Pressurized Water Reactor Fuel Research Articles

Computational Fluid Dynamics Modeling of a Pressurized Water Reactor Fuel Assembly to Estimate Loss Coefficients in Support of Subchannel Thermalhydraulics Modeling of Pressurized Water Reactor Small Modular Reactors With Advanced Fuels

Abstract Pressure loss coefficients are generally required by subchannel and system thermalhydraulics codes. These coefficients are not readily available for small modular reactors (SMRs) featuring nonconventional designs and novel coolants. In this study, the pressure loss coefficients were obtained using three-dimensional (3D) computational fluid dynamics (CFD) modeling for an advanced water-cooled reactor. A representative light water fuel assembly used in the Organization for Economic Co-operation and Development (OECD)/National Research Council Canada (NRC) pressurized water reactor subchannel and bundle tests (PSBT) benchmark was selected for CFD modeling and simulation under various working conditions. The fuel assembly includes three types of pressurized water reactor (PWR) spacer grids: simple grid (SG), nonmixing vane grid (NMVG), and mixing vane grid (MVG). Turbulent flow through subchannels of both nonheated and heated rod bundles was simulated to predict recoverable and nonrecoverable pressure distribution along the length of the bundle. It was observed that vortices were generated at the tips of spacer grids, affecting the cross-flow in subchannels significantly. The estimated pressure loss coefficients were found to be influenced by the flow conditions (Reynolds number or the upstream flow history) and spacer grid configuration. Pressure loss coefficient values ranged from 1.14 to 1.80, depending on the spacer grid type, design, and flow conditions. The CFD method used in this study was demonstrated to have the potential to generate input parameters required for the subchannel analysis and optimization of fuel assembly designs and serve as a surrogate for empirical correlations.

SiC and FeCrAl as Potential Cladding Materials for APR-1400 Neutronic Analysis

The aim of this study is to investigate the potential improvement of accident-tolerant fuels in pressurized water reactors for replacing existing reference zircaloy (Zr) fuel-cladding systems. Three main strategies for improving accident-tolerant fuels are investigated: enhancement of the present state-of-the-art zirconium fuel-cladding system to improve oxidation resistance, replacement of the current referenced fuel-cladding system material with an alternative high-performance oxidation-resistant cladding, and replacement of the current fuel with alternative fuel forms. This study focuses on a preliminary analysis of the neutronic behavior and properties of silicon carbide (SiC)-fuel and FeCrAl cladding systems, which provide a better safety margin as accident-tolerant fuel systems for pressurized water reactors. The typical physical behavior of both cladding systems is investigated to determine their general neutronic performance. The multiplication factor, thermal neutron flux spectrum, 239Pu inventory, pin power distribution, and radial power are analyzed and compared with those of a reference Zr fuel-cladding system. Furthermore, the effects of a burnable poison rod (Gd2O3) in different fuel assemblies are investigated. SiC cladding assemblies present a softer neutron spectrum and a lower linear power distribution compared with the conventional Zr-fuel-cladding system. Additionally, the SiC fuel-cladding system exhibits behaviors that are consistent with the neutronic behavior of conventional Zr fuel-cladding systems, thereby affording greater economic and safety improvements.

Flow and Heat Transfer Simulation in a Complete Pressurized Water Reactor Fuel Assembly Using Wall-Modeled RANS

Abstract Reproduction of turbulent flow and heat transfer inside a pressurized water reactor (PWR) fuel assembly is a challenging task due to the complex geometry and the huge computational domain. Capability of a wall-modeled Reynolds-averaged Navier–Stokes (RANS) simulation approach has been examined, which had already been validated against the measurements of the MATiS-H experiment. The method is here expanded to a larger computational domain aiming to reproduce flow and thermal field in the entire PWR fuel assembly. Namely, in the first part of this study, wall-modeled RANS is performed in a relatively short section of the representative PWR fuel assembly containing one single mixing grid with an array of 15 × 15 fuel rods. Linear and nonlinear eddy-viscosity turbulence models have been applied; however no significant difference is observed in the predicted pressure drop in the fuel assembly. The obtained predictions revealed an interesting pattern of swirl flow as well as diagonal cross flow downstream the mixing grid, which is driven by the applied design of split-type mixing vanes. In the second part, the computational model is extended to a domain representative of a complete PWR fuel assembly with ten mixing grids, inlet and outlet sections. Pressure drop and flow field are analyzed together with the predicted temperature and potential hot spots. In spite of a relatively coarse spatial resolution of the applied approach, the wall-modeled RANS provided promising results at least for the qualitative prediction of the pressure, flow field, and location of hot spots.

