Burnup Analysis Research Articles

Neutronic and burnup analysis of OECD-NEA MOX fuel benchmark with Dragon5 lattice code.

This paper presents a comprehensive and detailed analysis of the OECD-NEA MOX fuel benchmark based on different nuclear data libraries to investigate the reliability and accuracy of the Dragon5 lattice code developed by École Polytechnique de Montréal for the neutronic analysis of mixed uranium-plutonium oxide (MOX) fuel. The neutronics and burn-up calculations for rectangular pin and assembly geometries filled with different compositions of MOX fuel are computed. The performance of different nuclear data libraries is evaluated. Parameters such as infinite multiplication factor, reactivity change agree very well with the averaged reference values provided by the other institutions when the JEF2.2 library is used. Inventories of major actinides and fission product nuclides at different burn-up depths are also compared with published values at the MOX pin cell and assembly level, results of the Dragon5 lattice code are in agreement with averaged values provided by other codes. Furthermore, the deviation between newer libraries and reference solutions of the benchmark should be attributed to the differences in neutron data. Therefore, the Dragon5 lattice code is reliable for neutronics and burn-up analysis of MOX fuel and can be applied to the neutronic analysis of mixed uranium-plutonium oxide fuel.

Utilization of the Thorium of the Holos-Quad Micro-Reactor Concept

In this study, the criticality and burnup analyses have been performed for Holos-Quad micro nuclear reactor as a thorium burner which has 15 ton reactor mass and 22 MWth full core geometry by using continuous-energy multi-purpose three-dimensional Serpent and OpenMC Monte Carlo particle transport code with ENDF/ B-VII data libraries. As nuclear fuel, reactor grade (RG) and weapon grade (WG) Pu mixed with ThO2 in different proportions have been used. Firstly, effective criticality for RG-Pu and WG-Pu for operation period have been determined. In addition, since each of the 4 subcritical power modules must be subcritical due to the structure of the Holos-Quad reactor, the criticality analysis of the power modules has been performed. The criticality, burnup and fissile fuel production of the Holos-Quad micro nuclear thorium-burner have been calculated with different fuel fraction depending on the fuel type. Finally, thorium mixed fuels with 35% WG-Pu or 75% RG-Pu would allow burn up levels of ~ 25 GWd/ton and ~15 GWd/ton with reactor operation periods ~3.5 years without fuel change, respectively.

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

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

Evaluation of tritium production capability of the 5 MWe gas-cooled graphite-moderated reactor at Yongbyon using tritium-producing burnable absorber rods

The 5 MWe graphite-moderated gas-cooled reactor located at Yongbyon, North Korea, is a major facility for producing nuclear materials. The objective of this study is to evaluate the tritium production capability of the 5 MWe reactor when using a tritium-producing burnable absorber rod (TPBAR). This paper presents burnup analysis results for TPBAR-embedded core models performed by McCARD, a Monte Carlo neutron transport simulation code. For three types of cores with TPBARs, the numerical results show that the 5 MWe reactor can produce 1.83–7.51 g of tritium per year depending on the number of TPBARs in the core. The numerical results can serve as a basis for assessing the feasibility of producing tritium using graphite-moderated gas-cooled reactors.

Searching for optimum design and burnable absorber for controlling the reactivity of U.S. Supper critical water reactor (SCWR)

Numerous research endeavours are currently underway to advance supercritical water reactors (SCWR), acknowledged as a cornerstone among Generation IV nuclear reactor designs. Evaluation and managing the reactivity of the reactor is a vital issue in the reactor operation. This research seeks to identify efficacious burnable absorber (BA) materials and determine an optimal spatial distribution within the fuel assembly to regulate reactivity levels effectively. Gadolinium, erbium and Lutetium have been suggested as BA in the form of integral burnable absorber (IBA) rods ((UO2 + Gd2O3), (UO2 + Er2O3) and (UO2 + Lu2O3)). Two SCWR assembly models, each featuring varied quantities and distributions of BA, have been examined. Various concentrations of the suggested BAs have been examined in the suggested models to verify the optimum cases. Burnup analyses have been conducted to evaluate the proposed cases. Different alloys of BAs including B4C+Dy2O3 and B4C+Sm2O3 have been investigated in the control rod and compared with the standard alloy B4C.

