Monte Carlo Continuous Energy Burnup Research Articles

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.

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.

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.

Numerical design of thorium and uranium fuel samples irradiation in lead environment

The paper shows capabilities of thorium-lead fuel assembly for design of irradiation experiments on ThO2 and natural UO2 fuel samples using radioisotope neutron source. The main purpose of the current analysis was to determine the irradiation environment in the samples, especially: neutron spectrum, power, activity, reaction rates, production of 233Pa and 239Np as well as breeding of 233U and 239Pu. An advanced three-dimensional numerical model for Monte Carlo radiation transport and burnup simulations was developed using the Monte Carlo Continuous Energy Burnup Code (MCB). The versatility of the assembly gives a perfect opportunity to perform many irradiation experiments for R&D on the thorium and uranium fuel cycle in a different material and geometrical environments.

A criticality study on the LA-1 accident using Monte Carlo methods

The paper presents a criticality analysis of the first fatal nuclear accident with unexpected criticality of the fissionable system. The operator of the experiment, Harry Daghlian, received a radiation dose of about 5.1 Sv and died 28 days later due to acute radiation syndrome. The paper features anumerical reconstruction of the experimental set-up and the environment in the laboratory at the time of the accident, based on publicly available documentation. The Monte Carlo Continuous Energy Burnup Code was used in all criticality calculations. The study is a follow-up of the previous criticality analysis of the Louis Slotin accident (LA-2), in which the same plutonium core was used in a fatal experiment. In the paper, I present the numerical model developed, the influence of the system components on the criticality state, the approach to criticality in function of the tungsten-carbide reflector mass, as well as abenchmarking study using JEFF3.1 and ENDF/B-VII.1 nuclear data libraries.

Thorium based fuel loading pattern for PWR reactor

The Monte Carlo Continuous Energy Burnup Code was used for numerical simulations of the pressurized water reactor (PWR) with mixed U-Th fuel. The thorium fuel was introduce to the core of the Westinghouse 4-loop PWR in order to investigate the time evolutions of fissile 233U, 235U, 239Pu, 241Pu and of fertile 232Th and 238U. The core loading pattern was designed using the Whole Assembly Seed and Blanket (WASB) approach. The following calculations show the feasibility of the proposed incore fuel management patter using the available numerical tools, developed numerical model and methods.

Neutronic analysis of mixed thorium-uranium fuel bundle for CANDU reactors

The paper shows neutronic analysis of the CANDU (CANada Deuterium Uranium) nuclear reactor fuel channel with mixed thorium-uranium fuel bundles. The numerical model of the fuel channel was designed using The Monte Carlo Continuous Energy Burn-up Code – MCB developed at the AGH University of Science and Technology, Faculty of Energy and Fuels, Department of Nuclear Energy. The super-computer Prometheus available in the frame of the Pl-Grid Infrastructure at the Academic Computer Centre Cyfronet AGH was used for multi-scale calculations. The fuel bundles are composed of two clusters of fuel rods. The neutronic analysis considers detailed numerical simulation of neutron transport in fully heterogeneous geometry of the fuel channel. Moreover, burnup simulations were performed using Transmutation Trajectory Analysis method implemented in the MCB code. In the analysis we mainly consider time evolutions of neutron multiplication factor, fissile 233U, 235U and 239Pu and fertile 238U and 232Th. The simulations were performed for eight scenarios with various fuel composition.

Monte Carlo modelling of Th-Pb fuel assembly with californium neutron source

Abstract This paper describes the methodology developed for the numerical reconstruction and modelling of the thorium-lead (Th-Pb) assembly available at the Department of Nuclear Energy, Faculty of Energy and Fuels, AGH University, Krakow, Poland. This numerical study is the first step towards integral irradiation experiments in the Th-Pb environment. The continuous-energy Monte Carlo burnup (MCB) code available on supercomputer Prometheus of ACK Cyfronet AGH was applied for numerical modelling. The assembly consists of a hexagonal array of ThO2 fuel rods and metallic Pb rods. The design allows for different arrangements of the rods for various types of irradiations and experimental measurements. The intensity of the fresh neutron source intended for integral experiments is about 108 n/s, which corresponds to the mass of about 43 μg 252Cf. The source was modelled in the form of Cf2O3-Pd cermet wire embedded in two stainless steel capsules.

