Burnup Code Research Articles

Online inventory modeling of a CANDU-6 reactor for nuclear forensic applications

To develop the expected isotopic signatures of online nuclear power plants, a CANDU-6 quarter-core reactor was modeled using Serpent, a Monte Carlo and Burnup code. The reactor model was simulated using nominal operating parameters, steady power levels and standard refueling procedures, to set the baseline for online operations. The model was burned for 500 refuelings totaling 1400 effective full power days. The core was divided into 1140 spatially-discrete fuel bundles with each tracking the density of 237 isotopes. Instantaneous core inventory snapshots were recorded at the time of each refueling to create a continuous inventory database. These snapshots provide the expected isotopic densities and ratios for virtually any fuel bundle position or burnup under nominal operating parameters. These values are useful in the event of accidents, short-cycles, or nuclear proliferation. The time-dependent and spatially dependent results for xenon effluent are used to develop an analytical method for calculating the expected International Monitoring System xenon ratio measurements based on fuel bundle leak rates. A possible false-positive nuclear proliferation scenario for a CANDU-6 operating under nominal parameters is also identified.

Preliminary study of transuranic transmutation in a small modular chloride salt fast reactor

The management of radioactive waste poses a significant challenge to the sustainable development of nuclear energy. Efficient transmutation of nuclear wastes is crucial to minimize their accumulation. A small modular chloride salt fast reactor (sm-MCFR) capable of transmuting transuranic elements (TRU) is proposed in this paper, combining the advantages of the small modular reactor (SMR) and the molten salt reactor (MSR). The sm-MCFR is characterized by a high fuel loading and a compact core structure that can be quickly deployed around large commercial reactors to achieve TRU transmutation. To evaluate the TRU burnup capability of the sm-MCFR, several fuel salts and reprocessing modes were analyzed using the internally developed TRITON MODEC Coupled Burnup Code (TMCBurnup) tool. NaCl-MgCl3 with 98 % enrichment in 37Cl is chosen as carrier salt for the sm-MCFR, which can achieve 76.7 % TRU transmutation rate in average and 355 kg·GW−1·a−1 TRU transmutation quality at a continuous reprocessing rate of 10 L/d for 50 operation years. The optimized sm-MCFR reduced the radioactive toxicity of TRU by 84 %, thereby simplifying waste reprocessing. In addition, the sm-MCFR has a negative temperature feedback coefficient of −7.195 pcm/K, favoring safe reactor operation.

Deterministic neutron transport and burnup code coupling assessment

Nuclide evolution is a key aspect in nuclear reactor design and operation. Burnup (or depletion) codes evaluate the isotopic inventory evolution. The inherent need for nuclear safety, the development in computational capabilities, and new discoveries in related areas, such as mathematical methods, result in constant software re-evaluation and improvement. Now, many stochastic-based burnup codes are being developed in comparison to deterministic-based ones, which also present advantages and interesting aspects. In this work, nuclear libraries, processing requirements, data workflow, and methodologies are studied for the development of a brand-new burnup code linked to the VALKIN-FVM-Sn deterministic transport code. As a result, the requirements to create, and couple, a burnup code have been assessed, establishing a methodology. The burnup code has been developed and contrasted with an industry reference. The resulting coupling is a lattice code foundation to be used in a Multiphysics platform.

Neutronic and burn up analysis of VVER 1000 benchmark using EXCEL and TRIHEX-FA code system

Benchmarking and validation of the neutronic codes is an important aspect of the reactor design. This paper presents VVER-1000 X2 Benchmark steady state core physics analysis carried out using the lattice level burn-up code, EXCEL and core physics analysis using multi-group diffusion code, TRIHEXFA developed at BARC. These codes have been augmented with new features for modeling of LWR Cores. The VVER-1000 X2 Benchmark with modern fuel assembly (FA) design is a comprehensive benchmark proposed for validating and verifying package of codes and data libraries for reactor physics calculations of LWRs, which includes fuel assembly modeling, generation of homogenized cross section data base for all envisaged core conditions, and reactor core modeling. The trends of k∞ (k-inf) values as a function of various parametric variations and as a function of burn-up for all the fuel assemblies considered in benchmark are analyzed in detail and the deviations have been estimated.The VVER-1000 core follow-up simulations are done by using core diffusion code TRIHEXFA using the few group data generated by EXCEL code. Important core physical characteristics like critical soluble boron, 2-D radial power distribution, axial power distribution, radial peaking factors and volumetric power peaking factors, FA burn-up values obtained from the TRIHEXFA simulations are compared with the experimental core follow up data. All these parameters are found to be reasonably in good agreement with the values provided in benchmark.

