Monte Carlo Continuous Energy Research Articles

Development of a new Monte-Carlo neutron transport code for radiation shielding

A continuous energy Monte-Carlo method-based neutron transport code for fixed source in 3D solid geometry has been developed. The main objective is to integrate this code with recently developed Monte-Carlo gamma shielding code MCBLD, so as to build an in-house code for coupled neutron-gamma shielding study. Also it is intended to use of the new ENDF data library (ENDF/B-VIII) in this code. This paper gives an overview of the code in terms of processing neutron interaction data and numerical scheme to track scattered neutrons. Present capability of the code is demonstrated by presenting fluence due to mono-energetic neutron source as well as widely used spontaneous neutron source from 252Cf isotope, placed at the centre of different size iron spheres. Results are also generated from general purpose open-source Monte-Carlo code FLUKA. Overall good agreement between the two validates the numerical implementation of angle and energy dependent neutron cross-section data in this code.

Evaluation of the CNESTEN's TRIGA Mark II research reactor physical parameters with TRIPOLI-4® and MCNP

This paper focuses on the development of a new computational model of the CNESTEN's TRIGA Mark II research reactor using the 3D continuous energy Monte-Carlo code TRIPOLI-4 (T4). This new model was developed to assess neutronic simulations and determine quantities of interest such as kinetic parameters of the reactor, control rods worth, power peaking factors and neutron flux distributions. This model is also a key tool used to accurately design new experiments in the TRIGA reactor, to analyze these experiments and to carry out sensitivity and uncertainty studies. The geometry and materials data, as part of the MCNP reference model, were used to build the T4 model. In this regard, the differences between the two models are mainly due to mathematical approaches of both codes. Indeed, the study presented in this article is divided into two parts: the first part deals with the development and the validation of the T4 model. The results obtained with the T4 model were compared to the existing MCNP reference model and to the experimental results from the Final Safety Analysis Report (FSAR). Different core configurations were investigated via simulations to test the computational model reliability in predicting the physical parameters of the reactor. As a fairly good agreement among the results was deduced, it seems reasonable to assume that the T4 model can accurately reproduce the MCNP calculated values. The second part of this study is devoted to the sensitivity and uncertainty (S/U) studies that were carried out to quantify the nuclear data uncertainty in the multiplication factor keff. For that purpose, the T4 model was used to calculate the sensitivity profiles of the keff to the nuclear data. The integrated-sensitivities were compared to the results obtained from the previous works that were carried out with MCNP and SCALE-6.2 simulation tools and differences of less than 5% were obtained for most of these quantities except for the C-graphite sensitivities. Moreover, the nuclear data uncertainties in the keff were derived using the COMAC-V2.1 covariance matrices library and the calculated sensitivities. The results have shown that the total nuclear data uncertainty in the keff is around 585 pcm using the COMAC-V2.1. This study also demonstrates that the contribution of zirconium isotopes to the nuclear data uncertainty in the keff is not negligible and should be taken into account when performing S/U analysis.

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.

Development and verification of a Monte Carlo two-step method for lead-based fast reactor neutronics analysis

With the rise of economic and safety standards for nuclear reactors, new concepts of Gen-IV reactors and modular reactors showed more complex designs that challenge current tools for reactor physics analysis. A Monte Carlo (MC) two-step method was proposed in this work. This calculation scheme uses the continuous-energy MC method to generate multi-group cross-sections from heterogeneous models. The multi-group MC method, which can adapt locally-heterogeneous models, is used in the core calculation step. This calculation scheme is verified using a Gen-IV modular lead-based fast reactor (LFR) benchmark case. The influence of homogenized patterns, scatter approximations, flux separable approximation, and local heterogeneity in core calculation on simulation results are investigated. Results showed that the cross-sections generated using the 3D assembly model with a locally heterogeneous representation of control rods lead to an accurate estimation with less than 270 pcm bias in core reactivity, 0.5% bias in control rod worth, and 1.5% bias on power distribution. The study verified the applicability of multi-group cross-sections generated with the MC method for LFR analysis. The study also proved the feasibility of multi-group MC in core calculation with local heterogeneity, which saves 85% time compared to the continuous-energy MC.