Criticality analysis of the Louis Slotin accident

Abstract In this paper, I present a criticality study on the Louis Slotin accident, which happened in Los Alamos Scientific Laboratory (LANL) on 21 May 1946. For the numerical reconstruction of the nuclear system, I used the Monte Carlo Continuous Energy Burnup Code (MCB) developed at the AGH University, Krakow, Poland (Akademia Gorniczo-Hutnicza w Krakowie). I present the influence of the environment on the system criticality in the laboratory at the time of the accident. I consider locations, geometry and material composition of the elements forming the nuclear system: the plutonium core, the beryllium reflector, the human body. The numerical approach consists of three steps. Firstly, the isotopic composition of the core is estimated using the criterion of 10 cents excess reactivity achieved during the accident. Secondly, the effective neutron multiplication factor in the function of the Be hemisphere angle above the Pu core is shown. Lastly, the influence of each system component on Keff is calculated. Additionally, the influence of the position of Slotin’s hand on the criticality and the neutron spectrum in the core is presented. The study fills the gap in the numerical reconstruction of early criticality accidents and thus helps to preserve critical nuclear knowledge for future generations.

Development of numerical models for Monte Carlo simulations of Th-Pb fuel assembly

The thorium-uranium fuel cycle is a promising alternative against uranium-plutonium fuel cycle, but it demands many advanced research before starting its industrial application in commercial nuclear reactors. The paper presents the development of the thorium-lead (Th-Pb) fuel assembly numerical models for the integral irradiation experiments. The Th-Pb assembly consists of a hexagonal array of ThO2 fuel rods and metallic Pb rods. The design of the assembly allows different combinations of rods for various types of irradiations and experimental measurements. The numerical model of the Th-Pb assembly was designed for the numerical simulations with the continuous energy Monte Carlo Burnup code (MCB) implemented on the supercomputer Prometheus of the Academic Computer Centre Cyfronet AGH.

Reliability assessment of MVP-BURN and JENDL-4.0 related to nuclear transmutation of light platinum group elements

The Apres ORIENT research program, as a concept of advanced nuclear fuel cycle, was initiated in FY2011 aiming at creating stable, highly-valuable elements by nuclear transmutation from ↓ssion products. In order to simulate creation of such elements by ( n , γ) reaction succeeded by β − decay in reactors, a continuous-energy Monte Carlo burnup calculation code MVP-BURN was employed. Then, it is one of the most important tasks to con↓rm the reliability of MVP-BURN code and evaluated neutron cross section library. In this study, both an experiment of neutron activation analysis in TRIGA Mark I reactor at University of California, Irvine and the corresponding burnup calculation using MVP-BURN code were performed for validation of the simulation on transmutation of light platinum group elements. Especially, some neutron capture reactions such as 102 Ru( n , γ) 103 Ru, 104 Ru( n , γ) 105 Ru, and 108 Pd( n , γ) 109 Pd were dealt with in this study. From a comparison between the calculation (C) and the experiment (E) about 102 Ru( n , γ) 103 Ru, the deviation (C/E-1) was signi↓cantly large. Then, it is strongly suspected that not MVP-BURN code but the neutron capture cross section of 102 Ru belonging to JENDL-4.0 used in this simulation have made the big di↑erence as (C/E-1) >20%.

Dominance ratio evolution in large thermal reactors

The dominance ratio is a practical parameter relative to the proximity of the first two eigenvalues in critical nuclear systems. Its high value is a challenging problem for neutron transport simulations especially in the field of large thermal reactors.In this paper, we assess the variation of dominance ratio as a function of burnup for the whole core models of Pressurized Water Reactor and prismatic High Temperature Reactor. We applied the Dirac-delta source relaxation method and continuous-energy Monte Carlo Burnup code MCB5. The results suggest that the dominance ratio tends to increase over irradiation time and it is prone to spatial distribution of neutron absorbers. This stands as an additional difficulty in terms of reactor core control and operation. Concerning Monte Carlo burnup simulation, one can conclude that the number of initially sufficient inactive neutron cycles may be not enough at higher burnup. The evaluated increase of required number of discarded cycles is 45% for the analyzed HTR model and 141% for the PWR system.

