Cross-section Library Research Articles

Comprehensive analysis of ASTRA benchmark with MCU-HTR and Serpent codes

Key problems in the simulation of pebble-bed high-temperature gas-cooled reactors include approximations of TRISO fuel particles and pebble-bed models, choice of nuclear data libraries, and uncertainty in graphite impurity content. This study is focused on the investigation of these problems for the ASTRA critical facility simulation with MCU-HTR and Serpent Monte Carlo codes. This paper presents a comparative analysis of the ASTRA benchmark (IEU-COMP-THERM-008) through consistent code-to-code comparisons. Some studies were performed on the simplified models of the ASTRA facility. A special care was taken to develop similar models of the ASTRA facility for the calculations using MCU-HTR and Serpent. A good agreement in the multiplication factor (less than 160pcm) between the simulations of ASTRA critical configurations using MCU-HTR and Serpent with ENDF/B-VII.0 cross-section libraries was obtained. In addition, the reactivity impact of different graphite thermal scattering libraries included in ENDF/B-VIII.0 release on the calculated results for the ASTRA benchmark was examined. Graphite evaluations assuming 0, 10, and 30 % porosity were considered in Serpent simulations with ENDF/B-VIII.0 cross-section libraries. The spread of the calculated keff for the measured critical configurations having different core height was found to be approximately 200pcm for most selected models. Consequently, model’s ability to correctly capture the procedure of a reactor loading during startup was proven. However, a significant overestimation of the keff (1000–1500pcm) was observed when we used the equivalent boron content (EBC) in the reflectors graphite adjusted to reproduce the measured graphite absorption cross-section, the EBC in matrix graphite of 1 ppmwt., ENDFB/VIII.0 cross-sections, and 30 % or 10 % porous graphite thermal scattering libraries in the Serpent simulations.

MPOGen: A Python package to prepare few-group homogenized cross sections for core calculations by APOLLO3®

Since 2008, CEA has been developing a new generation of multi-purpose code called APOLLO3®. APOLLO3® integrates lattice calculation and core simulations into a single software package designed for Light Water Reactors (LWR) and Fast Reactors analysis. This paper introduces MPOGen, an object-oriented module written in Python3, developed at CEA to automatize the generation of input files for the preparation of few-group homogenized cross section libraries with the new lattice part of APOLLO3®. The module drives APOLLO3® through the computation of few-group homogenized cross sections (Schneider et al., 2016). It significantly expedites the input set-up process, resulting in substantial time savings for engineers.MPOGen still targets market of LWR, as PWR and VVER. The package supports an arbitrary number of parameter space. In this work, six parameters are used to span Cartesian and non-Cartesian grids. MPOGen also supports custom geometries generated by graphical interfaces (GUI), notably Silène and ALAMOS (Tomatis et al., 2022).

Transmutation of Transuranic Elements as Solid Coating in Molten Salt Reactor Fuel Channel

The accumulation of spent nuclear fuel (SNF) is presently considered as a hindrance of the massive deployment of nuclear power plant, especially regarding the transuranic (TRU) elements. Eliminating TRU through transmutation is one of the most feasible alternative as a technical solution to solve the issue. This study explores the possibility of TRU transmutation using molten salt reactor (MSR) in a heterogeneous configuration, where a solid TRU is coated inside the fuel channel filled with liquid salt fuel. Such configuration is proposed to allow higher TRU loading into fluoride salt mixture without compromising the safety of the reactor. TRU coating was applied in consecutively outward radial fuel channel layers with coating thicknesses of 2.5 mm and 5 mm. Calculation was performed using MCNP6.2 radiation transport code and ENDF/B-VII.0 neutron cross section library. From the results, TRU coating with smaller thickness and positioned closer to the centre of the core exhibit higher transmutation efficiency due to exposure to higher neutron flux. Highest transmutation efficiency was achieved at 67.93% after 160 days of burnup. This shows a potential of achieving highly efficient TRU using heterogeneous configuration in MSR core.