Development of a Simplified Pressurized Water Reactor Fuel Assembly Nonlinear Model

Abstract This paper proposes a computation technique to develop a simplified nonlinear model for a typical nuclear fuel assembly. Because more than a hundred fuel assemblies are packed in the reactor, simplistic model generation is critical to evaluate the motion during an anticipated event such as earthquake. Two straight beams are introduced to simplify the fuel assembly, and the beam properties are moderately defined to represent the skeleton structure and a bundle of slender fuel rods. Because nonlinearity is caused by the interaction between the rods and the spacer grids in the skeleton structure, the two beams are connected with multilinear joints that characterize the mechanical interaction between them. An equation of motion for the model is provided, and the degree of the freedom of the model can be reduced by using a few major modes of the beams. Significant mechanical parameters must be defined reasonably, so a method is proposed to identify unknown parameters through a deterministic calculation and an optimization process. All the information, including the identified parameters, are utilized to develop a nonlinear finite element model with a commercial code. The performance of the model is compared with the test results.

3-Dimensional Multiphysics Modeling of Behaviors of Pressurized Water Reactor Fuel Rods With Missing Pellet Surface

A 3-dimensional (3D) fuel performance analysis program, able to simulate normal operating conditions and accident conditions for PWR fuel behaviors, was developed based on the Multiphysics Object-Oriented Simulation Environment (MOOSE) finite-element framework. By taking fission products swelling, densification and expansion of pellet, thermal and irradiation creep, gap heat transfer, fission gas release, and cladding crack propagation into consideration, detailed fuel behaviors have been simulated in a multiphysics coupling way. Local defects in fuel pellet caused during manufacturing and filling processes known as the missing pellet surface (MPS) can cause abnormal stress distribution of the cladding and it could even lead to cladding failure. Taking Stress Corrosion Cracking (SCC) phenomenon into consideration, a simulation of PWR fuel rodlet that consists of a pellet with an MPS defect and an intact pellet was conducted. The fuel rod has experienced with sorts of events, including normal operating conditions and a high-power ramp event. The simulation results indicated that: 1) The MPS defect affects the temperature and displacement distribution in the vicinity of the MPS defect. When the pellets are in contact with the cladding, the inner surface of the cladding presents a large tensile hoop stress, which accelerates the crack propagation. 2) During the ramp event, the crack propagation rate was higher than that under normal condition and crack length expanded by about 0.1 µm.

Rewetting Analysis of a Pressurized Water Reactor Fuel Slab Using Improved Lumped Models

Abstract This work analyzes the rewetting of a pressurized water reactor (PWR) fuel slab using improved lumped parameter formulations based on Hermite approximations for integrals. The time-dependent two-dimensional rewetting problem is transformed into a quasi-steady heat conduction problem in a reference system moving with the rewetting front by assuming a constant rewetting velocity. The lumped parameter formulations are applied in the slab thickness direction, reducing the two-dimensional problem into a one-dimensional heat conduction equation in the longitudinal direction. The rewetting velocity is obtained by combining the analytical solutions of the average temperature distributions in the reflooded wet region and the liquid deficient dry region at the rewetting front. The Peclet number is obtained as a function of the Biot numbers, dimensionless internal heat generation rate, and dimensionless initial temperatures. The obtained solutions compare favorably with available solutions in the literature.