Monte Carlo modeling of Lead-Cooled Fast Reactor in adiabatic equilibrium state

Nuclear power would appear to be the only energy source able to satisfy the global energy demand while also achieving a significant reduction of greenhouse gas emissions. Moreover, it can provide a stable and secure source of electricity, and plays an important role in many European countries. However, nuclear power generation from its birth has been doomed by the legacy of radioactive nuclear waste. In addition, the looming decrease in the available resources of fissile U235 may influence the future sustainability of nuclear energy. The integrated solution to both problems is not trivial, and postulates the introduction of a closed-fuel cycle strategy based on breeder reactors. The perfect choice of a novel reactor system fulfilling both requirements is the Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state. In such a state, the reactor converts depleted or natural uranium into plutonium while consuming any self-generated minor actinides and transferring only fission products as waste.We present the preliminary design of a Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state with the Monte Carlo Continuous Energy Burnup Code – MCB. As a reference reactor model we apply the core design developed initially under the framework of the European Lead-cooled SYstem (ELSY) project and refined in the follow-up Lead-cooled European Advanced DEmonstration Reactor (LEADER) project. The major objective of the study is to show to what extent the constraints of the adiabatic cycle are maintained and to indicate the phase space for further improvements. The analysis includes burnup calculations in a two-batch fuel reloading pattern, assessment of the equilibrium fuel composition based on MOX with minor actinides, examination of power peaking with its evolution with burnup, and modeling of control rod operation for its influence on breeding gain and criticality.

Parametric studies of the PWR fuel assembly modeling with Monte-Carlo method

In this paper we present results of studies concerning modeling of critical nuclear systems performed by continuous energy Monte Carlo burnup code. We consider a geometry model of fuel assembly for Pressurized Water Reactor neutronic simulations with depletion calculated in the infinite medium. A variety of modeling parameters and common simplifications concerning geometry complexity have been investigated in order to observe their impact on key reactor parameters like: neutron multiplication factor, nuclide concentrations, conversion ratio and distribution of generated power. Comparison of different factors helps to draw conclusions about their significance and to choose the reliable modeling approach. The presented results should be valuable for a beginning reactor researcher, but the specialists will also find a part of results interesting due to their seldom direct consideration in literature. The conclusions are general and valid not only in case of Monte Carlo simulations.We found that the lack of reactivity control in the model biases the concentration of 235U by more than 10% at end of irradiation. Next, calculation time steps longer than 50days lead to average discrepancy of neutron multiplication factor of 100pcm or more. Axial discretization of fuel requires more than 20 burnable zones to provide asymptotic behavior of burnup calculation. Lack of thermal collision tables results in underestimation of minor actinides concentrations of the order of 10% at low burnup and several percent at the end-of-life. The discrepancies caused by: the radial discretization of fuel pins, the presence of the clad–pellet gap, the applied precision of Monte Carlo calculations and the separate treatment of fuel rods, have been observed at the end-of-life to lead to a maximum relative difference in concentrations of 235U and 239Pu below 1%. Nevertheless, some studies should take also these effects into account.

Monte Carlo burnup in HTR system with various TRISO packing

Monte Carlo codes allow for the calculation of precise representation of complex HTR geometry including the double heterogeneity of a fuel. Various approaches to coated fuel particles (TRISO) placement can be applied and the impact on criticality calculations have been already assessed by several authors. Because static neutron transport simulations are affected by different fuel lattice modeling, it can be expected that the systematic discrepancies should appear during burnup calculation as well.In this paper we compare the impact of different TRISO placement and packing factor in the model on the results of the depletion simulation. For numerical tests we chose simplified model of the fuel block from High Temperature Reactor of prismatic type and the continuous energy Monte Carlo burnup code MCB5. Burnup dependent discrepancies of kinf and nuclide concentrations have been successfully observed. Five arrangement models of fuel kernels have been compared in our work: cuboidal lattice truncated by fuel rod, cubic and hexagonal lattices without rod truncation, cubic lattice with stochastic geometry option and rod filled with one layer of randomly located particles. We put emphasis on the constant mass of fuel kernels in the system and correlate obtained results with physical and statistical aspects of TRISO packing so as to explain observed discrepancies.