Automated Burnup Analysis from Metallic, Ceramic, and Amorphous Fuels using Atom Probe Tomography

Automated Burnup Analysis from Metallic, Ceramic, and Amorphous Fuels using Atom Probe Tomography

A Novel Adjoint-Based Reduced-Order Model for Depletion Calculations in Nuclear Reactor Physics

The licensing of new reactors implies the use of verified and validated neutronic codes. Numerical validation can rely on sensitivity and uncertainty studies, but they require repeated execution of time-consuming neutron flux and depletion calculations. The computational costs can be reduced by using perturbation theories. However, the uncoupled Depletion Perturbation Theory is restricted to single integral values such as nuclide density. Relying on reduced-basis approaches, which reconstruct all nuclide densities at once, is one way to get around this restriction. Furthermore, the adjoint-based reduced-order model uses the direct and adjoint equations for projection. For diffusion or transport calculations, the Exact-to-Precision Generalized Perturbation Theory was developed. Still, no models for depletion calculations are readily available. Therefore, this paper describes a novel adjoint-based reduced-order model for the Bateman Equation. It uses a range-finding algorithm to create the basis and the uncoupled Depletion Perturbation Theory for the reconstruction of the first order replaced by with a first order formulation. Our paper shows that for several perturbed cases, the depletion reduced-order model successfully reconstructs the nuclide densities. As a result, this serves as a proof of concept for our adjoint-based reduced-order model, which can perform sensitivity and uncertainty burn-up analysis in a shorter time.

Neutronic assessment and optimization of ACP-100 reactor core models to achieve unit multiplication and radial power peaking factor

This study focuses on the neutronic and burnup assessment of the PWR-based Chinese ACP-100 SMR core. Four different core models with varying fuel compositions and assembly configurations were analyzed using the Monte Carlo code SERPENT 2. Three of these models shared uniformly enriched fuel assemblies of 3 wt.%, 4 wt.%, and 4.45 wt.% 235U respectively, while the fourth model (the proposed design for the ACP-100 in this study) featured a mix of differently enriched fuel assemblies along with the inclusion of burnable absorbers. The burnup analysis of the cores spanned 1570 EFPDs, evaluating several neutronic parameters (beta effective, reactivity coefficients) as well as fuel cycle characteristics (cycle burnup, discharge burnup and cycle length). Changes in the isotopic concentration of fissile, fertile, and poison materials were observed and the assembly-wise linear power distribution, neutron flux spectra, and radial power peaking factors of the models were determined. In the mixed core, a combination of borated water, IFBAs and control rods were employed to achieve exact criticality (unit keff) and uniform power distribution (unit radial power peaking factor). Insertion of control rods into 41 of the 57 assemblies, in conjunction with IFBAs placed strategically, attained a near unit radial power peaking factor (PPF) of 1.0586. Finally, the combination of IFBAs, sol-bor of 936 ppm concentration, and CRs with rod worth of 414.33 pcm per assembly brought down the keff to 1.000, but increased the PPF to 1.117. While the achievement of uniform power distribution came at the cost of fuel burnup, the proposed mixed core successfully remained critical for over 1.5 years and reached a discharge burnup of 23.37 GWd/t for a two-batch refueling scheme.

Experimental burn-up analysis of ITU Triga Mark II reactor fuel elements

In this study, burn-up values ​​of spent fuels were determined using Non-Destructive Analysis (NDA) with the help of 137Cs burn-up indicator. Four spent fuel elements (F30, D18, B5 and E14) that burned in the core of the Istanbul Technical University Triga Mark II research reactor over the years were investigated with two different methods. In addition, the individual activity of the fuel elements along the axial fuel length was determined by the High Purity Germanium (HPGe) detector. As expected, it was determined that the fuel element B5, located in the center of the reactor core, exhibited the highest burn-up value with both methods. The calculations conducted using method 1 revealed a burn-up rate of 5.40% for the B5 fuel element, while method 2 yielded a burn-up value of 205.59 MWd/tHM. Experimental results were compared with previous simulation burn-up studies.