A new generation of uranium coincidence fast neutron collars for assay of LWR fresh fuel assemblies

The active uranium neutron coincidence collar provides a means of non-destructively assaying the fissile linear density of Light Water Reactor fresh fuel assemblies containing low enriched uranium. These neutron collars can operate in two modes: a thermal and a fast mode. In fast mode, a neutron collar has an added cadmium (Cd) liner in the sample cavity of the detector to reduce the impact of the burnable poison (thermal neutron absorber) on the detector signal (doubles). The main advantage for operating in fast mode is a detected signal that is less dependent of the burnable neutron poison content and thus less dependent on facility operator declarations. The drawback is that operating in fast mode requires a longer measurement time (∼hour vs tens of minutes for thermal mode) to achieve the statistically needed precision in the measurements. The trend in the modern reactor fuel assemblies is moving to higher burnup by using higher initial enrichment and, consequently, a higher number of burnable poison rods to compensate the initial neutron reactivity. The increase of the burnable poison loading has motivated the development of a new generation of high efficiency fast neutron collars to allow practical measurements in-field by nuclear inspectors. This paper describes the development and performance evaluation of a new generation of neutron collars, for both boiling water reactor (BWR) and pressurized water reactor (PWR) fuels, jointly developed between the US Department of Energy, through Los Alamos National Laboratory, and the Euratom Safeguards Directorate of the European Commission. We present here calibrations with reference fuel assemblies at Los Alamos National Laboratory as well as the results of in-field measurement campaigns in fuel fabrication plants with modern commercial fuel assemblies. The experimental results show that a typical PWR verification can be made in a total time of 30 min with an uncertainty in the measured mass of 2% at one standard deviation (1σ). A BWR verification can be made in 47 min with an uncertainty in the measured mass of 1.9% at 1σ, or a total time of 20 min with 1σ uncertainty in the measured mass of 2.5%.

Effect of TiN Coating on the Fouling Behavior of Crud on Pressurized Water Reactor Fuel Cladding

A titanium nitride film was applied to mitigate the deposition of fouling deposits (crud) on pressurized water reactor fuel cladding. Titanium nitride was chosen as it reduces the van der Waals force, believed to be the dominant surface force at pressurized water reactor conditions, between crud and the substrate. Crud deposition experiments were conducted in a heated, flowing pressurized water reactor-simulant loop with uncoated and titanium nitride-coated Zr-Nb-Sn alloy tubes for 1–2 weeks. Crud morphology and thickness were investigated using focused ion beam–scanning electron microscopy, and crud chemical composition was measured via energy-dispersive X-ray spectroscopy. Results demonstrate that the titanium nitride coating effectively reduced crud deposition thickness compared to uncoated Zr-Nb-Sn alloy in the 2-week experiment. In addition, the titanium nitride coating increased the Ni/Fe ratio of the crud, which may help mitigate crud-induced power shifts.

EFFECT OF IODINE AND MOISTURE ON THE MICROSTRUCTURE OF ZIRCALOY-4 UNDER SERVICE CONDITION IN PHWR

Fuel failures are always a cause of concern in any nuclear reactors as it increases the manrem consumption of radiation workers. Although performance of the fuels in pressurized heavy water reactors is good, but still fuel failures occur occasionally. Post irradiation examination (PIE) of the failed fuel elements indicates internal hydriding, not deuteriding, as a major cause for the failures, although secondary deuteriding occurs and, in a few cases, failures are associated with defects in the end plug weld. The sources of hydrogen are either fuel pellets or the clad or the graphite coating. Restriction has been imposed on maximum content of total hydrogen in the fuel element to 1 mg to prevent hydriding of the Zircaloy clad tube. Accidental pick up of hydrogen occurs, which could lead to failure of the fuel bundles. Experimental investigations have been conducted to understand the individual effect of iodine and accidental pick up of moisture on the microstructure of Zircaloy-4 end cap welded samples with graphite coating. Results indicate that severe hydriding in Zircaloy-4 samples due to the existence of internal moisture in presence of graphite under service condition may result in fuel failure and justifies the findings of PIE.

Pb(Mg1/3Nb2/3)-PbTiO3-Based Ultrasonic Transducer for Detecting Infiltrated Water in Pressurized Water Reactor Fuel Rods.

