Axial Power Profile Research Articles

Study on potential burnable poison materials for a small modular block-type HTGR design using MgO-BeO as a composite-based moderators

Preliminary analysis of MgO-BeO composite material used as a moderator in a 50 MWt block-type high-temperature gas-cooled reactor (HTGR) was performed in our previous study. The target burnup of 80 GWd/t was achieved with a uniform fuel composition of 17 wt% 235U enrichment and 6 kg of heavy metal per fuel block. However, this resulted in high excess reactivity and a peak in axial power distribution at the core center. Therefore, this study aims to reduce excess reactivity by incorporating burnable poison (BP) material and optimize the axial power profile by introducing a nonuniform fuel composition in the core. Neutronic calculations were performed using the Monte Carlo MVP3.0 code developed by the Japan Atomic Energy Agency (JAEA). In this study, three fuel enrichments of 235U, ranging from 15 wt% to 20 wt%, were distributed across the core while maintaining a constant fuel packing fraction of 45 %. The results showed that the higher power density distribution shifted from the core’s center to its upper part, leading to lower power density in the bottom region than the top. In addition, excess reactivity was reduced by inserting BP rods. Several parametric calculations were performed to achieve minimal excess reactivity without compromising the burnup target. The results showed that the BP rod with a radius of 0.7 cm and 12 wt% of Gd2O3 can reduce the excess reactivity from 25.5 %Δk/k to 13.47 % Δk/k.

Numerical simulation of axial and radial decay heat distribution in spent RBMK-1500 fuel assembly

The RBMK-1500 reactor is a boiling water reactor. Thus, the coolant density and the neutron moderation efficiency vary axially over the fuel assembly height influencing the distribution of decay heat in the discharged fuel. The underestimation of this effect on the decay heat may lead to biases in safety assessments of the spent fuel management. Axial coolant density and power profiles were used in this work to determine axial and radial decay heat distributions in the RBMK-1500 fuel assembly over the first 100 years of cooling using the SCALE code system. It was found that the application of axial profiles influences the increase of estimated total decay heat by 3 % just 2 years after the irradiation is finished. However, the estimation of total decay heat is lower by 2 % after 100 years of cooling compared to the case with no applied axial profiles. The maximum of the normalized decay heat profile also varies between 1.31 and 1.40 over the cooling time. The estimated decay heat in outer pins over the cooling time is higher by 10 ÷ 14 % compared to the decay heat produced in inner pins. The variation of results over the cooling phase and reasons behind the difference in estimated decay heat was explained by analysing nuclides which mainly contribute to the total decay heat.

Towards a more realistic MELCOR model for a dry cask for spent nuclear fuel. Part II: application.

Nowadays, a great deal of attention is devoted to the development of best-estimate models able to produce more realistic outcomes. This is also the case for system codes, such as MELCOR, that are being mostly used in a conservative way especially when dealing with the licensing process. The above-mentioned need for more realistic results is at the core of this two-paper series related to the creation of a more accurate MELCOR model for the HI-STORM 100S dry cask. The findings obtained from the sensitivity studies carried out in the Part I are leveraged to set up an improved MELCOR model, the characteristics of which are consistent with the typical features of Spent Nuclear Fuel (SNF), and with geometrical and material properties of the cask itself. The addition of an axial power profile in the Fuel Assembly (FA), the better characterization of the flow losses in the air gap between internal metallic canister and external concrete-based overpack, and the choice of an appropriate value for the concrete thermal conductivity, are taken into account conjointly in this Part II. The outcomes from the improved MELCOR simulation are reported mainly in terms of the Peak Cladding Temperature (PCT), being the variable under regulatory surveillance. However, in addition to PCT, calculated temperature profiles are displayed and compared against the ones resulting from the previous model.