Neutronic Assessment of High-Temperature Gas-Cooled Thorium Burner using Monte Carlo Calculation Method with Full Core Model

In this study, the effective reactivity and burnup analyses have been performed for heterogeneous three-dimensional high-temperature gas-cooled thorium reactor (HTGR) which has 60 MWth full core geometry by using continuous-energy multi-purpose three-dimensional Monte Carlo particle transport Serpent code with ENDF/ B-VII data libraries. Nuclear fuel have been selected as 50 % ThO2+50% RG-PuO2. Firstly, effective reactivity for three different qualities of graphite for operation period have been determined. The effective reactivity showed better performance with increasing densities of graphite. Secondly, it has been also examined to ZrC and SiC cladding materials effect on the effective reactivity. It is observed that SiC has a positive effect on reactivity compared to ZrC. As a results, the full core life low-power thorium-burner HTGR have been calculated as up to ~4500 days depending on the graphite material whereas, the corresponding burn−ups came out to be ~ 189 GWd/ton, for end of life.

Inverted fuel geometry implementation for a fast reactor: Potential improvements in neutronic and thermalhydraulic performance

This paper presents a detailed procedure for implementing the inverted fuel geometry to a fast reactor to improve its safety system and economy. The study is starting from the fuel unit cell, fuel assembly, reactor core, and burnup analysis. A proposed multivariable graph (v, ΔP, TFmax–Dco, VF) introduced at the fuel unit cell level provides comprehensive thermalhydraulic and neutronic parameters in a single graph, allowing for an efficient optimization process. The fuel unit cell study reveals that the inverted fuel design has a higher fuel volume fraction and lower core pressure drop than conventional pin-typed fuel. This is beneficial for the reactor economy and enhances the reliability of the safety system. With the inverted fuel design, the primary loop can save pumping power by up to 40 % and provides an excess driving force for natural circulation. The male–female axial grid structure separating the fuel assembly potentially eliminates coolant flow path restriction and fretting issues. The core is named the Inverted Core Fast Reactor (IC-FR), an LBE-cooled fast SMR designed to generate 60 MWth for at least 40 years of full-power operation without refueling and fuel shuffling. IC-FR is a transportable reactor and has load following capability that can be deployed for many applications, including marine and land-based applications, and stand alone or mixing power grid. The burnup study of IC-FR reveals that the balance of neutron leakage and fissile inventory yields a small reactivity swing (<1$) for 40 years. This study extensively utilizes Monte Carlo (MC) code MCS for neutronic calculation. Owing to the high computational expense of MC code, the approaches to optimize the MC usage are also presented.

Performance evaluation of a currently in-use dry storage cask design for spent accident tolerant fuel loading case under normal operation condition

Abstract The aim of this study is to assess the performance of an existing dry storage cask design for accident tolerant fuel loading case and to examine the compliance with the safety limits applied currently for dry storage. In the study, for a dry storage cask design currently in use, criticality calculations, dose rate evaluation and thermal analyses are performed in case of loading with the accident tolerant fuel discharged from a PWR. Firstly, for an accident tolerant fuel selected among the concepts proposed for use in light water reactors, burnup analyses are performed by utilizing the Serpent continuous energy code and spent fuel characteristics are determined. Then, criticality analyzes are carried out by using the Serpent Monte Carlo code for the case of loading the accident tolerant fuel into the selected dry storage cask design. Gamma and neutron dose rates at the outer surface and close distances of the storage cask are determined with the Serpent code. To evaluate the thermal performance of the storage cask, thermal analyzes are performed by using the ANSYS Fluent computational fluid dynamics code. The analysis results are compared with the nuclear safety criteria applied to dry storage casks. Results of the analysis show that the dry storage cask design currently in-use does not exceed the criticality, dose rate and maximum surface temperature limits when loaded with spent accident tolerant fuel.