The influence of graphite moderator density on the neutronic performance and fuel economy of a Molten Salt Reactor

Typical thermal molten salt reactor (MSR) employs graphite as its neutron moderator, due to its chemical compatibility and low neutron absorption. It is not without issues, however, since graphite provides positive temperature feedback when the MSR is fueled with U-233. Reducing the graphite density can possibly ameliorate this issue, but at the cost of other neutronic parameters, such as fertile conversion into fissile nuclide, and fuel economy, such as fissile makeup cost. This study evaluates the influence of graphite moderator density on the neutronic performance and fuel economy of an MSR. Passive Compact Molten Salt Reactor (PCMSR) design was used as the MSR model. To evaluate the neutronic parameters, MCNP6.2 radiation transport code with ENDF/B-VII.0 neutron cross section library were used. The evaluated neutronic parameters were temperature coefficient of reactivity (TCR), kinetic parameters, and conversion ratio (CR). Meanwhile, the fuel economy evaluation comprised the initial fuel inventory cost and fissile makeup cost. From the calculation results, lowering graphite density remarkably improves TCR, but also decreased CR in a notable amount. While effective delayed neutron fraction (βeff) was not particularly affected, mean neutron generation time (Λ) was reduced in lower graphite density. Considering the fuel economy, initial fuel cost is not significantly affected, but the fissile makeup cost can be sensitive of graphite density.

Cross section generation for a conceptual horizontal, compact high temperature gas reactor

A macroscopic cross section generation model was developed for the conceptual horizontal, compact high temperature gas reactor (HC-HTGR). Because there are many sources of spectral effects in the design and analysis of the core, conventional LWR methods have limitations for accurate simulation of the HC-HTGR using a neutron diffusion core neutronics simulator. Several super-cell model configurations were investigated to consider the spectral effect of neighboring cells. A new history variable was introduced for the existing library format to more accurately account for the history effect from neighboring nodes and reactivity control drums. The macroscopic cross section library was validated through comparison with cross sections generated using full core Monte Carlo models and single cell cross section for both 3D core steady-state problems and 2D and 3D depletion problems. Core calculations were then performed with the AGREE HTR neutronics and thermal-fluid core simulator using super-cell cross sections. With the new history variable, the super-cell cross sections were in good agreement with the full core cross sections even for problems with significant spectrum change during fuel shuffling and depletion.

Shut-down dose rate analysis for final design of shielding tent in port interspace of ITER UP#18

Upper Port #18 (UP18) of ITER is one of the procurement items of the Korea Domestic Agency (KODA), and the final design is being carried out. UP18 as a diagnostics port is divided into 3 sections: port plug (PP), port interspace (PI), and port cell (PC) depending on the distance from the plasma. Because the PI and PC include diagnostics devices requiring maintenance such as valves and detectors, there is a design requirement for human access in a radiation environment. A separate shielding is necessary to satisfy the requirement. In the PI of UP18 the port interspace of UP18, a shielding tent concept was proposed to effectively mitigate the radiation cross-talk which is dose from neighboring ports as well as UP18 PP.Therefore, in this study, neutronics analysis was conducted to assess the shielding tent design. A parametric study for the shielding frame materials, design options, and shielding materials was performed. A shut-down dose rate (SDDR) in a maintenance region was evaluated using SRC-UNED v3.1.4 with FENDL 2.1 and 3.1 cross-section libraries. The analysis results show that the contribution of the shielding frames to SDDR is significantly high in the reference shielding tent design and especially the SDDR difference is attributed to the weight fraction of impurities such as cobalt (Co), niobium (Nb), and tantalum (Ta) within the frame. Borated heavy concrete (BHC) is the effective shielding material in the PI because it shields the neutrons and photons during operation and maintenance. Based on these results, the shielding tent material and design was optimized, considering both the manufacturing cost and the SDDR reduction. The calculated SDDR is 83.9 μSv/hr, which satisfies the design requirement.

