo transport code. Said models are dedicated to two of the most common nuclear fuel technologies for pressurized water reactors: UOX and MOX. Results showed a nice agreement between the new and the classical approach. At the same time, a quicker response in simulations was reported, especially for complex scenarios that involve multi-recycled fuel strategies (known as closed cycle). Thanks to the new method the prebuilt libraries needed in the previous module can be avoided, and so are the simplifications brought by the use of these.