Fluence calculations for the TRISO- particle fuel ION implantation experiment

The TRistructural ISOtropic (TRISO) nuclear fuel is commonly used in High-Temperature Reactors (HTRs). The characteristic feature of this fuel is the presence of four covering layers. Due to the TRISO-particle fuel irradiation in the reactor core, damage in the coating layers might occur. The investigation of the damage occurring in the TRISO nuclear fuel is a key factor for the failure-free operation of HTR reactors. To determine the level of damage in the TRISO fuel, the ion implantation technique may be used. This method allows to quickly simulate the fuel irradiation phenomenon in the reactor core, through parameters such as displacements per atom (DPA). In order to determine the level of damage to the sample that corresponds to the final neutron flux of the HTR reactor, it is first necessary to determine the fluence with which the ion implantation experiment should be performed. This work aims to determine the relationship between ion fluence dose that corresponds to the neutron fluence in the HTR reactor generating similar radiation damages at the TRISO layers. For that purpose, the authors performed detailed calculations using Stopping and Range of Ions in Matter (SRIM) and Monte Carlo Continuous Energy Burn-up Code (MCB) codes. The obtained results indicate that the fluence of 3.8*1016 ions/cm2 corresponds to the neutron flux of 1.6*1013 neutrons/cm2s over one year of constant irradiation.

Validation of the burnup code MOTIVE with respect to fuel assembly decay heat data

The burn-up code MOTIVE is a 3D code for fuel assembly inventory determination developed at GRS in recent years. It modularly couples an external Monte Carlo neutron transport code to the in-house inventory code VENTINA. In the present publication, we report on the validation of MOTIVE with respect to full-assembly decay heat measurements of light water reactor fuel. For this purpose, measurements on pressurized water reactor and boiling water reactor fuel assemblies from different facilities have been analyzed with MOTIVE. The calculated decay heat values are compared to the measured data in terms of absolute and relative deviations. These results are discussed and compared to other published validation analyses. Moreover, the observed deviations between measurements and calculations are analyzed further by taking into account the results of the validation of nuclide inventory determination with MOTIVE. The influence of possible biases of calculated nuclide densities important to decay heat at the given decay times are investigated and discussed.

Inventories of Short-Lived Fission Gas Nuclides in Nuclear Reactors

Taking inventories in reactor cores is critical for understanding their radioactive source terms and establishing the relationship between the activity concentration in the primary loop and the status of the reactor core’s fuel. However, there is a niche in which a simple but accurate relationship between reactor conditions and nuclide inventories can reliably predict the fission gas nuclide activities of the reactor core in the primary loop. In this study, a simple and efficient model called “Inventories of a Point Reactor for Fission Gas Nuclides” (IPRFGN) was proposed to calculate and interpret such inventories, in which a 10 MW high-temperature gas-cooled experimental reactor (HTR-10) was used as the test case. The present study findings were consistent with those of a general point–depletion burnup code such as the KORIGEN code. Here, the relative error was <1%. Based on the application of the IPRFGN model in HTR-10, the results indicate that the proposed IPRFGN model has provided the relationship between the inventories of fission gas nuclides in the core and the reactor conditions in all types of nuclear fission reactors. In the future, the IPRFGN model will be used for calculating fission gas nuclide inventories in various reactors.

Advances on the Development of a Deterministic Based Burnup Code. Code Coupling and Data Processing Assessment

Advances on the Development of a Deterministic Based Burnup Code. Code Coupling and Data Processing Assessment

Inter-Code Comparison of Computational VERA Depletion Benchmark Using OpenMC, OpenMC-ONIX and DRAGON

This research focuses on the comparative analysis of the PWR fuel assembly based on VERA depletion benchmark problems using community-developed open source Monte Carlo code OpenMC, python based burnup code system ONIX (a coupling interface for Monte Carlo code OpenMC), and deterministic DRAGON code. The depletion analysis was performed using OpenMC and ONIX with ENDF/B-VII.1 nuclear data library, and DRAGON with SHEM-361 based DRAGLIB format library (ENDF/B-VII.1). The code-to-code analysis on the evolution of k eff , atom number density, and power distribution as a function of burnup has been performed and the result shows a good agreement with the maximum difference within 200 pcm at EOC. However small discrepancy around 90 pcm has been observed in k eff calculated by DRAGON compared to OpenMC in the presence of integral fuel burnable absorbers (IFBA). The above-mentioned codes have been validated successfully for the first time against PWR fuel assembly based on VERA depletion benchmark problems. It can be concluded that initial implementation of these codes at the Department of Nuclear Science and Engineering under Military Institute of Science and Technology, Dhaka, was successful and that further research works are to be performed to utilize these codes for depletion/neutronics calculation of existing 3MW TRIGA Mark-II research reactor and VVER-type power reactor that is to be commissioned in Bangladesh.