Модель реактора ВВЕР-1000 у прецизійному коді SERPENT 2 для розрахунку розподілу енерговиділення в активній зоні

A VVER-1000 reactor model using the Monte Carlo Serpent 2 code for core power distribution calculation is presented. The core and zones located near to the core were modeled in detail, without simplification. The assembly power distribution and axial power profiles were calculated for fresh core of the X2 VVER-1000 benchmark, namely the core of the KhNPP2 first loading for the hot zero power. The results were compared with the data obtained by specialists from Helmholtz-Zentrum Dresden-Rossendorf.

Fuel Performance Analysis of Fast Flux Test Facility MFF-3 and -5 Fuel Pins Using BISON with Post Irradiation Examination Data

Using the BISON fuel-performance code, simulations were conducted of an automated process to read initial and operating conditions from the Pacific Northwest National Laboratory (PNNL) database and reports, which contain metallic-fuel data from the Fast Flux Test Facility (FFTF) MFF Experiments. This work builds on previous modeling efforts involving 1977 EBR-II metallic fuel pins from experiments. Coupling the FFTF PNNL reports to BISON allowed for all 338 pins from MFF-3 and MFF-5 campaigns to be simulated. Each BISON simulation contains unique power and flux histories, axial power and flux profiles, and coolant-channel flow rates. Fission-gas release (FGR), fuel axial swelling, cladding profilometry, and burnup were all simulated in BISON and compared to available post-irradiation examination (PIE) data. Cladding profilometry, FGR, and fuel axial swelling simulation results for full-length MFF metallic pins were found to be in agreement with PIE measurements using FFTF physics and models used previously for EBR-II simulations. The main two peaks observed within the cladding profilometry were able to be simulated, with fuel-cladding mechanical interaction (FCMI), fuel-cladding chemical interaction (FCCI), and thermal and irradiation-induced creep being the cause. A U-Pu-Zr hot-pressing model was included in this work to allow pore collapse within the fuel matrix. This allowed better agreement between BISON-simulated cladding profilometry and PIE measurements for the peak caused by FCMI. This work shows that metallic fuel models used to accurately represent fuel performance for smaller EBR-II pins may be used for full-length metallic fuel, such as FFTF MFF assemblies and the Versatile Test Reactor (VTR). As new material models and PIE measurements become available, FFTF MFF assessment cases will be reassessed to further BISON model development.

Neutronic-coupled thermal hydraulic analysis of flow blockage in the hottest wire-wrapped LBE-cooled fuel assembly

Flow blockage in lead-cooled fast reactors is considered a realistic accident for fuel assemblies, whereas reactivity feedback and axial power profiles are considered in few studies due to the complexity and lack of neutronic modules in commercial computational fluid dynamic software. This study proposes a neutronic-coupled thermal–hydraulic model to analyze the coupling effect on flow blockage in a 19-pin wire-wrapped fuel assembly. Neutronic and thermal hydraulics were validated by comparing the reported benchmark and experimental data, respectively. Simulation results with and without neutronic coupling schemes were compared, and case studies of different reactivity insertions were conducted. The wall overheating of blockages with coupling schemes is more conservative than the constant heat source condition with the lower dimensionless temperature T¯. After a step change of cross-section, the response of Rw and T¯ shows synchronous behavior. For a larger reactivity insertion, it leads to higher wall heat flux, lower wall thermal resistance and lower dimensionless temperature for blockage regions.

On the Role of Axial Neutron Flux Mode Excitation in BWR Stability and Nonlinear Oscillations