Thermal hydraulic modeling of solid-fueled nuclear thermal propulsion reactors part II: Full-core coupled neutronic and thermal hydraulic analysis

This paper focuses on the extension and application of the previously validated thermal–hydraulic engineering code ntpThermo to perform multi-channel analysis for full-core coupled neutronic and thermal–hydraulic simulations. This paper also introduces the Basilisk code, which enables coupled full-core neutronic and thermal–hydraulic simulations by integrating Serpent, the continuous energy Monte-Carlo neutron transport code, and ntpThermo within a flexible object-oriented framework. Today, limited data exists pertaining to coupled multiphysics analysis of low-enriched uranium nuclear thermal propulsion reactors. More specifically, the impact of thermal–hydraulic feedback on the neutronic solution for full core applications is not well quantified. In this paper full-core coupled neutronic and thermal–hydraulic simulations are performed to ascertain the importance of thermal–hydraulic feedback on the predicted reactivity and local power distributions. The reactor design analyzed adheres to the current industry ground rules in an effort to provide useful insights for the current industry design effort. The results in this paper demonstrate that thermal–hydraulic feedback must be considered to accurately predict reactivity; otherwise, errors on the order of hundreds to thousands of pcms can be observed. Additionally, neglecting thermal–hydraulic feedback can trigger inconsistencies between the radial power distribution and the fuel orificing pattern. A design approach that doesn’t account for thermal–hydraulic feedback can lead to a final design that may experience thermal failure via excessive fuel temperatures, and would require a reduction of reactor power. The latter will cause a subsequent decrease in exit propellant temperatures and thus specific impulse of the rocket engine. Sensitivity studies were conducted to determine the importance of specific thermal–hydraulic fields and revealed that the local moderator temperature has a notable impact on the local power distributions.

On the Dimensions Required for a Molten Salt Zero Power Reactor Operating on Chloride Salts

Molten salt reactors have gained substantial interest in the last years due to their flexibility and their potential for simplified closed fuel cycle operation for massive expansion in low-carbon electricity production, which will be required for a future net-zero society. The importance of a zero-power reactor for the process of developing a new, innovative rector concept, such as that required for the molten salt fast reactor based on iMAGINE technology, which operates directly on spent nuclear fuel, is described here. It is based on historical developments as well as the current demand for experimental results and key factors that are relevant to the success of the next step in the development process of all innovative reactor types. In the systematic modelling and simulation of a zero-power molten salt reactor, the radius and the feedback effects are studied for a eutectic based system, while a heavy metal rich chloride-based system are studied depending on the uranium enrichment accompanied with the effects on neutron flux spectrum and spatial distribution. These results are used to support the relevant decision for the narrowing down of the configurations supported by considerations on cost and proliferation for the follow up 3-D analysis. The results provide for the first time a systematic modelling and simulation approach for a new reactor physics experiment for an advanced technology. The expected core volumes for these configurations have been studied using multi-group and continuous energy Monte-Carlo simulations identifying the 35% enriched systems as the most attractive. This finally leads to the choice of heavy metal rich compositions with 35% enrichment as the reference system for future studies of the next steps in the zero power reactor investigation. An alternative could be the eutectic system in the case the increased core diameter is manageable. The inter-comparison of the different applied codes and approaches available in the SCALE package has delivered a very good agreement between the results, creating trust into the developed and used models and methods.

Multi-group neutron cross section generation capability for FRENDY nuclear data processing code

ABSTRACT The multi-group cross-section generation capability for neutrons is implemented in the FRENDY nuclear data processing code. ACE-formatted files are used as the source of nuclear data instead of ENDF-formatted files since FRENDY already has the capability to generate pointwise cross sections in the ACE format. The several distinguished features are implemented for the multi-group generation capability, e.g. explicit consideration of resonance interference effect among nuclides, enhanced resonance treatment for various nuclear reactions, and accurate numerical integration of thermal cross sections. Verification calculations of the newly implemented capability are carried out through the comparison with the NJOY nuclear data processing code. Cross-section generations for all nuclides in JENDL-4.0, -4.0u, -, ENDF/B-VII.1, -VIII.0, JEFF-3.3, and TENDL-2019 are carried out without unexpected processing issues, except for Pu-238 of TENDL-2019 that includes inconsistent data. The verification results indicate that the multi-group cross sections generated by FRENDY are consistent with those generated by NJOY. Now FRENDY can generate not only the pointwise cross sections for continuous energy Monte-Carlo codes but also the multi-group cross sections for deterministic neutronics analysis codes.

