Burnup Simulations Research Articles

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.

Математическое моделирование обгара углеродного материала в стационарном высоконапорном воздушном потоке

Математическое моделирование обгара углеродного материала в стационарном высоконапорном воздушном потоке

Experimental validation of a nuclear forensics methodology for source reactor-type discrimination of chemically separated plutonium

Abstract An experimental validation of a nuclear forensics methodology for the source reactor-type discrimination of separated weapons-useable plutonium is presented. The methodology uses measured values of intra-element isotope ratios of plutonium and fission product contaminants. MCNP radiation transport codes were used for various reactor core modeling and fuel burnup simulations. A reactor-dependent library of intra-element isotope ratio values as a function of burnup and time since irradiation was created from the simulation results. The experimental validation of the methodology was achieved by performing two low-burnup experimental irradiations, resulting in distinct fuel samples containing sub-milligram quantities of weapons-useable plutonium. The irradiated samples were subjected to gamma and mass spectrometry to measure several intra-element isotope ratios. For each reactor in the library, a maximum likelihood calculation was utilized to compare the measured and simulated intra-element isotope ratio values, producing a likelihood value which is proportional to the probability of observing the measured ratio values, given a particular reactor in the library. The measured intra-element isotope ratio values of both irradiated samples and its comparison with the simulation predictions using maximum likelihood analyses are presented. The analyses validate the nuclear forensics methodology developed.

Calculations of spent fuel isotopic composition for fuel rod from VVER-440 fuel assembly benchmark using several evaluated nuclear data libraries

Abstract This article presents a comparison of results of simulation of VVER-440 reactor fuel burnup by means of ISTAR set of tools using ENDF/B-VII.0 and JEFF-3.1 evaluated nuclear data libraries. Code verification is required on the basis of both experimental data and computational uncertainty analysis. A VVER-440 fuel rod burnup simulation using ENDF/B-VII.0 and JEFF-3.1 evaluated nuclear data libraries was performed as preliminary step before comparison of nuclear data uncertainty analysis and calculation/experiment. Fuel irradiation conditions and SNF nuclide composition experimental data were obtained from radiochemical assays of irradiated VVER-440 fuel report. Fuel burnup simulation was carried out by means of ISTAR set of tools using Monte-Carlo code for spectrum averaged reaction rates evaluation required for solution of Bateman's equations. As a result, Kinf and nuclide composition across burnup were estimated for the sample considered. On the basis of two nuclear data libraries’ implementation for depletion simulation it was shown that Kinf uncertainties rise across burnup from 0.4% at the beginning of life up to 0.7% at the end of life. Heavy nuclides composition uncertainties run up to 9.5% for Am-241 at the end of the 4th life time; however for main uranium and plutonium isotopes the uncertainties do not exceed 1.2%. Most fission products composition uncertainties stay below 7.3% except Sm-150 nuclide (15.1%). The next step will be a comparison with experimental spent nuclear fuel isotopic composition data considering the above mentioned uncertainties.

Calculation chain for the analysis of spent nuclear fuel in long-term interim dry storage

In this paper, a calculation chain for the analysis of spent nuclear fuel in long-term interim dry storage is demonstrated. The calculation chain consists of reactor physics burnup simulation for obtaining the decay heat, Computational Fluid Dynamics (CFD) simulation for the temperatures of fuel rods and the dry storage cask, and finally, cladding integrity analysis. Each succeeding calculation step uses the results from the previous step as boundary conditions. The applied codes are Serpent Monte Carlo reactor physics burnup code, open-source OpenFOAM CFD code, and VTT-ENIGMA fuel performance code. The peak cladding temperatures are calculated with OpenFOAM at different instants of time, ranging from 3.4 to 300 years after unloading from the reactor. The chosen demonstration case considers 17 × 17 PWR fuel stored in a CASTOR® V/21 type dry storage cask. The main parameters of interest in the last stage of the analysis are the cladding creep hoop strain and stress during dry storage. The developed analysis methodology helps to ensure the safety of long-term dry storage.