The Application of Radiochemical Measurements of PWR Spent Fuel for the Validation of Burnup Codes

The paper shows the usage of destructive assay data from PWR fuel samples for the validation of the NFCSS burnup system developed by the IAEA. The results of radiochemical measurements of spent fuel isotopic composition were compared with the results of numerical modelling. In total, 254 samples from 15 PWRs, incorporated in the SFCOMPO database, were applied in the validation study. The paper shows the trends observed in the calculated-to-experimental ratios for eight major actinides and six minor actinides present in spent nuclear fuel. The data were quantified and analysed considering the enrichment, burnup and origin. The obtained results were compared with other studies on burnup validation using different numerical tools. In general, the results of numerical modelling for major actinides show rather good consistency with radiochemical measurements, while the results for minor actinides are less accurate.

Assessment of the Radiotoxicity of Spent Nuclear Fuel from a Fleet of PWR Reactors

The paper presents the methodology for the estimation of the long-term actinides radiotoxicity and isotopic composition of spent nuclear fuel from a fleet of Pressurized Water Reactors (PWR). The methodology was developed using three independent numerical tools: the Spent Fuel Isotopic Composition database, the Nuclear Fuel Cycle Simulation System and the Monte Carlo Continuous Energy Burnup Code. The validation of spent fuel isotopic compositions obtained in the numerical modeling was performed using the available experimental data. A nuclear power embarking country benchmark was implemented for the verification and testing of the methodology. The obtained radiotoxicity reaches the reference levels at about 1.3 × 105 years, which is common for the PWR spent nuclear fuel. The presented methodology may be incorporated into a more versatile numerical tool for the modeling of hybrid energy systems.

Serpent and Monte Carlo Continuous Energy Burnup Code Comparison for Neutronics Calculations of the European Lead-Cooled Fast Reactor

Abstract As part of the generation IV reactors (GENIV), the European lead-cooled fast reactor (ELFR) is one of the most promising candidates to be part of the European energy framework in the near future. Alike most of the GENIV systems, the ELFR is still under development and having reliable computational tools that allow fast and accurate results become an important task in the modeling and simulation levels. In this work, the Serpent code is assessed. Serpent is a continuous-energy Monte Carlo (MC) code suitable for reactor physic calculations and is used for the modeling of the ELFR system. The results were compared with the reference data that were obtained with the Monte Carlo continuous energy burnup code (MCB). In order to verify the ELFR Serpent model, several neutronic parameters were compared to the reference results: the effective neutron multiplication factor (keff), the Doppler constant (KD), the reactivity effect of the coolant density, the effective delayed neutron fraction (βeff), and the effective prompt neutron lifetime (Λ). In addition, the axial and radial power distributions were also obtained and verified. A good approximation between Serpent and MCB values was obtained.

MPACT VERIFICATION WITH MAGNOX REACTOR NEUTRONICS PROGRESSION PROBLEMS

MPACT is a state-of-the-art core simulator designed to perform high-fidelity analysis using whole-core, three-dimensional, pin-resolved neutron transport calculations on modern parallel computing hardware. MPACT was originally developed to model light water reactors, and its capabilities are being extended to simulate gas-cooled, graphite-moderated cores such as Magnox reactors. To verify MPACT’s performance in this new application, the code is being formally benchmarked using representative problems. Progression problems are a series of example models that increase in complexity designed to test a code’s performance. The progression problems include both beginning-of-cycle and depletion calculations. Reference solutions for each progression problem have been generated using Serpent 2, a continuous-energy Monte Carlo reactor physics burnup calculation code. Using the neutron multiplication eigenvalue ke_ as a metric, MPACT’s performance is assessed on each of the progression problems. Initial results showed that MPACT’s multigroup cross section libraries, originally developed for pressurized water reactor problems, were not sufficient to accurately solve Magnox problems. MPACT’s improved performance on the progression problems is demonstrated using this new optimized cross section library.