Machine learning approaches to equilibrium burnup analysis for Molten Salt Reactor

During the operation of a liquid-fueled molten salt reactor (MSR), the deep depletion to an equilibrium state can be attained with online refueling and reprocessing techniques. However, the equilibrium burnup simulation of an MSR using current depletion codes requires considerable computational cost and time, and even fails when the equilibrium conditions are unsatisfied. To efficiently evaluate the specific characteristics of the equilibrium burnup in MSRs, a fast-filtering model for equilibrium state and a fast prediction model for equilibrium neutronic properties are developed based on the machine learning (ML) technique. A database of 4860 cases including the neutronic information of equilibrium and non-equilibrium of a fuel assembly with variable geometric parameters in an MSR is first generated, and then the neutronic properties of 232Th-233U, 232Th-EU (enriched uranium) and 232Th-Pu fuel cycles are evaluated. Based on the MSR database, fifteen different ML classification algorithms are implemented to establish the fast-filtering model for equilibrium burnup state. Moreover, a model for the fast prediction of equilibrium neutronic parameters of neutron flux (Φ) and breeding ratio (BR) is developed with fifteen ML regression algorithms. Considering the various performance metrics for measuring the predicted performances of ML models in the classification and regression, the LightGBM (LGBM) model is the most favorable for filtering the burnup state and predicting the neutronic parameters in MSRs. With the test data, the LGBM model can obtain the accuracy of about 0.98 for classification and the relative error of about 1.10% for regression.

Validation and application of the Dragon5 lattice code for neutronics and burnup analysis of VVER-1000 pin cell and assembly model

This work deals with the validation and assessment of neutronics and burnup analysis capacity of the deterministic lattice code Dragon5 for VVER-1000 hexagonal pin and assembly model configuration. Various VVER-1000 hexagonal pin-cell benchmark cases including the fresh uranium fuel and mixed oxide fuel of ORNL are simulated and compared with the HELIOS and OpenMC code. The data library differences of certain nuclides will lead to large deviations of the results. Accurate results can be presented by the Dragon5 lattice code when compared with OpenMC code. The OECD LEU and MOX assembly benchmarks are computed to further validate the prediction ability for VVER-1000 assembly, reasonable agreement between Dragon5 solutions and published values of other codes is observed for burnup parameters including infinite multiplication factors, concentrations of important actinides and fission product nuclides. Moreover, the post-irradiation of UO2 sample experiment of the Russian Kalinin-1 VVER-1000 reactor is simulated and compared with laboratory measured values to provide the experiment validation of the Dragon5 lattice code. The predicted concentrations of actinides up to Cm244 are compared with HELIOS and Serpent Monte Carlo code. The results show that the Dragon5 lattice code presents better prediction accuracy for Calculated-to-Experimental ratios (C/E's) of most actinide concentrations than the HELIOS and Serpent code. Therefore, the École Polytechnique de Montréal Dragon5 lattice physics code can be applied to the neutronics and burnup analysis of VVER-1000 hexagonal pin-cell and assembly model.

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 temperature and hydrogen content on the transverse mechanical properties of Zircaloy-4 fuel cladding

This paper presents the results of the investigation on the influence of hydrogen on the transverse mechanical properties of the Zircaloy-4 fuel cladding at various temperatures ranging from room temperature to 600οC. Data pertaining to the mechanical properties of the fuel cladding in the hoop direction are required for the analysis of high burn-up fuel behaviour under loss of coolant accident (LOCA) and reactivity initiated accident (RIA) conditions. Using the ring tension test method, the mechanical properties of the Zircaloy-4 fuel cladding in the transverse direction were evaluated as functions of hydrogen concentration and test temperature. The effect of hydrogen on the mechanical properties was observed to be marginal but the effect of temperature was significant in the investigated hydrogen concentration range (47–475 ppmw).