Initial thermal-hydraulic assessment by OpenFOAM and FLUENT of a subcritical irradiation facility

Aim of this work is the preliminary thermal-fluid-dynamic assessment of a subcritical system to perform integral measurements on transmutation processes, designed in the frame of EU CHANDA project: in particular, a low power Accelerator Driven System (ADS) may represent an attractive intermediate step to fill the gap between existing and future facilities like MYRRHA (or possibly EFIT). The activity finds its place within the roadmap for the evaluation of transmutation processes in ADSs, where currently measurements are performed on the zero-power ADS Guinevère at SCK·CEN (Mol, Belgium), while in the future it is expected that MYRRHA will be the major high-power facility for performing this kind of studies and assessing the performance of a fast ADS.The neutronic characterization and burn-up simulations based on MCNP6 code of the reactor core allowed the definition of the geometry and the composition of the fuel assembly and, therefore, the thermal power to be removed. On this basis a thermo-fluid-dynamic assessment, through simple analytical accounts and detailed 3D CFD calculations by OpenFOAM-v1612+ and ANSYS FLUENT v17.0 was made.The present paper is an extended version including the main conclusions and observations emerged in the 2nd AIGE-IIETA International Conference in Genoa.

Conceptual study of an accelerator-driven ceramic fast reactor with long-term operation

To improve both safe operation and high resource utilization in nuclear power, we propose and investigate the concept of an accelerator-driven ceramic fast reactor (ADCFR). This reactor type has the potential to operate continuously throughout a 40-year core life, without fuel shuffling or supplementation. The ADCFR consists of a high-power superconducting linear accelerator, a gravity-driven dense granular spallation-target, and a ceramic fast reactor. The performance of the ADCFR was assessed by using a neutron-physics simulation, thermal calculations, and a characteristic analysis. The results show that the peak position for the neutron spectrum in the ADCFR is at about 0.1 MeV. This means that it falls with the fast neutron spectrum, and it can convert loaded nuclear fertile material into fissile fuel. Using a burnup simulation, the ideal effective multiplication-factor (Keff) was calculated by using a combination of subcritical (accelerator-driven) and critical modes. In 40 year of operation, Keff is obtained from the initial 0.98 to the peak ~1.02 and then to ~0.99. Different granular coolant materials were selected to compare neutron performance. In breeding, the differences are relatively small. The thermal calculation indicates that heat transfer performance of granular makes it possible to meet the required specifications in theory. Finally, the corresponding characteristics, with regard to the 2-phase coolant, ceramic materials, nuclear safety performance, operation modes, economics, and range of applications were analyzed. Accelerator-driven ceramic fast reactors can achieve very high levels of inherent safety, good breeding performance, high power-generation efficiency, and high flexibility in wide range of applications.

The fuel behaviour code FUROM and its high burn-up simulation capabilities

The code FUROM has been developed for the computational simulation of VVER and PWR fuel behaviour in quasi-static conditions. The applicability and flexibility of the code were ensured in the fields of cladding mechanical response, fission gas release, isotopic content of the fission gas etc. Special attention was given to the validation of the code by using experimental results from international sources. The code is regularly used for the systematic evaluation of the fuel-related safety parameters of Paks NPP. The satisfactory conclusion of this evaluation is a pre-requisite of licensing all modified fuel designs.

Construction of minor actinides reduction scenario in Japan utilizing fusion reactors

This study demonstrates by numerical simulations how transmuting minor actinides in fusion reactors can contribute to nuclear fuel cycles in Japan. Minor actinides are installed into blankets of a helical-type fusion reactor to transmute them rather moderately so that neither large neutron wall loading nor removal of large heat are required. Neutron transport and burn-up simulations are conducted to optimize the position of minor actinides and the blanket's structure for transmutation. Whereas the ratio of transmuted minor actinides is not large, the total amount of transmuted minor actinides is revealed to be sufficient because of the system's large inventory. The simulations are followed by a plan that introduces fusion reactors intended for transmutation. Discussion based on Japan's present situation confirms that introduction of fusion reactors with a thermal fusion output of 1 GW by the year 2050 would mitigate many issues concerning disposal of high-level radioactive waste.