Neutron flux modal decomposition is a key tool for analytical and reduced order modeling of boiling water reactor (BWR) stability and oscillations. As a minimum, the fundamental flux mode is used for representing global oscillations while the addition of at least one azimuthal harmonic is needed for simulating the regional out-of-phase mode. Unlike the fundamental and first azimuthal modes, the excitation of an axial flux mode alters the axial power shape but not the total power in the channel and therefore cannot be self-sustained when coupled to density wave–generated reactivity, presumably explaining why it has not been explicitly included in previously published models. Although not self-sustained, the axial mode excitation driven by density wave propagation and interactions with other spatial modes play important roles in interpreting observed BWR stability and oscillations particularly in the nonlinear regime when the oscillation magnitude is large. In this paper, the characteristics of the steady-state axial modes are presented, and their impact on oscillation dynamics for small and large amplitudes of both the global and the regional oscillations is studied using reduced order analytical tools. Aside from the oscillating component, our research results identify an average nonzero axial mode component to develop during limit cycle oscillations that causes the average axial power profile to shift toward the bottom of the core and thus contributes a negative reactivity component. The emergence of this nonzero average axial mode component and the associated negative reactivity were found to diminish the power increase due to global mode power oscillations and contribute to nonlinear stabilization of regional oscillations. The physical interpretation of nonlinear power oscillations with the inclusion of the axial mode component resolves previously unexplained results obtained from high-fidelity numerical models.

Modeling of sodium-cooled fast reactor core – A sensitivity study on number of core channels and axial meshes

System thermal–hydraulic codes are used in the dynamics analysis of sodium-cooled fast reactors (SFRs). In such simulations, the accuracy of the reactor core model is very important to establish plant safety. Coupled physical phenomena, namely, heat transfer, fluid dynamics, and neutronics, need to be simulated in the core. Due to the complexity, a simplified core modeling approach has been traditionally followed in system dynamics codes. Firstly, steady-state core neutronics calculations are carried out using a 2D axisymmetric model to obtain perturbation worth and power distributions. Later, this data is mapped to several core channels (usually 10 to 20), each representing a set of fuel subassemblies (SAs) for thermal-hydraulics calculations.In this paper, a more sophisticated core modeling approach has been followed to understand the importance of such an exercise. A 3D neutronics model was used to obtain subassembly-wise perturbation worths and power distributions. For thermal-hydraulics calculations, the number of core channels was increased with the idea of one subassembly per channel. The number of axial meshes in each channel was also increased. The traditional and new modeling approaches were used to simulate the benchmark problem of FFTF LOFWOS test#13 for comparison. Steady-state parameters from both the new and the old models were comparable. Notably, there was a peaking of the axial power profile in the new model compared to the old model, which resulted in slightly higher peak fuel temperatures in the new model. However, the difference in the evolution of transient parameters between the two models was insignificant, and both models compared reasonably well with the benchmark data.

Thermal-hydraulic stability investigations of once through helically coiled steam generators for SMR applications

Once-through steam generators with helically-coiled tubes (OTSG-HC) are susceptible to density wave instability phenomena. In order to investigate the stability performance of the CAREM-25 OTSG-HC, a new stability model is presented. Such a model being linear and strictly non-diffusive is able to represent arbitrary heat flux profiles, which is essential to study realistic cases. The OTSG-HC tubes are modeled by three different regions representing the liquid zone, the boiling zone, and the superheated vapor zone. Moreover, the coolant from the primary side is coupled with a model representing the phenomena present in CAREM-25. A convenient parameter space for stability representation is found and used to characterize the CAREM-25 OTSG-HC stability performance.In this work, special attention is paid to the calculation of the friction pressure losses in the coolant flowing through the helically coiled tubes. For this reason, a best estimate correlation is implemented and used in the model together with a sophisticated axial power profile model which allows representing realistic boundary conditions. As a result, it is confirmed the crucial importance of correctly representing the axial power profile transferred in the helically coiled tubes. In addition, it is found the commonly used uniform axial power profile leads to non-conservative stability predictions particularly at low power conditions. The latter is explained by the underestimation of the boiling and superheating lengths, which is proved to have a destabilizing effect.Stability maps are presented in terms of the transferred power and the primary coolant operational pressure, i.e. QSG-PPri plane. It is observed that when modeling realistic power profiles, the less stable operational condition of the OTSG-HC is found at very low powers, for all analyzed range of primary pressures. In addition, a weak stability dependence of the OTSG-HC with the pressure of the primary side is observed. Finally, the system is stabilized in the region of interest and its stability performance quantified.