A feasibility study of SMART reactor power performance optimizations-part 1: Steady-state and burn-up analysis

SMART is an integral small pressurized water nuclear reactor design with a rated power output of 100 MWe from 330 MWth, but it needs a higher power output for the United Kingdom energy market. This study applies Monte Carlo code OpenMC to build a full-core model and innovatively adjust the simulation coefficients to approach the reactor operating conditions. The analysis results point out the reasonable optimization’s technical direction. The model’s sensitivity to ENDF and JEFF nuclear data libraries and spatial division is tested and verified. Then it performs a series of simulations to obtain the core’s neutronic parameters, such as neutron energy and spatial distributions, effective neutron multiplication factor keff and its variation versus depletion. The analysis found that the initially designed core’s keff is 1.22906, and the temperature reactivity defect is 11612 pcm. In 1129 full-power operating days, the keff will decrease to 0.99126, and the reactor depletes 8.524 × 1026235U atoms. However, the outermost fuel assemblies’ 235U depletion rate is lower than 45% in this extended refuelling cycle, and their ending enrichment is higher than 2.4%. That means the fuel economy of the original design’s two-batch refuelling scheme core layout is insufficient. Improving the thermal neutron fluence in these assemblies may optimize the SMART power performance effectively.

Development of Linkage Program Code OpenMC and ORIGEN2.2 for Neutronic Analysis and Burnup Nuclear Reactor Program

This paper discusses the development of program code that coupling the ORIGEN2.2 for burn-up analysis with the Monte Carlo program OpenMC for neutron analysis program called OpenMC-ORIGEN (Op-OR). The results of the program have been compared with benchmark results from previously published results and MCNP6 program. The acquired results show a good agreement with benchmark results. This linkage program codes performed well for designing a new advanced reactor and analysing the neutronics parameter and burnup/depletion calculation.

CORE BURN-UP ANALYSIS OF THE RSG-GAS RESEARCH REACTOR USING DETERMINISTIC AND STOCHASTIC METHODSS

Due to several characteristics, such as geometry, compact core, high coolant flow, and high neutron flux, the burn-up study of the RSG-GAS multi-purpose reactor provides challenges when employing a neutronic calculation. For the burn-up analysis, two calculating methodologies are used in the RSG-GAS: deterministic and probabilistic methods. The deterministic codes such as WIMSD-5B and Batan-FUEL are utilized, whereas the continuous-energy Monte Carlo code Serpent 2 is used for the stochastic method. WIMSD-5B is being used to produce a four-group cross-section that is needed by Batan-FUEL to do full core diffusion calculations. Burn-up calculations were performed at the whole fuel assemblies inside the core to see if the deterministic code, WIMSD-5B/Batan-FUEL, could accurately replicate the burn-up behavior of the RSG-GAS research reactor. The Serpent 2 calculation was also done with the exact models to provide a comparison. The results show that both Serpent 2 and WIMSD-5B/Batan-FUEL can perform the RSG-GAS burn-up analysis if appropriate treatments are made to the deterministic codes at both the assembly and core levels. There is a 5% difference in calculated fuel burn-up between deterministic and stochastic approaches.

Neutronic Evaluation of Light Water Reactor-Based Small Modular Reactor With Transuranic Fuel and Comparison With Different Core Configuration of UO2 and Mixed Oxide Fuel

Abstract Neutronic and burnup assessment of a light water reactor (LWR)-based small modular reactor (SMR) core with transuranic fuel (TRU) was carried out and compared with UO2 fuel and mixed oxide fuel (MOX) for different core composition and configurations. Eight different cores were modeled by changing the fuel composition and loading pattern of four different fuel assemblies, namely, UO2, MOX, TRU, and annular UO2 fuel and simulated in open-source Monte Carlo code OpenMC. The radial neutron flux distribution and maximum to average flux ratio were studied. The lowest flux ratio was found in annular-2 models, whereas TRU models showed the highest value. The effect of enrichment, fuel burnup, moderator to fuel ratio, and soluble boron concentration on effective multiplication factor (keff) was investigated. Burnup analysis of fissionable materials and neutron poison was performed. Burnup-dependent fission rate of U-235, Pu-239, and Pu-241 was also studied. The composition of spent fuel at end of the cycle (EOC) was investigated and lower minor actinides (MAs) production was observed for TRU models. This analysis would help to determine the effectiveness and feasibility of these potential core configurations and to predict their neutronic behavior.