Validation of LOOP through PWR measured data

This paper serves two purposes. First it demonstrates the validation of newly developed and verified burnup code, Linkage of ORIGEN and OpenMC (LOOP) against measured data of Post Irradiation Experiments (PIE) performed at Takaham-3 PWR. Secondly, it assesses the qualification of OpenMC model of Chashma Nuclear Power Plant unit-II (CNPP-II) PWR to predict different operating conditions of the reactor core. The LOOP code is then employed to perform the burnup simulations of CNPP-II first cycle with fresh UO2 fuel. The LOOP calculated results of power distribution of the first cycle at Beginning of Cycle (BOC), Middle of Cycle (MOC) and End of Cycle (EOC) are then compared with the operational data of the reactor. The maximum difference between the LOOP calculated radial and axial power distribution results and the measured data is 9.1% and 2.3% at BOC. At EOC, this difference further decreases to 7.2% and 1.5%. The depletion of burnable poison rods in the reactor was also compared with the FSAR predicted data which are in exact agreement with each other. The production of Cs-137 and Cs-134 was also investigated for the validity of LOOP burnup simulations.

Thorium-based mixed oxide fuel in a pressurized water reactor: A burnup analysis with MCNP

A three-dimensional MCNP model of a Westinghouse-type 17×17 PWR was used to assess the feasibility of enhancing the current fuel supply with a thorium-plutonium mixed oxide fuel (ThMOX). A conventional UO2-fueled configuration was compared to a configuration which incorporated ThMOX fuel. The safety of the ThMOX configuration was compared to that of the UO2 configuration at several time steps of interest within all three cycles (beginning of cycle, peak excess reactivity of cycle, and end of cycle) using the following metrics: axial and radial nuclear hot channel factors FZNandFRN, moderator and fuel temperature coefficients (MTC and FTC), delayed neutron fraction (βeff), and shutdown margin (SDM). Additionally, the performance of the ThMOX configuration was assessed by tracking cycle lengths, the amount of plutonium destroyed, and fission product poison concentration.The ThMOX configuration’s FZN (1.455±0.222) was slightly less favorable than that of the UO2 configuration (1.515±0.246), though its FRN was statistically identical (3.623±0.179 compared to 3.430±0.066 for UO2). The delayed neutron fraction was another area of concern for the ThMOX configuration. It was lower than that of the UO2 configuration, but the lowest βeff occurred at the beginning of Cycle 3 (3.324E−03±8.919E−07). The βeff of the UO2 configuration was 46% higher (6.177E−03±1.510E−06) at this time step. Similarly, the ThMOX configuration’s SDMs were less favorable than those of the UO2 configuration, though even the lowest (3971±43pcm) remained well within the NRC prescribed limit of 1300pcm. However, the MTC of the ThMOX configuration was more favorable for all time steps except the beginning of Cycle 3, and even then it was negative as required by the NRC (−2.325±0.00098pcm/°F). The FTC results were also more favorable for the ThMOX configuration. Its highest FTC value (−6.718±0.001pcm/°F at the peak of Cycle 1) was more negative than all but the lowest FTC of the UO2 configuration (−7.335±0.002pcm/°F at the end of Cycle 3).Overall, incorporating ThMOX fuel into the fuel supply for a conventional PWR seems feasible from a neutronics standpoint. However, this analysis would benefit from a more detailed burnup simulation and optimization of the ThMOX fuel composition. Also, a full thermal hydraulic analysis in conjunction with a neutronics analysis is required.

The results of the transmutation of fission fragments in the spectrum of neutrons of thermal and fast reactors