Influence of FIMA burnup on actinides concentrations in PWR reactors

In the paper we present the study on the dependence of actinides concentrations in the spent nuclear fuel on FIMA burnup. The concentrations of uranium, plutonium, americium and curium isotopes obtained in numerical simulation are compared with the result of the post irradiation assay of two spent fuel samples. The samples were cut from the fuel rod irradiated during two reactor cycles in the Japanese Ohi-2 Pressurized Water Reactor. The performed comparative analysis assesses the reliability of the developed numerical set-up, especially in terms of the system normalization to the measured FIMA burnup. The numerical simulations were preformed using the burnup and radiation transport mode of the Monte Carlo Continuous Energy Burnup Code – MCB, developed at the Department of Nuclear Energy, Faculty of Energy and Fuels of AGH University of Science and Technology.

Numerical design of the Seed-Blanket Unit for the thorium nuclear fuel cycle

In the paper we present the Monte Carlo modelling by the means of the Monte Carlo Continuous Energy Burn-up Code of the 17x17 Pressurized Water Reactor fuel assembly designed according to the Radkowsky Thorium Fuel concept. The design incorporates the UO2 seed fuel located in the centre and (Th,U)O2 blanket fuel located in the peripheries of fuel assembly. The high power seed region supplies neutrons for the low power blanket region and thus induces breeding of fissile 233 U from fertile 232 Th. The both regions are physically separated and thus this approach is also known as either the heterogonous approach or the Seed-Blanket Unit. In the numerical analysis we consider the time evolutions of infinite neutron multiplication factor, axial/radial power density profile, 233 U, 235 U and 232 Th.

Feedback on neutron capture cross sections of 238Pu and 241Am from analyses of measured isotopic compositions of irradiated LWR fuels and MOX core physics experiments

The atomic fractions of 238Pu and 241Am in MOX fuels recycled in light water reactors are 1% to 2% and not significant compared with those of major Pu isotopes. On the other hand, recent evaluated nuclear data libraries, such as JENDL-4.0 and JEFF-3.2, give noticeably different thermal and epithermal neutron capture cross sections for 238Pu and 241Am. The thermal neutron capture cross sections of 238Pu and 241Am in JEFF-3.2 are 31% and 9% larger than those of JENDL-4.0, respectively. This paper shows the effect of the differences in the neutron cross sections on analysis results of two different integral experiments. The first is the isotopic compositions of 238Pu on UO2 and MOX fuels irradiated in BWR and PWR, and the second is the critical experiments of the water moderated cores fully loaded with MOX fuels. The former was analyzed by using the continuous energy Monte Carlo burnup calculation code MVP-BURN and the latter by the continuous energy Monte Carlo calculation code MVP. The comparisons between the calculated and measured results indicate that the most likely thermal and epithermal neutron capture cross sections of 238Pu and 241Am should be around at the middle between those of JEFF-3.2 and JENDL-4.0.

Comparative analysis between measured and calculated concentrations of major actinides using destructive assay data from Ohi-2 PWR

Abstract In the paper, we assess the accuracy of the Monte Carlo continuous energy burnup code (MCB) in predicting final concentrations of major actinides in the spent nuclear fuel from commercial PWR. The Ohi-2 PWR irradiation experiment was chosen for the numerical reconstruction due to the availability of the final concentrations for eleven major actinides including five uranium isotopes (U-232, U-234, U-235, U-236, U-238) and six plutonium isotopes (Pu-236, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242). The main results were presented as a calculated-to-experimental ratio (C/E) for measured and calculated final actinide concentrations. The good agreement in the range of ±5% was obtained for 78% C/E factors (43 out of 55). The MCB modeling shows significant improvement compared with the results of previous studies conducted on the Ohi-2 experiment, which proves the reliability and accuracy of the developed methodology.