High-Fidelity Steady-State and Transient Simulations of an MTR Research Reactor Using Serpent2/Subchanflow

In order to join efforts to develop high-fidelity multi-physics tools for research reactor analysis, the KIT is conducting studies to modify the coupled multi-physics codes developed for power reactors. The coupled system uses the Monte Carlo Serpent 2 code for neutron analysis and the Subchanflow code for thermo-hydraulic analysis. Serpent treats temperature dependence using the target motion sampling method and Subchanflow was previously extended and validated with experimental data for plate-type reactor analysis. This work present for the first time the steady-state and transient neutron and thermo-hydraulic analysis of an MTR core defined in the IAEA 10 MW benchmark using Serpent2/Subchanflow. Important global and local parameters for nominal steady-state conditions were obtained, e.g., the lowest and highest core plate/channel power/temperature, the radial and axial core power profile at the plate level, and the core coolant temperature distribution at the subchannel level. The capabilities of Serpent2/Subchanflow to perform transient analysis with on-the-fly motion of the control plates were tested, namely with fast and slow reactivity insertion. Based on the unique results obtained for the first time at the subchannel and plate level, it can be stated that the coupled Serpent2/Subchanflow code is a very promising tool for research reactor safety-related investigations.

Investigation of Critical Heat Flux for Plutonium-Based Mixed Oxide Advanced Fuel Bundle Design

Abstract The critical heat flux (CHF) performance of an advanced plutonium-based mixed oxide (MOX) fuel for potential use in a pressure tube heavy water reactor (PT-HWR) has been studied experimentally at Canadian Nuclear Laboratories with an electrically heated string simulator of 43-element fuel bundles. The fuel simulator has a uniform axial power profile and a radial power profile that is representative of the plutonium-based MOX fuel. The measurements of CHF caused by liquid-film dryout were made in the MR-3 heat transfer loop facility using R-134a refrigerant as the working fluid. The test matrix included system pressures from 1.47 to 2.11 MPa, mass flow rates from 12.7 to 14.7 kg/s and inlet temperatures from 31 to 59 °C, which are representative of the water-equivalent reactor operating conditions of 9 to 12.5 MPa pressure, 13.5 to 21.3 kg/s mass flow rate and the desired inlet subcoolings. Compared to conventional natural uranium fuel, the radial power profile of a MOX fuel exhibits a steeper and less even distribution across the fuel element rings, with a relatively higher power in the outer ring. It was found that CHF values of the MOX fuel are significantly lower than those of the natural uranium fuel. Based on the experimental data, a correlation has been derived to account for the effect of radial power profile on CHF. This correlation can be used to evaluate the relative CHF values of advanced or nonconventional fuel designs with radial power profiles deviating from that of natural uranium fuel.

Cladding Profilometry Analysis of Experimental Breeder Reactor-II Metallic Fuel Pins with HT9, D9, and SS316 Cladding