Radioactivity of spent nuclear fuel (SNF) discharged from nuclear reactor core is determined during the first 100 years by fission fragments (FF), after that the main contribution in the SNF activity is made by actinides. Existing scenarios of SNF handling are based on the transmutation of minor actinides (MA) into fission fragments accomplished in fast reactors. Scenarios of transmutation of fission fragments in thermal and fast neutron spectra and time-dependent radiation characteristics are examined in the present study. Nuclide composition of fission fragments is taken from the results of simulation of burnup of 439GT fuel assembly (TVSA type) for VVER-1000 nuclear reactor during 3 years performed using MCU-5 software complex. The obtained data were used for determining starting nuclide composition for different cooling-down times prior to the beginning of transmutation (irradiation in neutron fluxes) to be input in the ORIGEN2 code.The following three options of irradiation of fission fragments are presented: transmutation without cooling down, cooling down fission fragments during 4 years prior to irradiation, cooling down fission fragments during 30 prior to irradiation. Duration of irradiation was selected to be equal to 3 and 15 years. Efficiency of transmutation was determined using time-dependent “transmutation factor” equal to the ratio of radioactivity of nuclides in the process of transmutation and after its completion to their radioactivity without transmutation.The calculated values of transmutation factors proved to be noticeable only during irradiation in reactor core: these values reached 5–10 and were dependent only on the duration of fission fragment cooling down prior to the beginning of transmutation. After removal of fission fragments from neutron flux transmutation factor decreased to unity within several years. After one hundred more years after irradiation in neutron spectrum of thermal reactor transmutation factor reduces to 0.8–0.5 depending on the duration of the transmutation process. Slight growth of transmutation factor to the values of 1.2–1.8 was observed after irradiation in fast reactor spectrum within the time interval of 200–1000 years and after 1000 years following this its reduction to the value of 0.9–0.7 was noted.The main conclusion is that purposeful incineration of fission fragments is senseless because only insignificant gain in radioactivity (a little less than by the factor two) is achieved after 1000 years.The indifference of fission fragments with regard to transmutation can be partially explained by the fraction of stable nuclides which increases with extension of the period of fission fragments cooling down. Upon completion of the cycle of fuel use it contains approximately 15% of stable nuclides among fission fragments, and after 30 years of cooling down the fraction of stable isotopes reaches 85%.

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.

Underestimation of nuclear fuel burnup – theory, demonstration and solution in numerical models

Monte Carlo methodology provides reference statistical solution of neutron transport criticality problems of nuclear systems. Estimated reaction rates can be applied as an input to Bateman equations that govern isotopic evolution of reactor materials. Because statistical solution of Boltzmann equation is computationally expensive, it is in practice applied to time steps of limited length. In this paper we show that simple staircase step model leads to underprediction of numerical fuel burnup (Fissions per Initial Metal Atom – FIMA). Theoretical considerations indicates that this error is inversely proportional to the length of the time step and origins from the variation of heating per source neutron. The bias can be diminished by application of predictor-corrector step model. A set of burnup simulations with various step length and coupling schemes has been performed. SERPENT code version 1.17 has been applied to the model of a typical fuel assembly from Pressurized Water Reactor. In reference case FIMA reaches 6.24% that is equivalent to about 60 GWD/tHM of industrial burnup. The discrepancies up to 1% have been observed depending on time step model and theoretical predictions are consistent with numerical results. Conclusions presented in this paper are important for research and development concerning nuclear fuel cycle also in the context of Gen4 systems.

Assesment of advanced step models for steady state Monte Carlo burnup calculations in application to prismatic HTGR

Abstract In this paper, we compare the methodology of different time-step models in the context of Monte Carlo burnup calculations for nuclear reactors. We discuss the differences between staircase step model, slope model, bridge scheme and stochastic implicit Euler method proposed in literature. We focus on the spatial stability of depletion procedure and put additional emphasis on the problem of normalization of neutron source strength. Considered methodology has been implemented in our continuous energy Monte Carlo burnup code (MCB5). The burnup simulations have been performed using the simplified high temperature gas-cooled reactor (HTGR) system with and without modeling of control rod withdrawal. Useful conclusions have been formulated on the basis of results.

Study on a method for loading a Li compound to produce tritium using high-temperature gas-cooled reactor

Tritium production using high-temperature gas-cooled reactors and its outflow from the region loading Li compound into the helium coolant are estimated when considering the suppression of tritium outflow. A Li rod containing a cylindrical Li compound placed in an Al2O3 cladding tube is assumed as a method for loading Li compound. A gas turbine high-temperature reactor of 300MW electrical nominal capacity (GTHTR300) with 600MW thermal output power is considered and modeled using the continuous-energy Monte Carlo transport code MVP-BURN, where burn-up simulations are carried out. Tritium outflow is estimated from equilibrium solution for the tritium diffusion equation in the cladding tube. A GTHTR300 can produce 400–600g of tritium over a 180-day operation using the chosen method of loading the Li compound while minimizing tritium outflow from the cladding tube. Optimizing tritium production while suppressing tritium outflow is discussed.

