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Advanced Numerical Modelling Techniques for Nuclear Reactors

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A special issue of Energies (ISSN 1996-1073). This special issue belongs to the section "B4: Nuclear Energy".

Deadline for manuscript submissions: closed (31 July 2022) | Viewed by 20168

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Special Issue Editors

Dr. Victor Hugo Sánchez Espinoza

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Guest Editor

Karlsruhe Institute of Technology (KIT), Institute of Neutron Physics and Reactor technology (INR), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany
Interests: nuclear engineering; reactor safety; numerical safety assessment methods for reactors of Gen II and III including SMRs; computational reactor physics; safety-related thermal hydraulics; uncertainty quantification of numerical codes

Dr. Andrea Zoia

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Guest Editor

Université Paris-Saclay, CEA, Service d'Etudes des Réacteurs et de Mathématiques Appliquées, 91191 Gif-sur-Yvette, France
Interests: Monte Carlo methods for particle transport; nuclear engineering; computational reactor physics; Eigenvalue problems; variance reduction methods; perturbations and sensitivities

Special Issue Information

Dear Colleagues,

Nuclear reactor analysis is experiencing rapid development driven by the cheap and widely available computational power of high performance computer (HPC) clusters, the increased use of Monte Carlo methods, the development of new deterministic parallelized transport solvers, the upgrade of subchannel thermal hydraulic codes for full core analysis at the subchannel level, the rapid spread of the CFD analysis of safety-related reactor and core analysis, and—last but not least—the modernization of fuel performance codes, including the emergence of first-principles thermo-mechanic solvers. Today, the focus of reactor and core analysis is the development of multi-physics and multiscale flexible and modular coupling approaches among the different solvers. In this context, the validation of advanced numerical tools for core and reactor analysis based on experimental or plant data is primordial for their acceptance and use in industry-like applications. The ultimate goal is to increase the spatial resolution and prediction accuracy, allowing a reduction in safety margins.

We invite the nuclear community (academia, research institutions, industry, regulators, and TSO) to submit original and unpublished manuscripts to this Special Issue that is focused on recent developments relevant to “Core and Reactor Analysis”.

The goal of the Special Issue is to publish the most recent research results on computational reactor physics and analysis relevant for design optimization and safety evaluation. The topics suited for this Special Issue include, but are not limited to:

New deterministic transport solvers for pin-based core analysis (static and transients);
Advances in Monte Carlo methods for the analysis of core transients, e.g., REA;
New methods for the detailed multi-physics analysis of core depletion (neutronic, thermal hydraulic, and thermo-mechanics);
Multiscale reactor analysis method based on the coupling of different thermal hydraulic solvers (CFD, subchannel, and system);
New reactor physical methods for the improved prediction of safety parameters based on different transport approximations (diffusion, SP3, MOC, etc.);
Applications of the methods and code to LWR including SMR.

Dr. Victor Hugo Sánchez Espinoza
Dr. Andrea Zoia
Guest Editors

Manuscript Submission Information

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Keywords

Core analysis
Multi-physics
Monte Carlo
Deterministic solvers
Multiscale
HPC
Neutronics
Thermal hydraulics
Thermo-mechanics
LWR
SMR

Published Papers (10 papers)

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Research

15 pages, 3216 KiB

Open AccessArticle

Transmutation and Breeding Performance Analysis of Molten Chloride Salt Fast Reactor Using a Fuel Management Code with Nodal Expansion Method