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.

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.

Whole-core depletion calculation using domain decomposed continuous-energy Monte Carlo simulation via McBOX with p-CMFD acceleration and inline feedback

This paper presents an approach to the continuous-energy Monte Carlo (MC)-based neutronics-thermal/hydraulics-depletion (N-T/H-D) coupled whole-core calculation. To reduce the memory demand for the detailed depletion calculation, the fission and surface source (FSS) iteration method is applied to the domain decomposed MC criticality calculation. For each iteration (i.e., MC cycle), inline feedback calculations are performed for the stable MC-based coupling calculation, where the partial current-based finite difference (p-CMFD) method is embedded to accelerate the convergence of both the critical boron concentration (CBC) and the neutron source distributions. Then, the subpin-level depletion calculations are performed by the Chebyshev rational approximation method (CRAM). The proposed framework was implemented in the continuous-energy MC code, McBOX, developed at Korea Advanced Institute of Science and Technology (KAIST). The detailed depletion analysis on a typical pressurized water reactor problem shows the improved performance of the proposed framework in terms of memory and computing time.

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.

Application of TULIP/STREAM code in 2-D fast reactor core high-fidelity neutronic analysis

Abstract The deterministic MOC code STREAM of the Computational Reactor Physics and Experiment (CORE) laboratory of Ulsan National Institute of Science and Technology (UNIST), was initially designed for the calculation of pressurized water reactor two- and three-dimensional assemblies and cores. Since fast reactors play an important role in the generation-IV concept, it was decided that the code should be upgraded for the analysis of fast neutron spectrum reactors. This paper presents a coupled code - TULIP/STREAM, developed for the fast reactor assembly and core calculations. The TULIP code produces self-shielded multi-group cross-sections using a one-dimensional cylindrical model. The generated cross-section library is used in the STREAM code which solves eigenvalue problems for a two-dimensional assembly and a multi-assembly whole reactor core. Multiplication factors and steady-state power distributions were compared with the reference solutions obtained by the continuous energy Monte-Carlo code MCS. With the developed code, a sensitivity study of the number of energy groups, the order of anisotropic PN scattering, and the multi-group cross-section generation model was performed on the keff and power distribution. The 2D core simulation calculations show that the TULIP/STREAM code gives a keff error smaller than 200 pcm and the root mean square errors of the pin-wise power distributions within 2%.

Experimental study on fission reaction rate induced by D-T neutron in depleted uranium shell

Fission reaction rate is an important index for validating and checking the neutron transportation and fission power in nuclear engineering. The experimental data can be used in benchmark validation of cross sections, and in studying the correlation of fission power with the thickness of uranium sphere shell. There are five assemblies of depleted uranium shells used in this work, the inner radii of which are all fixed at 13.1 cm, while their outer radii are 18.1, 19.4, 23.35, 25.4 and 28.5 cm, respectively. The D-T neutron source is generated in the center of the assemblies, the yield of which is about 3 × 10<sup>10</sup>−4 × 10<sup>10</sup> s<sup>–1</sup>. In horizontal plane across the center of the assemblies, the fission rates at positions along the radial direction are measured in the direction with 45° inclining with respect to the incident D<sup>+</sup> beam. Due to the disturbance to assemblies and neutron field, the activation foil of uranium is a suitable choice rather than fission chamber or capture detector. The material of activation foil is the same as that in the experimental assemblies. Considering the accurate fission yield of <sup>143</sup>Ce, the objective nuclides are selected. The total fission yield of <sup>143</sup>Ce is contributed by <sup>238</sup>U and a little <sup>235</sup>U. For calculating the total fission yiled of <sup>143</sup>Ce, the neutron energy range of 0−15 MeV is divided into eight subranges. By measuring the 293 keV gamma rays from the fission product <sup>143</sup>Ce in activation foils with a TRANS-SPEC-DX100 HPGe detector, with a relative efficiency 40%, the fission rates and the trends at positions along the radial direction in the five assemblies are obtained based on the <sup>143</sup>Ce fission product yield. The fission rate ranges from 5.28 × 10<sup>–29</sup> to 7.58 × 10<sup>–28</sup> sn<sup>-1</sup>·nuclide<sup>–1</sup>, with the relative uncertainty in a range from 6% to 11%. The Monte Carlo transport code MCNP5 and continuous energy cross section library ENDF/BV.8 are used for analyzing the fission rate distribution in the assemblies, and the experiemtal configuration, including the wall of the experimental hall is described in detail in the model. The calculated results are compared with the experimental ones and their agreement is found to be in an uncertainty range.