In this study, a high-sensitivity Pb() (PMN-PT)-based ultrasonic transducer was developed for detecting defective pressurized water reactor (PWR) fuel rods. To apply the PMN-PT substance to nuclear power plant facilities, given the need to guarantee their robustness against radioactive materials, the effects of neutron irradiation on PMN-PT were investigated. As a result, the major piezo-electric constants of PMN-PT, such as the electrical impedance, dielectric constant, and piezo-electric charge constant, were found to vary within acceptable ranges. This means that the PMN-PT could be used as the piezo-electric material in the ultrasonic transducer for nuclear power plants. The newly developed ultrasonic transducer was simulated using a modified KLM model for the through-transmission method and fabricated under the same conditions as in the simulation. The through-transmitted waveforms of normal and defective PWR fuel rods were obtained and compared with simulated results in the time and frequency domains. The response waveforms of the newly developed ultrasonic transducer for pressurized water reactor (PWR) fuel rods showed good agreement with the simulation outcome and could clearly detect defective specimens with high sensitivity.

Neutronic Study of Burnable Poison Materials and Their Alternative Configurations in Fully Ceramic Microencapsulated Loaded Pressurized Water Reactor Fuel Assembly

Use of FCM fuel in light water reactors is an attractive option for existing and future generations of these reactors to make them accident tolerant in nature. This work focuses on the neutronic study of the use of burnable material in various configurations to control the excess reactivity and to keep the moderator temperature coefficient of reactivity (MTC) feedback negative for entire cycle length. Erbia and gadolinia, two conventional materials are used in three different configurations including quadruple isotropic (QUADRISO), bi-isotropic (BISO), and Matrix Mix forms. The results obtained from the implicit random treatment of the double heterogeneity of tri-structural isotropic (TRISO), QUADRISO, and BISO particles show that the erbia is the best material to be used in QUADRISO and Matrix Mix configurations with lowest reactivity swing for the life cycle and residual poison well below 0.5%. Gadolinia is usable in FCM environment only in the BISO form where enhanced self-shielding controls the depletion performance of the material. The gadolinia has almost zero residual poison at end of cycle (EOC); however, it has relatively large reactivity swing, which will need more micromanagement of the control rods during the plant operations. At the beginning of cycle (BOC), erbia-loaded assemblies have shown an increase in negative value of MTC compared with reference due to presence of resonance peak in erbium near 1 eV. The finally recommended material-configuration combinations have shown the excess reactivity containment in desired manner with good depletion performance and negative feedback of the MTC for life cycle.

Multiphysics Modeling of Pressurized Water Reactor Fuel Performance

Nuclear fuel rods operate under complex radioactive, thermal, and mechanical conditions. Nowadays, fuel rod codes usually make great simplifications on analyzing the multiphysics behavior of fuel rods. The present study develops a three-dimensional (3D) module within the framework of a general-purpose finite element solver, i.e., abaqus, for modeling the major physics of the fuel rods. A typical fuel rod, subjected to stable operations and transient conditions, is modeled. The results show that the burnup levels have an important influence on the thermomechanical behavior of fuel rods. The swelling of fission products causes a dramatically increasing strain of pellets. The variation of the stress and the radial displacement of the cladding along the axial direction can be reasonably predicted. It is shown that a quick power ramp or a reactivity insertion accident can induce high tensile stress in the outer regime of the pellet and may cause further fragmentation to the pellets. Fission products migration effects and differential thermal expansion become more severe if there are flaws or imperfections on the pellet.

Thermal hydraulic assessment of a new design of PWR fuel assembly

Abstract The recycling of the spent fuel of nuclear reactors is the important aspects in the fuel cycle technology. One of the recycling of the spent fuel is reuse of spent PWR (pressurized water reactor) fuel in CANDU reactor after re-fabricated into CANDU (CANda deuterium uranium) fuel bundles; this technology called Direct Use of Spent PWR Fuel In CANDU Reactor (DUPIC). This technology was modified however; it introduced a new design for the PWR fuel assembly. The new fuel can be transferred directly to CANDU reactors without processing, only after a certain cooling time. In this study the evaluation from thermal hydraulic point of view is done. A Westinghouse PWR fuel assembly was selected as a reference. Thermal hydraulic simulation for reference and new design fuel occurred by using RELAP5 (Reactor Excursion and Leak Analysis Program) code, then comparing the both results. The hottest rod was considered a worst condition in the reactor so the model simulated it rather than whole core. The new design achieves a good results comparing with a reference one.