Monte Carlo methods for numerical simulations of the Lead-Cooled Fast Reactor

Closed nuclear fuel cycle is one of the most promising options for the efficient use of nuclear energy resources with full recycling of long-lived transuranic elements. However, it can be implemented only in fast breeder reactors of Generation IV. The paper shows our methodology applied in the analysis of the lead-cooled fast reactor equilibrium fuel cycle using the Continuous Energy Monte Carlo Burnup Code – MCB. The implementation of novel modules for nuclear transmutation trajectory folding allows us to trace the life cycle of crucial minor actinides from the beginning of the reactor’s life towards the state of adiabatic equilibrium. Changes in the mass contribution to gateway isotopes 242Pu, 243Am and 244Cm during 124.2 years were considered in the study. Numerical demonstration was performed for the reactor core designed within the European Lead-cooled System (ELSY) project and redefined in the follow-up Lead-cooled European Advanced Demonstration Reactor (LEADER) project.

SFCOMPO DATABASE OF SPENT NUCLEAR FUEL ASSAY DATA – THE NEXT FRONTIER

SFCOMPO is the world’s largest database for measured spent nuclear fuel assay data. An international effort coordinated by the Nuclear Energy Agency (NEA) resulted in a significant expansion of the database and its release online in 2017 as a downloadable application. The SFCOMPO Technical Review Group (TRG) was recently formed under the direction of NEA’s Nuclear Science Committee/Working Party on Nuclear Criticality Safety and was mandated to maintain and further coordinate the development of SFCOMPO. This TRG is currently focused on (1) critical evaluation of the experimental assay data by independent experts and (2) development of benchmarks and benchmark models that can be applied to validate burnup codes. This will improve the quality and documentation of the experimental datasets and enable their use by the international community to support code validation for design and safety analysis of spent nuclear fuel transportation, storage, and repository applications. It follows the precedent and draws on the experience gained from similar NEA efforts in the International Reactor Physics Experiment Evaluation Project and the International Criticality Safety Benchmark Experiment Project. Ongoing SFCOMPO evaluations have served as a test bed to develop templates for documenting evaluations, develop review guidance, improve approaches for a global uncertainty analysis, and devise a strategy focused on providing practical information of highest value to the user community. The current effort, status, and associated challenges are discussed.

Monte Carlo Investigation of the UK’s First EPR Nuclear Reactor Startup Core Using Serpent

Computationally modelling a nuclear reactor startup core for a benchmark against the existing models is highly desirable for an independent assessment informing nuclear engineers and energy policymakers. For the first time, this work presents a startup core model of the UK’s first Evolutionary Pressurised Water Reactor (EPR) based on Monte Carlo simulations of particle collisions using Serpent 2, a state-of-the-art continuous-energy Monte Carlo reactor physics burnup code. Coupling between neutronics and thermal-hydraulic conditions with the fuel depletion is incorporated into the multi-dimensional branches, obtaining the thermal flux and fission reaction rate (power) distributions radially and axially from the three dimensional (3D) single assembly level to a 3D full core. Shannon entropy is quantified to characterise the convergence behaviour of the fission source distribution, with 3 billion neutron histories tracked by parallel computing. Source biasing is applied for the variance reduction. Benchmarking the proposed Monte Carlo 3D full-core model against the traditional deterministic transport computation suite used by the UK Office for Nuclear Regulation (ONR), a reasonably good agreement within statistics is demonstrated for the safety-related reactivity coefficients, which creates trust in the EPR safety report and informs the decision-making by energy regulatory bodies and global partners.

The PyNE-Based Burnup Analysis Method for Accelerator-Driven Subcritical Systems

To study the burnup features of accelerator-driven subcritical systems (ADSs), simplified transmutation trajectories are imperative to make the simulation process more effective with acceptable precision. This process has long been considered a challenging task since the construction of simplified burnup chains often need complex judgments and experiences. Additionally, the burnup analysis of ADSs requires more specific burnup chains for some important isotopes with minor actinides (MAs) and long-lived fission products (LLFPs) included. However, some general burnup codes lack these chains or pack some particularly important isotopes into a kind of pseudo nuclide. In this context, a PyNE-based burnup module (PyNE-Burn) has been developed to solve the burnup problem in ADSs, where three types of isotopes have been considered to construct the simplified burnup chains and weight-sorted judgment criteria have been proposed to determine which nuclides should be included. Moreover, the scan-mode-method-based high-order differential expression has been employed to substitute the legacy method in solving the linearized burnup chains. Finally, numerical tests have been carried out to demonstrate that the PyNE-Burn module has acceptable accuracy and can be used in dealing with the burnup problem in ADSs.