Comparison of nuclear data uncertainty propagation methodologies for PWR burn-up simulations

Several methodologies using different levels of approximations have been developed for propagating nuclear data uncertainties in nuclear burn-up simulations. Most methods fall into the two broad classes of Monte Carlo approaches, which are exact apart from statistical uncertainties but require additional computation time, and first order perturbation theory approaches, which are efficient for not too large numbers of considered response functions but only applicable for sufficiently small nuclear data uncertainties. Some methods neglect isotopic composition uncertainties induced by the depletion steps of the simulations, others neglect neutron flux uncertainties, and the accuracy of a given approximation is often very hard to quantify. In order to get a better sense of the impact of different approximations, this work aims to compare results obtained based on different approximate methodologies with an exact method, namely the NUDUNA Monte Carlo based approach developed by AREVA GmbH. In addition, the impact of different covariance data is studied by comparing two of the presently most complete nuclear data covariance libraries (ENDF/B-VII.1 and SCALE 6.0), which reveals a high dependency of the uncertainty estimates on the source of covariance data. The burn-up benchmark Exercise I-1b proposed by the OECD expert group “Benchmarks for Uncertainty Analysis in Modeling (UAM) for the Design, Operation and Safety Analysis of LWRs” is studied as an example application. The burn-up simulations are performed with the SCALE 6.0 tool suite.

Validation of the burn-up code EVOLCODE 2.0 with PWR experimental data and with a Sensitivity/Uncertainty analysis

A validation of the burn-up simulation system EVOLCODE 2.0 is presented here, involving the experimental measurement of U and Pu isotopes and some fission fragments production ratios after a burn-up of around 30GWd/tU in a Pressurized Light Water Reactor (PWR). This work provides an in-depth analysis of the validation results, including the possible sources of the uncertainties. An uncertainty analysis based on the sensitivity methodology has been also performed, providing the uncertainties in the isotopic content propagated from the cross sections uncertainties. An improvement of the classical Sensitivity/Uncertainty (S/U) model has been developed to take into account the implicit dependence of the neutron flux normalization, that is, the effect of the constant power of the reactor. The improved S/U methodology, neglected in this kind of studies, has proven to be an important contribution to the explanation of some simulation-experiment discrepancies for which, in general, the cross section uncertainties are, for the most relevant actinides, an important contributor to the simulation uncertainties, of the same order of magnitude and sometimes even larger than the experimental uncertainties and the experiment-simulation differences. Additionally, some hints for the improvement of the JEFF3.1.1 fission yield library and for the correction of some errata in the experimental data are presented.

Burnup simulations of different fuel grades using the MCNPX Monte Carlo code

Global energy problems range from the increasing cost of fuel to the unequal distribution of energy resources and the potential climate change resulting from the burning of fossil fuels. A sustainable nuclear energy would augment the current world energy supply and serve as a reliable future energy source. This research focuses on Monte Carlo simulations of pressurized water reactor systems. Three different fuel grades - mixed oxide fuel (MOX), uranium oxide fuel (UOX), and commercially enriched uranium or uranium metal (CEU) - are used in this simulation and their impact on the effective multiplication factor (Keff) and, hence, criticality and total radioactivity of the reactor core after fuel burnup analyzed. The effect of different clad materials on Keff is also studied. Burnup calculation results indicate a buildup of plutonium isotopes in UOX and CEU, as opposed to a decline in plutonium radioisotopes for MOX fuel burnup time. For MOX fuel, a decrease of 31.9% of the fissile plutonium isotope is observed, while for UOX and CEU, fissile plutonium isotopes increased by 82.3% and 83.8%, respectively. Keff results show zircaloy as a much more effective clad material in comparison to zirconium and stainless steel.

A Simplified Supercritical Fast Reactor with Thorium Fuel

Super-Critical water-cooled Fast Reactor (SCFR) is a feasible option for the Gen-IV SCWR designs, in which much less moderator and thus coolant are needed for transferring the fission heat from the core compared with the traditional LWRs. The fast spectrum of SCFR is useful for fuel breeding and thorium utilization, which is then beneficial for enhancing the sustainability of the nuclear fuel cycle. A SCFR core is constructed in this work, with the aim of simplifying the mechanical structure and keeping negative coolant void reactivity during the whole core life. A core burnup simulation scheme based on Monte Carlo lattice homogenization is adopted in this study, and the reactor physics analysis has been performed with DU-MOX and Th-MOX fuel. The main issues discussed include the fuel conversion ratio and the coolant void reactivity. The analysis shows that thorium-based fuel can provide inherent safety for SCFR without use of blanket, which is favorable for the mechanical design of SCFR.