Flexibility of ADS for minor actinides transmutation in different two-stage PWR-ADS fuel cycle scenarios

A two-stage Pressurized Water Reactor (PWR)-Accelerator Driven System (ADS) fuel cycle is proposed as an option to transmute minor actinides (MAs) recovered from the spent PWR fuels in the ADS system. At the second stage, the spent fuels discharged from ADS are reprocessed by the pyro-chemical process and the recovered actinides are mixed with the top-up MAs recovered from the spent PWR fuels to fabricate the new fuels used in ADS. In order to lower the amount of nuclear wastes sent to the geological repository, an optimized scattered reloading scheme for ADS is proposed to maximize the discharge burnup and lower the burnup reactivity loss. Then the flexibility of ADS for MA transmutation is evaluated in this research. Three aspects are discussed, including: different cooling time of spent ADS fuels before reprocessing, different reprocessing loss of spent ADS fuels, and different top-up MAs recovered from different kinds of spent PWR fuels. The ADS system is flexible to be combined with various pyro-chemical reprocessing technologies with specific spent fuels cooling time and unique reprocessing loss. The reduction magnitudes of the long-term decay heat and radiotoxicity of MAs by transmutation depend on the reprocessing loss. The ADS system is flexible to transmute MAs recovered from different kinds of spent PWR fuels, regardless of UOX or MOX fuels. The reduction magnitudes of the long-term decay heat and radiotoxicity of different MAs by transmutation stay on the same order.

Metal Fuel Development and Verification for Prototype Generation IV Sodium-Cooled Fast Reactor

Abstract Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU)–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic–martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

Analysis of high burnup pressurized water reactor fuel using uranium, plutonium, neodymium, and cesium isotope correlations with burnup

Abstract The correlation of the isotopic composition of uranium, plutonium, neodymium, and cesium with the burnup for high burnup pressurized water reactor fuels irradiated in nuclear power reactors has been experimentally investigated. The total burnup was determined by Nd-148 and the fractional 235U burnup was determined by U and Pu mass spectrometric methods. The isotopic compositions of U, Pu, Nd, and Cs after their separation from the irradiated fuel samples were measured using thermal ionization mass spectrometry. The contents of these elements in the irradiated fuel were determined through an isotope dilution mass spectrometric method using 233U, 242Pu, 150Nd, and 133Cs as spikes. The activity ratios of Cs isotopes in the fuel samples were determined using gamma-ray spectrometry. The content of each element and its isotopic compositions in the irradiated fuel were expressed by their correlation with the total and fractional burnup, burnup parameters, and the isotopic compositions of different elements. The results obtained from the experimental methods were compared with those calculated using the ORIGEN-S code.

Hydraulic lift-off issues for application of high performance annular fuels in pressurized water reactors

In the PWR core, the fuel assembly is firmly seated on the lower core plate during operation. However, if the hydraulic force exerted on the fuel assembly by coolant flow is too large and the fuel assembly is lifted-off from the lower core plate, the excessive vibration will cause fuel failure. Therefore, the hydraulic lift-off issue needs to be addressed when the advanced fuel assembly is developed. It has been shown that the advanced annular fuel design with internal cooling allows power uprating up to 50% while the peak temperature of the fuel can be reduced and the MDNBR can be maintained. However, if the coolant condition in the core is kept unchanged, increasing the core power by 50% requires the core flow rate also increase proportionally, which will give rise to the hydraulic lift-off, an important issue to be addressed. In this paper, taking the 17×17 solid fuel design as the reference, the hydraulic lift-off issue is investigated for proposed 12×12 and 13×13 annular fuel designs. Both the steady-state and start-up operating conditions are evaluated. It is found that the hydraulic lift-off indeed is an issue for annular fuel design which requires careful analysis. By comparison, the lift-off forces and hold-down forces required for the externally and internally cooled annular fuels (13×13 and 12×12 arrays) are several times larger than that of the referenced solid fuel (17×17 array). Therefore, the hold-down mechanism for annular fuel needs to be carefully designed.