BISON finite element method fuel performance simulations were conducted using an existing automated process that couples the Fuels Irradiation & Physics Database (FIPD) and the Integral Fast Reactor Materials Information System database by writing input files and comparing the BISON output to post-irradiation fuel pin profilometry measurements contained within the databases. The importance of this work is to demonstrate the ability to benchmark fuel performance metallic fuel models within BISON using Experimental Breeder Reactor-II fuel pin data for a number of similar pins, while building off previous modeling efforts. Changes to the generic BISON input file include implementing pin specific axial power and flux profiles, pin specific fluences, frictional contact, and irradiation-induced volumetric swelling models for cladding. A statistical analysis of irradiation-induced volumetric swelling models for HT9, D9, and SS316 was performed for experiments X421/X421A, X441/X441A, and X486. Between these three experiments, there were 174 post-irradiation examination (PIE) profilometries used for validating the swelling models presented using a standard error of the estimate (SEE) method. Implementation of the volumetric swelling models for D9 and SS316 claddings was found to have a significant impact on the BISON profilometry simulated, where HT9 clad pins had an insignificant change due to low fluence values. BISON profilometry simulated for HT9, D9, and SS316 fuel pins agreed with PIE profilometry measurements, with assembly SEE values being 4.4 × 10−3 for X421A, 2.0 × 10−3 for X441A, and 2.8 × 10−3 for X486. D9 clad pins in X421/X421A had the highest SEE values, which is due to the BISON simulated profilometry being shifted axially. While this work accomplished its purpose to demonstrate the modeling of multiple fuel pins from the databases to help validate models, the results suggest that the continued development of metallic fuel models is necessary for qualifying new metallic fuel systems to better capture some physical performance phenomena, such as the hot pressing of U-Pu-Zr and the fuel cladding chemical interaction.

DEVELOPMENT OF 3D PIN-BY-PIN CORE SOLVER TORTIN AND COUPLING WITH THERMAL-HYDRAULICS

Currently, safety analyses mostly rely on codes which solve both the neutronics and the thermal-hydraulics with assembly-wise nodes resolution as multiphysics heterogeneous transport solvers are still too time and memory expensive. The pin-by-pin homogenized codes can be seen as a bridge between the heterogeneous codes and the traditional nodal assembly-wise calculations. In this work, the pin-by-pin simplified transport solver Tortin has been coupled with a sub-channel code COBRA-TF. The verification of the 3D solver of Tortin is presented at first, showing very good agreement in terms of axial and radial power profile with the Monte Carlo code SERPENT for a small minicore and with the state-of-the-art nodal code SIMULATE5 for a quarter core without feedback. Then the results of Tortin+COBRA-TF are compared with SIMULATE5 for one assembly problem with feedback. The axial profiles of power and moderator temperature show good agreement, while the fuel temperature differ by up to 40 K. This is caused mainly by different gap and fuel conductance parameters used in COBRA-TF and in SIMULATE5.

VVER-1000 BENCHMARK INTERPRETATION WITH MONTE CARLO CODE MCS

An interpretation of the NEA-1517/82 benchmark from the SINBAD shielding database has been conducted with the MCS Monte Carlo code developed at the Ulsan National Institute of Science and Technology (UNIST) and the ENDF/B-VII.1 nuclear data library. The NEA-1517/82 benchmark corresponds to experiments on a VVER-1000 critical mock-up (thermal reactor with hexagonal fuel lattice) inside the LR-0 research reactor operated by the Nuclear Research Institute (NRI) in the Czech Republic. A new 3D model of the VVER-1000 mock-up core is developed for MCS based on the SINBAD documentation. The model includes the top and bottom parts of fuel pins, the spacer grids and core components: baffle, barrel, downcomer, tank, reactor pressure vessel (RPV) and concrete block used as biological shielding. The quality of the model is verified first by code/code comparison of MCS against MCNP6 for criticality and power distributions (pin-by-pin and axial power). The validation of MCS results is then performed against six critical cases, 260 measured pin powers and benchmark calculations of the axial power profile. Finally, a comparison of calculated and measured neutron spectra inside the mock-up core is presented as a preliminary study for upcoming works on the deep-penetration shielding capability of MCS.