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.

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.

Burnup analysis for the pebble-bed fluoride-salt-cooled high-temperature reactor based on the energy-dependent elastic scattering cross-sectional model

To carry out accurate burnup calculations for a pebble-bed fluoride-salt-cooled high-temperature reactor, the energy-dependent cross-sectional model based on the Doppler broadening rejection correction method has been proposed to develop the energy-dependent elastic scattering cross-sectional model. In this study, the Monte Carlo continuous energy code PSG2/Serpent was used to examine the difference between the constant cross-sectional model and the energy-dependent cross-sectional model during burnup. For the cases analyzed in this study, numerical simulations show that the multiplication coefficient was improved by hundreds pcm and 239Pu concentration was improved by approximately 1–2% during burnup when the energy-dependent elastic scattering cross-sectional model is considered.

Inline critical boron concentration search with p-CMFD feedback in whole-core continuous-energy Monte Carlo simulation

For the pressurized water reactors (PWRs) depletion analysis, the critical boron concentration (CBC) search is required. In the Monte Carlo (MC) simulation, the CBC can be estimated by the neutron balance approach, which is called as the inline CBC search method in this paper. This method is based on the fixed-point iteration, which results in a number of inactive cycles for the MC simulation. For the acceleration of the convergence and real variance reduction, the p-CMFD feedback is applied to the inline CBC search method. Furthermore, a rejection technique is introduced in the p-CMFD calculation to suppress the error propagation from the p-CMFD parameters to the p-CMFD solution. For a near critical system with zero boron concentration, the inline CBC search method may get in trouble with the negative boron concentration. To solve this problem, the extended particle-tracking algorithm is proposed and verified in a simplified toy problem. In realistic PWR test problems, the numerical results show that the proposed method accurately estimates the CBCs, while the p-CMFD feedback effectively reduces the number of inactive cycles. In conclusion, for the PWR depletion analysis, the continuous-energy MC simulation under the CBC condition can be efficiently performed by the inline CBC search method with the p-CMFD feedback.

Radially and azimuthally dependent resonance self-shielding treatment for general multi-region geometry based on a unified theory

ABSTRACT A unified resonance self-shielding method, which can treat general sub-divided fuel regions, is developed for lattice physics calculations in reactor physics field. In a past study, a hybrid resonance treatment has been developed by theoretically integrating equivalence theory and ultra-fine-group slowing-down calculation. It can be applied to a wide range of neutron spectrum conditions including low moderator density ranges in severe accident states, as long as each fuel region is not sub-divided. In order to extend the method for radially and azimuthally sub-divided multi-region geometry, a new resonance treatment is established by incorporating the essence of sub-group method. The present method is composed of two-step flux calculation, i.e. ‘coarse geometry + fine energy’ (first step) and ‘fine geometry + coarse energy’ (second step) calculations. The first step corresponds to a hybrid model of the equivalence theory and the ultra-fine-group calculation, and the second step corresponds to the sub-group method. From the verification results, effective cross-sections by the new method show good agreement with the continuous energy Monte-Carlo results for various multi-region geometries including non-uniform fuel compositions and temperature distributions. The present method can accurately generate effective cross-sections with short computation time in general lattice physics calculations.