Neutronic analysis of the BYU molten salt micro reactor

Background: With the rising need for reliable and clean energy, researchers at Brigham Young University created a molten salt microreactor (MSMR) concept to help meet the world’s growing energy demands. The MSMR is rated at 45 MWth and is uniquely capable of passively dissipating all decay heat using conduction only. Methods: This work presents the results of the simulated neutronic behavior of the MSMR using the Monte Carlo neutronics code OpenMC with the ENDF/B-VII.1 cross-section library at steady state conditions. The specific characteristics discussed in this work are: keff, power profile, axial and radial peak power factors, Doppler coefficient of reactivity, and the moderator temperature feedback coefficient. Results: At steady state operating conditions, keff was calculated to be 1.07091 ± 0.00087 indicating that this configuration can achieve criticality. The axial peaking power factor was found to be 1.003 and the radial peaking power factor was found to be 1.45. Feedbacks were calculated by offsetting the temperature of the fuel and the moderator respectively. Doppler coefficient of reactivity to be -12.585 pcm/K and the moderator temperature coefficient of reactivity to be -0.34332 pcm/K. Conclusions: It is worth noting that the largest contributor of the negative reactivity comes from the fuel salt and not the moderator, further demonstrating the safety of the reactor. Future work will include a full neutronic optimization of the core, fuel burn up analysis, and exploration of a suitable burnable neutron poison.

Overview of Method of ChAracteristics based Neutron TRAnsport Code (MANTRA) and its validation

An in-house computer code MANTRA is developed to solve the neutron transport equation at lattice or assembly level (in 2D) applying Method of Characteristics (MOC) for reactor design calculation. The code has the capability to divide the lattice into triangular meshes. By default, the code assumes constant neutron source inside the meshes. If required, it is possible to use linear source assumption to reduce the number of meshes for speeding up the solution. Using MOC, the solution of transport equation is obtained along a number of parallel lines or rays traced through the meshes in different directions and connect them through reflective or periodic boundary condition. In the code, there is option to choose either Power or Krylov subspace iteration technique to converge the solution with desired accuracy. The code is coupled with multigroup cross section library WIMSD. Performance of MANTRA is tested against a number of benchmark problems and overall good agreement is observed with the results obtained from internationally recognized codes like WIMSD, DRAGON, MCNP etc.

Fine mesh pin-by-pin analysis of IPWR equilibrium core

Detailed fine mesh reactor physics simulations are warranted for reactor cores in order to estimate the important safety parameters such as detailed pin power distribution in the core and local pin peaking factors. A pin-by-pin reactor analysis code HEXPIN has been developed to perform 3D reactor core analysis with a pincell size mesh. HEXPIN uses the conventional two step calculation methodology for performing reactor calculations. India is developing a 2700 MWth/900 MWe Indian Pressurized Water Reactor (IPWR) for developing indigenous capability for commercial light water technology. It uses enriched uranium oxide fuel. The batch refueling scheme is adopted for fuel cycle management. A fine mesh pin-by-pin analysis of equilibrium core of IPWR has been performed using HEXPIN code system. The pin homogenised parametric cross section library, required for HEXPIN calculation, has been generated using lattice burn up code EXCEL. The equilibrium core calculation has been performed by a set of repeated calculations till a converged core configuration is achieved. The present paper gives the methodology and brief results of the analysis.

A deep-learning representation of multi-group cross sections in lattice calculations

To compute few-group nodal cross sections, lattice codes must first generate multi-group cross sections using continuous energy cross-section libraries for each material in each fuel cell. Since the processing cost is significant, we propose representing the multi-group cross sections during lattice calculations using a pre-trained deep-learning-based model. The model utilizes a combination of Principal Component Analysis (PCA) and fully connected Neural Networks (NN). The model is specifically designed to manage extensive multi-group cross-section data sets, which contain data for several dozen nuclides and encompass more than 50 energy groups. Our testing of the trained model on a PWR assembly with a realistic boron letdown curve revealed an average relative error of around 0.1% for both fission and total macroscopic cross sections. The average time required for the model to generate the cross sections was approximately 0.01% of the time needed to execute the cross-section processing module.