The impact of gadolinium on the reactor production of 153Sm

Radioisotopes represent major sources of ionizing radiation, not least for use in medical applications, brachytherapy and nuclear medicine included. In this, the nuclear reactor is the main source of β- - γ emitting isotopes, an example product being 153Sm used in the treatment of pain arising from bone metastases. Present analysis relates to the potential of gadolinium neutron capture reaction, its impact on reactor production of radioisotopes and the proliferation resistant potential of thorium fuel cycle. A comparative analysis has been made of the impact of gadolinium on the production of 153Sm by UO2 and (Th, U) O2 fuels in a Westinghouse small modular reactor. Five fuel assemblies were investigated: one containing no gadolinium, the other four containing 16, 24, 34 or 44 gadolinium fuel rods. The code Monte Carlo N-Particle eXtended (MCNPX) integrated with the CINDER90 burn-up code was used for calculations. In the production of 153Sm the same trend is followed for the fuels containing gadolinium, increasing significantly with the number of gadolinium fuel rods. Zero production results from fuel assemblies without gadolinium. The concentration of 153Sm increases significantly with burn-up, indicating that gadolinium has a positive impact on the production of 153Sm.

Neutronics conceptual research on a hybrid blanket of china fusion engineering test reactor

• A hybrid CFETR blanket using natural uranium U-10Zr alloy and water is proposed. • Multi-layer alternate arrangement of fission zone and tritium breeding zone is used. • 3D detailed neutronics model is built for JMCT by CMGC. • MCNP results are crosschecked by JMCT. • TBR and M are 1.26 and 3.18 respectively. The engineering concept research of China Fusion Engineering Test Reactor (CFETR) is underway. In order to achieve tritium self-sufficiency and lower tritium startup inventory, a fusion fission hybrid blanket scheme is proposed in this paper as a backup for the traditional pure fusion blanket concepts. In the hybrid blanket design, natural uranium is used as fuel, light water as coolant and Li 4 SiO 4 as tritium breeder. MCORGS, a burnup code coupled by MCNP and ORIGENS, is used in burnup calculation. JMCT, a Joint Monte Carlo Transportation code developed by Institute of Applied Physics and Computational Mathematics, is used to crosscheck the MCNP results. One dimensional design and optimization is firstly made to propose a multi-layer alternate arrangement of fission zone and tritium breeding zone so as to maximum Tritium Breeding Ratio (TBR) and a moderate energy Multiplication (M). A 3D neutronics model of CFETR is converted from MCNP benchmark files to a GDML file which is used in JMCT. At the beginning of the core, TBR 3D is 1.26 which is bigger than the max achievable TBR for pure fusion blanket (around 1.15), M is 3.18 which means the tritium startup inventory will be 3 times lower than the pure fusion blanket in case of a fixed blanket thermal power. After 12 years burnup with 200 MW fusion power, TBR and M will be 1.28 and 4.05 respectively. The hybrid system is in an ultra deep subcritical state, the effective multiplication constant varies from 0.161 to 0.227. If the natural uranium in fission zone is replaced by spent fuel or LEU, more higher TBR and M will be obtained.

Development and validation of Burn-up Calculation Code IMPC-Burnup2.0 for accelerator-driven sub-critical system

Fuel depletion analysis is important for the evolutionary behavior of nuclear materials in the reactor. For the accelerator-driven sub-critical system (ADS), the main research focuses on the transmutation of minor actinides (MA) and long-lived fission products (LLFP). To perform nuclide inventory calculation of ADS, a new nuclide point-depletion code IMPC-Depletion has been developed by utilizing Chebyshev Rational Approximation Method (CRAM) and Transmutation Trajectory Analysis method (TTA) to solve the Bateman equation. Besides, a new burnup code system named IMPC-Burnup2.0 which couples the OpenMC neutron transport code and IMPC-Depletion is also developed to perform burnup behavior analysis of ADS sub-critical system. Finally, the decay of Np237 and fixed neutron flux irradiation of 3.4% UO2 are calculated and compared with other reference results to verify IMPC-Depletion. Moreover, IAEA-ADS and OECD/NEA ADS Minor Actinide Burner benchmark are calculated using IMPC-Burnup2.0 to testing the capability for burnup analysis for ADS sub-critical core. Numerical results show that IMPC-Burnup2.0 is reliable when compared with reference values and it can be used to solve burnup problem of ADS.