Multiphysics Modeling and Validation of Spent Fuel Isotopics Using Coupled Neutronics/Thermal-Hydraulics Simulations

Multiphysics coupling of neutronics/thermal-hydraulics models is essential for accurate modeling of nuclear reactor systems with physics feedback. In this work, SCALE/TRACE coupling is used for neutronic analysis and spent fuel validation of BWR assemblies, which have strong coolant feedback. 3D axial power profiles with coolant feedback are captured in these advanced simulations. The methodology is applied to two BWR assemblies (2F2DN23/SF98 and 2F2D1/F6), discharged from the Fukushima Daini-2 unit. Coupling is performed externally, where the SCALE/T5-DEPL module transfers axial power data in all axial nodes to TRACE, which in turn calculates the coolant density and temperature for each of these nodes. Within a burnup step, the data exchange process is repeated until convergence of all coupling parameters (axial power, coolant density, and coolant temperature) is observed. Analysis of axial power, criticality, and coolant properties at the assembly level is used to verify the coupling process. The 2F2D1/F6 benchmark seems to have insignificant void feedback compared to 2F2DN23/SF98 case, which experiences large power changes during operation. Spent fuel isotopic data are used to validate the coupling methodology, which demonstrated good results for uranium isotopes and satisfactory results for other actinides. This work has a major challenge of lack of documented data to build the coupled models (boundary conditions, control rod history, spatial location in the core, etc.), which encourages more advanced methods to approximate such missing data to achieve better modeling and simulation results.

Thermal-hydraulics analysis of flow blockage events for fuel assembly in a sodium-cooled fast reactor

The subchannel code ATHAS-LMR has been improved for analyzing thermal-hydraulics characteristics of flow blockage events in the fuel assembly of a sodium-cooled fast reactor (SFR). Based on the in-depth study of the phenomenon of flow blockage and combined with the characteristics of fast reactor assembly, enhanced models for the wire-wrap spacer and convective heat transfer have been added to account for changes in transverse flow in the presence of blockages. And appropriate transverse mixing and local resistance model are also implemented in ATHAS-LMR-FB. The improved code has been assessed against experimental data obtained at adiabatic and diabatic conditions in SFR assemblies with simulated flow blockages. Good agreement between predictions and experimental data on velocity and temperature distributions downstream of the blockage has been observed. Comparison of predictions of various analytical tools have also been performed. Predictions of the improved ATHAS-LMR code are shown to be compatible with those of other analytical tools. An in-depth analysis of effects of location and size of blockage has been carried out for the fuel assembly of the China Experimental Fast Reactor. Boiling has been predicted to occur for blockage ratios greater than 55.61% at the central location of the fuel assembly. Peak cladding and coolant temperatures are predicted at the axial location 70% of the rod length (just downstream of the maximum local power position of the non-uniform axial power profile).

Advanced BWR criticality safety part II: Cask criticality, burnup credit, sensitivity, and uncertainty analyses

In this study, an analysis on burnup credit for cask criticality safety in BWR spent fuel is conducted. Accurate burnup credit can be used to reduce overly conservative safety margins to increase shipping and storage efficiency while maintaining criticality level within regulatory limits. This analysis is based on advanced lattice depletion models that capture various complexities associated with BWR operation. This paper describes the second part of the two-part study which performs an out-of-core analysis of spent fuel in a transportation/storage cask. The first part of the study (Radaideh et al., 2019) developed the set of depletion models used here. In this paper, the spent fuel compositions resulting from these depletion models are used for cask criticality calculations. Uncertainty quantification of cask keff is performed by combining the uncertainty in isotope inventory, nuclear data, and the statistical sampling in KENO-V.a. The uncertainty in isotopic inventory is quantified by performing a validation analysis by comparing spent fuel compositions calculated by 2D TRITON to experimentally determined spent-fuel assay data for three reactors: Fukushima Daini-2, Cooper-1, and Gundermmingen-A. The validation results demonstrate good agreement for the uranium isotopes as compared to the plutonium isotopes. Also, it was found that the uncertainty in cask keff is dominated by the isotopic uncertainty and can reach about 2500 pcm, and as low as about 1700 pcm. Final results show that axial power profile, axial coolant density, control rod modeling, and the presence of gadolinium in 3D simulations have the largest effects on BWR burnup credit. This implies the need for detailed 3D modeling for accurate BWR burnup credit analysis. In addition, based on the UQ analysis considering both actinide only and actinide and fission products sets, the cask remains subcritical within 2σ for all depletion cases analyzed (C0-C9), even though the cask is assumed to be flooded with water and the lattices are discharged at their peak reactivity.