Multi-physics coupling simulation of small mobile nuclear reactor with finite element-based models

This paper describes the multi-physics coupling simulation of a small transportable nuclear reactor, SIMONS, using the neutronics/thermo-mechanics coupling three-dimensional high-fidelity code platform FEMAS (FEM-based Multiphysics Analysis Software for Nuclear Reactor). FEMAS builds multi-physics field models utilizing the open-source finite element codes deal.II and FEniCS. Using coupling simulations of neutronics, thermal-conduction, and thermo-mechanics, the SIMONS micro mobile reactor is examined. For the generation of cross-sections libraries, the open-source Monte Carlo code OpenMC is utilized, and the (n,xn) correction and SPH method is developed to maintain neutron equilibrium and eliminate homogenization errors. The approaches yield neutronic solutions that are in good agreement with the reference solutions, as the keff error decreases from 5066 pcm to 7.2 pcm and the maximum relative power error decreases from 5.2% to 0.3%. With a thermal power of 240 kW, the local temperature difference of the entire reactor is 354.68 K, and the deformation rate due to thermal stress is 2.4%, resulting in a reactivity feedback of 4459.6 pcm. At the wall in contact with the outermost fuel rods, where the temperature gradient grows substantially and thermal stress rises, the monolith stress is measured to peak at 9.65 MPa. This research has paved the way for a more thorough multi-physics examination of micro nuclear reactors in the era of high-precision computation.

Applying approximate Bayesian computation to reduce uncertainty in multigroup 235-U cross-sections using ICSBEP experimental data

The multiplication factor (keff) and its uncertainty are critical design parameters in nuclear reactors. The keff uncertainty must be considered for operation, safety, and economic reasons. Consequently, reducing uncertainty in keff has been of interest to the nuclear industry for as long as nuclear reactors were designed. Two methods of reducing this uncertainty are in use – Generalized Linear Least Squares (GLLS) and A General Monte Carlo-Bayes Procedure for Improved Predictions of Integral Functions of Nuclear Data (MOCABA). Both methods have the ability to reduce the keff uncertainty by reducing uncertainty in cross-sections through the assimilation of measured critical systems’ or nuclear reactors’ operational data. The algorithms require means of simulating the experimental data and that the calibrated parameters are described by specific statistical distributions. GLLS is limited to linear models and multivariate normal prior and posterior, while MOCABA can use any (non-linear) model but is also limited to multivariate normal prior and posterior. This work implements a universal and rigorous algorithm called Sequential Monte Carlo – Approximate Bayesian Computation (SMC-ABC) for the same application. It is found that SMC-ABC, GLLS, and likely MOCABA give essentially the same results for the same problems. The paper also presents a reliable method for validating the results based on so-called “synthetic experiments”. Synthetic experiments are computationally generated data used in place of experiments. The method allows for verifying whether there is a risk of overfitting the calibrated parameters (cross-sections in case of this work) during the data assimilation. This novel method shows the potential of using SMC-ABC and other Bayesian calibration methods to generate improved general-purpose neutron cross-section libraries.

Measurement of Photoneutrons from Depleted Uranium and Comparison Study Using MCNPX-PoliMi

High-energy photon interrogation is a nondestructive technique that is used to detect special nuclear materials and characterize nuclear waste. The development of such systems is complex and requires Monte Carlo simulations to optimize system performance. Monte Carlo simulations rely on various scattering, absorption, and photonuclear cross-section data. While the scattering and absorption cross-section data have been extensively studied and validated with experiments, the results obtained from photonuclear simulations are often found to underpredict measured results, indicating uncertainties in the photonuclear cross sections themselves. Thus, there is a need for new measured results that can be used to quantify underpredictions in simulations using photonuclear cross-section data. In the present work, we interrogated depleted uranium with a 9-MV electron linac and detected photoneutrons with trans-stilbene organic scintillators. The measurement of photoneutrons with organic scintillators is challenging due to the presence of the intense photon flux, which causes issues such as pulse pile-up, detector saturation, and poor signal-to-background ratio. To mitigate these challenges, we used iron and polyethylene shielding of varying thicknesses around the depleted uranium target and a neural network–based digital pulse processing algorithm to recover neutron and photon information from piled-up events. Our goal was to compare the measured photoneutron count rate with the simulated rate obtained using the MCNPX-PoliMi transport code. For a light output window of 0.28 to 2.67 MeVee (1.66- to 6.85-MeV proton recoil energy), we found that the simulated count rate obtained using the ENDF/B-VII photonuclear cross-section library underpredicts the measured rate by 32.8% ± 3.2%. Additionally, we compared the simulated and measured photoneutron light output distributions. For the least thicknesses of shielding, the simulation was found to underpredict measurements in the 0.70- to 2.67-MeVee light output window. For the greatest thicknesses of shielding, the simulation was found to underpredict the measurement across the entire light output window of 0.28 to 2.67 MeVee.