Preliminary Economic Evaluation of Thorium-Based Fuels in PWRs

Many studies have proven that thorium-based fuel cycles are more expensive than current uranium fuel cycles, which is confirmed in this paper as well. This research, however, goes beyond the conventional nuclear plant refueling cycle, aiming to optimize the refueling cycle in line with specific advantages realized from the thorium-based refueling cycle. These benefits are the good thermal neutronic characteristics of fertile 232Th and fissile 233U, resulting in longer refueling cycles, higher capacity factors, and reduced volumes of spent fuel. This paper focuses on once-through, homogeneously mixed, thorium-uranium fuel cycles and explores the economic advantages to introducing thorium as a fertile component in pressurized water reactor fuels as compared to once-through conventional uranium-only cycles. The economic evaluation compares the operational savings incurred as a result of longer fuel cycles and reduced reactor downtime with increased fuel expenses, due to higher initial enrichment of uranium and downblending with thorium. Uranium fuel is compared with thorium-uranium fuel in terms of the fuel cycle costs, reactor downtime cost due to refueling, and income due to electricity sales.Thorium-uranium fuel costs more than uranium, but the income from electricity sales and reactor downtime refueling cost is advantageous. These factors have also been taken into account and prove that the advantage of increased income outweighs the higher fuel costs. Issues related to longer fuel cycles and higher burnups need to be addressed, but the development and adoption of once-through, homogeneous, thorium-uranium fuel can be justified by a clear economic benefit for utilities operating nuclear power plants.

Uncertainty Analysis on Reactivity and Discharged Inventory due to 235,238U, 239,240,241Pu, and Fission Products: Application to a Pressurized Water Reactor Fuel Assembly

Uncertainty analysis on reactivity and discharged inventory for a typical pressurized water reactor fuel element as a result of uncertainties in 235,238U, 239,240,241Pu, and fission products nuclear data was performed. A typical Westinghouse three-loop fuel assembly fueled with UO2 fuel with 4.8% enrichment was selected. The Total Monte Carlo method was applied using the deterministic transport code DRAGON. This code allows the generation of the few-groups nuclear data libraries by directly using data contained in the nuclear data evaluation files. The nuclear data used in this study are from the JEFF3.1 evaluation, with the exception of the nuclear data files for U, Pu, and fission products isotopes (randomized for the generation of the various DRAGON libraries). These are taken from the TALYS evaluated nuclear data library TENDL-2012. Results show that the calculated total uncertainty in keff (as a result of uncertainties in nuclear data of the considered isotopes) is virtually independent of fuel burnup, and amounts to 700 pcm. The uncertainties in the inventory of the discharged fuel are dependent on the element considered and lie in the range 1% to 15% for most fission products, and are <5% for the most important actinides. The total uncertainty on the reactor parameters was also split into different components (different nuclear reaction channels), and the main sources of uncertainties were identified.

Reactivity Loss Validation of High-Burnup PWR Fuels with Pile-Oscillation Experiments in MINERVE

The ALIX experimental program relies on the experimental validation of the spent fuel inventory, by chemical analysis of samples irradiated in a pressurized water reactor (PWR) between five and seven cycles, and also on the experimental validation of the spent fuel reactivity loss with burnup, obtained by pile-oscillation measurements in the MINERVE reactor. These latter experiments provide an overall validation of both the fuel inventory and the nuclear data responsible for the reactivity loss. This program also offers unique experimental data for fuels with a burnup reaching 85 GWd/tonne, as spent fuels in French PWRs have never exceeded 70 GWd/tonne up to now.The analysis of these experiments is done in two steps with the APOLLO2/SHEM-MOC/CEA2005v4 package. In the first step, the fuel inventory of each sample is obtained by assembly calculations. The calculation route consists of the self-shielding of cross sections on the 281-energy-group SHEM mesh, followed by flux calculation by the method of characteristics in a two-dimensional exact heterogeneous geometry of the assembly, and finally a depletion calculation by an iterative resolution of the Bateman equations. In the second step, the fuel inventory is used in the analysis of pile-oscillation experiments in which the reactivity of the ALIX spent fuel samples is compared to the reactivity of fresh fuel samples. The comparison between experiment and calculation shows satisfactory results with the JEFF3.1.1 library, which predicts the reactivity loss within 2% for burnup of ˜75 GWd/tonne and within 4% for burnup of ˜85 GWd/tonne.