Time resolved temperature diagnosis in dynamic hohlraum experiments on JULONG-I facility

Axial radiation properties of Z-pinch dynamic hohlraum have been investigated experimentally on JULONG-I facility in China. A new method of time resolved temperature diagnostics in dynamic hohlraum experiments has been developed via a one-dimensional (1-D) imaging system consisting of a streak camera. With the help of the axial radiation power profile measured by an X-ray power measurement system using the same type of scintillator as in the 1-D imaging system, time resolved temperature distribution along the one-dimensional space in dynamic hohlraum experiments are extracted by a special mathematical transform process of the image. The peak radiation temperature in the centre is more than 100 eV. The radiation temperatures are also estimated by the 1-D Multi-IFE-Z program.

Verification and application of SuperMC3.3 to lead-bismuth-cooled fast reactor

Lead-cooled fast reactors have multilayered designs and large internal temperature differences, which cause challenges in simulating reactor physics. SuperMC, a large-scale integrated software system for neutronics design, is inherently able to address complex geometries and multi-temperature problems. The purpose of this study is to verify the applicability of SuperMC to the lead-bismuth-cooled fast reactor RBEC-M. The multi-temperature cross-section generation function of SuperMC was employed and showed good performance. Based on the ENDF/B-VII.1 library, the effective multiplication factor keff obtained by SuperMC showed good agreement with those from previous works. The relationship of keff and 15N enrichment applied to the fuel material was also studied, with the results showing that in creased 15N could significantly improve keff. The axial power profile and kinetics parameters for the bench mark were then calculated and analyzed. This work thus verified the applicability of SuperMC for comprehensive neutronics simulations for lead-bismuth-cooled fast reactors.

Neutronic design of homogeneous thorium/uranium fuel for 24 month fuel cycles in the European pressurized reactor using MCNP6

Abstract A new approach to designing thorium based fuels for a European pressurized reactor (EPR) core is introduced by considering a three dimensional full core MCNP6 model. Existing information on designing thorium/uranium fuel for Generation II pressurized water reactors (PWRs) have been analyzed and the new fuel was proposed for the EPR (Gen III+). The newly designed fuel assembly (Th-B1) can reach 24-month fuel cycles without altering the geometry or disregarding any design limits. The Th-B1 was developed by reducing and ultimately removing Gd2O3 and slightly reducing the overall initial fissile content compared to the uranium only EPR. The same overall initial fissile content of the U-EPR was maintained, with the exception for fuel pin sections where pure ThO2 replaced the UO2 and Gd2O3 containing sections of the original U-EPR design. Results showed that there is no need to increase the initial fissile content or soluble boron enrichment as predicted by previous studies. Both xenon and samarium fission product yields for the Th-B1 are lower than for the U-B1, which contributes to a higher value for k ∞ and better breeding. Results also showed that the total fissile content for the Th-B1 shows a smaller decrease with time, which implies a smaller decrease for k ∞ with time and subsequently an extension of the fuel cycle. The innovative optimized fuel design of the EPR (five axial zones), enabled fuel enrichment savings and longer fuel cycles for Th-EPR. Not only does one save on fuel costs by keeping the overall initial fissile content similar, but this also results in reactor properties that are similar to those of the U-EPR and no serious changes need to be made. The only concerns are a lower delayed neutron fraction and boron worth for the Th-B1 at EOL. However, not all assemblies in a full core will be at the end of their life and if a three batch reloading scheme is assumed, the boron worth and delayed neutron fraction will most likely be within acceptable limits. Other differences between the Th-EPR and U-EPR such as the axial power profile and difference in boron letdown curve can be accounted for by amending the operating procedures during the design phase.