Neutronic Analysis of the RSG-GAS Fuel Using Burnable Poison

Control and safety of nuclear reactors are significantly influenced by the determination of safety parameters. The three most crucial safety factors for assessing reactor status are the infinite multiplication factor, reactivity coefficients, and power peaking factor. The objective of the present study is to examine how the RSG-GAS fuel safety parameters behave in a typical reactor operation state. A lattice cell fuel model of the fuel lattice of the RSG-GAS reactor core was modeled using WIMSD-5Bwith cross-section library data based on ENDF/B-VIII.0. The value of the infinite multiplication factor with various burnable poison concentrations, as well as the moderator and fuel temperatures, were the variables that were examined. The reactivity coefficient parameters were similarly analyzed. By comparing the WIMSD-5B code results with information from the SAR document, the WIMS model for RSG-GAS fuel was verified, and it was inferred that the parameters are in good agreement. Safe behavior uses the predicted reactivity coefficient values as an example.

Depletion chain simplification using pseudo-nuclides

This paper introduces a method for simplifying the depletion chains and reducing the number of nuclides tracked in a given nuclear system while adding pseudo-nuclides in order to preserve decay heat. Pseudo-nuclides are demonstrated to preserve the decay heat effects of fission product irradiation with fewer nuclides tracked overall than decay heat precursors. The functionality is implemented and tested in the reactor physics application Griffin. Extensive testing is performed using the comprehensive ENDF-B/VIII.0 decay data and microscopic cross-section libraries from ORIGEN and MC2-3. The impact on the number of nuclides that must be tracked based on effects such as changing cross-section data, changing energy group structure, changing neutron flux spectrum, and tracking decay energy release is explored in this work. Overall, the addition of pseudo-nuclides allows the number of nuclides tracked to be reduced from 1,700 to 400 or fewer nuclides, based on the criteria used in this work.

Characterization of Fast Neutron Transmission Through an Iron Shield

In this paper we give an analysis of the neutron transmission through an iron sphere using Monte Carlo and transport theory methods based on ENDF/B-VII.1 general purpose library. The motivation for this investigation comes from a well-known deficiency in the iron inelastic data from the older library evaluation (ENDF/B-V), giving a concern for a fast neutron flux underestimation within the reactor pressure vessels. In order to benchmark the next-generation ENDF/B-VI iron data, the U.S. Nuclear Regulatory Commission and the former Czechoslovakian National Research Institute have jointly preformed several experiments in 1990s, addressing neutron leakage spectra obtained for a 252Cf fission source in a centre of an iron sphere. It was shown that the ENDF/B-VI iron cross section, containing several improvements over previous evaluations, will not entirely resolve the neutron spectrum discrepancies observed at high neutron energies. Since safety analyses of reactor pressure vessel embrittlement are often based on neutron transport calculations using specific multigroup cross section libraries, simulation of this benchmark was performed using a hybrid shielding methodology of ADVANTG3.0.3 and MCNP6.1.1b codes. Comparison of calculated and referenced dosimeter activation rates are presented for several "standard" nuclear reactions, often used in reactor pressure vessel dosimetry. For that purpose, the new IRDFF-II special library from the IAEA Nuclear Data Services was used as a reference source of dosimetry cross sections. The MCNP6.1.1b code was used for calculation of reaction rates, which were also compared with previous IRDFF-1.05 special library and general purpose ENDF/B-VII.1 library.

Characterization of the GBC-32 Fuel Assembly Source Terms

This paper presents burnup/depletion calculations of the typical Westinghouse 17x17 fuel assembly to be used as a radioactive waste package in a Generic Burnup Credit cask benchmark problem with 32 elements (GBC-32). This first phase is addressing spent fuel source terms calculation while evaluation of the shielding performance of the GBC-32 cask is planned for the second phase. The TRITON-NEWT methodology of the SCALE6.1.3 program package was used in a tandem with ORIGEN-S code for deterministic 2D calculation of the GBC-32 fuel assembly neutron multiplication factor, providing spatial-temporal fluxes and isotopic concentration change. The burnup simulation was done up to 60 GWd/tU with sensitivity analysis of relevant physical parameters influenced by the working cross-section library. This approach also allowed generation of the specific user-defined collapsed cross-section libraries as a function of fuel enrichment and burnup level. Calculation of isotopic concentrations, decay heat, neutron-gamma spectra and major actinides activity for different fuel assembly cooling periods was performed using ORIGEN-ARP module.

Comparison of MCNPX and EGSnrc Monte Carlo Codes in the Calculation of Nano-Scaled Absorbed Doses and Secondary Electron Spectra around Clinically Relevant Nanoparticles

Purpose: Absorbed dose enhancement due to the presence of high atomic number Nanoparticles (NP)s has been estimated and modeled by Monte Carlo (MC) simulation methods. In the current study, two MC codes of Monte Carlo N‐Particle eXtended (MCNPX) and EGSnrc codes were compared by calculation of secondary electron energy spectra and nano-scaled dose values around four types of spherical NPs.
 Materials and Methods: The MC model was composed of a spherical nanoparticle with a diameter of 50 nm and mono-energetic sources of photons with energies of 30,60, and 100 keV. The secondary electrons emitted from the nanoparticle were scored on the nanoparticle surface and the delivered dose to water around the nanoparticle was tallied using concentric shells with a thickness of 25 nm. Four different elements were used as materials of NPs, including Gold, Bismuth, Gadolinium, and Hafnium.
 Results: Our results showed a considerable difference in the number of emitted electrons per incident photon between the two codes. There were also discrepancies between the two codes in the energy spectra of secondary electrons. Calculated radial dose values around NPs in nano-scale had a similar pattern for both codes. However, significant differences existed for some elements.
 Conclusion: It can be concluded that the results of nano-scaled MC modeling for nanoparticle-based radiation therapy are dependent on the code type and its algorithm for electron transport as well as exploited cross-section libraries.

Extended one-group cross-section libraries in precision nuclide composition calculations of irradiated steel

Extended one-group cross-section libraries in precision nuclide composition calculations of irradiated steel

Steady-state analysis of ROSTOV-2 benchmark problem using SERPENT-2 and PARCS codes

Recently, a new benchmark problem has been introduced by the OECD/NEA on the Rostov-2 reactor to address the verification and validation of novel High-Fidelity Multi-Physics tools. In line with the first phase of this benchmark, the current research was dedicated to the steady state reactor physics analysis, using “traditional multi-physics” codes for further and future comparisons. A set of homogenized two-group cross-section libraries were developed using SERPENT-2 Monte Carlo code. Some sophisticated techniques have been employed for detailed consideration of radial as well as axial reflectors, spacer grids and assembly discontinuity factor evaluation. Since PARCS code requires the group constants in PMAXS format, the generated libraries were re-formatted using the GenPMAXS interface. The PMAXS library verification, as well as the steady state analysis have been conducted in three operational states of the reactor using PARCS nodal diffusion code, independent SERPENT-2 Monte Carlo simulation and some available reference data. A set of criticality calculations have been conducted to evaluate the total reactivity worth of different control rod groups, the group #10 integral reactivity curve, fuel temperature coefficient and the critical boron concentration in Hot Zero Power (HZP) and Hot Full Power (HFP) states. The maximum reactivity difference between PARCS and SERPENT-2 simulations was less than 100 pcm in all criticality calculations. In HFP state, the relative power distribution in radial and axial directions were calculated and compared. The maximum Power Peaking Factors (PPFs) and the location of the “hot assembly” from PARCS and SERPENT-2 are in good agreement with each other. Finally, the burn-up calculation was conducted with PARCS code to evaluate neutronic parameters at some specific burn-up values. The calculated reactor physics parameters including the critical boron concentration, the power peaking factors, the power and burn-up distributions at the Beginning of Experiment (BOE) were consistent with available reference data.