by Kaicheng Yu, Maosong Cheng, Xiandi Zuo and Zhimin Dai

Energies 2022, 15(17), 6299; https://doi.org/10.3390/en15176299 - 29 Aug 2022

Cited by 3 | Viewed by 1567

Abstract

The transmutation of transuranic (TRU) elements produced by pressurized water reactors (PWRs) can effectively reduce their radioactive hazards. The molten chloride salt fast reactor (MCSFR) is a type of liquid-fueled molten salt reactor (MSR) using fuel in the form of molten chloride salts. The MCSFR utilizing a fast neutron spectrum and high actinide fraction is considered to be a potential reactor type for TRU transmutation. An online refueling and reprocessing scenario is the unique feature of liquid-fueled MSRs. On account of this characteristic, a new fuel management code named ThorNEMFM with a nodal expansion method (NEM) was developed and validated with the molten salt breeder reactor (MSBR) and the molten salt fast reactor (MSFR) benchmarks. Then, the transmutation and breeding performances of the MCSFR were simulated and analyzed with the ThorNEMFM code. The MCSFR adopts TRU elements as initial fissile loads and online feeding fissile materials. The results show that the transmutation ratio of TRU elements in the MCSFR can reach 50%, and the breeding ratio can reach 1.359. Moreover, the MCSFR has low radiotoxicity due to lower buildup of fission products (FPs). Full article

(This article belongs to the Special Issue Advanced Numerical Modelling Techniques for Nuclear Reactors)

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25 pages, 1658 KiB

Open AccessFeature PaperArticle

A Detailed Analysis of the H.B. Robinson-2 Reactor Pressure Vessel Dosimetry Benchmark

by Romain Vuiart, Mariya Brovchenko, Julien Taforeau and Eric Dumonteil

Energies 2022, 15(14), 5098; https://doi.org/10.3390/en15145098 - 12 Jul 2022

Viewed by 2331

Abstract

The operation of many nuclear pressurized water reactors is being extended beyond their design lifetime limit. From the perspective of possible further lifetime extension, safety requirements are a priority. Therefore, the quantification of the neutron irradiation embrittlement of the reactor pressure vessel (RPV) is an important issue, as this is a guiding parameter that influences the reactor lifetime. In this context, the Institut de Radioprotection et de Sûreté Nucléaire developed a calculation scheme for the analysis of RPV aging under neutron irradiation, named VACS (vessel aging calculation scheme). VACS couples a deterministic approach (CASMO5 and SIMULATE5) to evaluate the full-core fission neutron source term and a Monte Carlo modeling (MCNP6) approach to model the neutron attenuation from the core to sites of interest (RPV, surveillance capsules, etc.). To ensure the reliability of aging predictions, this paper describes a detailed analysis of the neutron H.B. Robinson-2 reactor pressure vessel dosimetry benchmark. The results indicate that VACS shows satisfactory accuracy when the ENDF-B/VII.1 or JEFF-3.3 nuclear data libraries are used in the attenuation calculation. However, the use of ENDF-B/VIII.0 leads to significantly worse results. Full article

(This article belongs to the Special Issue Advanced Numerical Modelling Techniques for Nuclear Reactors)

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18 pages, 11321 KiB

Open AccessArticle

An Advanced TRACE Modeling Approach: Automatic Connection of 3D Cartesian and Cylindrical VESSEL Components in Integral Plant Models

by Kanglong Zhang and Victor Hugo Sanchez-Espinoza

Energies 2022, 15(12), 4384; https://doi.org/10.3390/en15124384 - 16 Jun 2022

Viewed by 1363

Abstract

Best estimate system thermal-hydraulic codes in the nuclear engineering community, e.g., TRACE, RELAP3D, CATHARE-3, etc., were extended with 3D coarse-mesh components to better describe the 3D Thermal-Hydraulic (TH) phenomena taking place within the Reactor Pressure Vessel (RPV) and the core. The RPV is usually shaped like a cylinder while the core is mostly a cube. Hence, the TRACE code is equipped with a Cylindrical VESSEL and a Cartesian VESSEL. The former one is to represent the RPV (including core), pressurizer, and steam generator. The latter one is more appropriate to represent the core. The two components are connected by two Vessel-Junctions (VJ) at the core inlet and outlet. Due to the different nodalization between the two VESSELs, the analyst needs to do repetitive and error-prone work defining the cell-to-cell junctions and their TH parameters. To facilitate this process, the Karlsruhe Institute of Technology (KIT) has developed an automatic approach based on a mesh-constructing and field-mapping library, namely the MEDCoupling. These new capabilities of TRACE are demonstrated by the analysis of the coolant mixing for an academic case and the AP1000 reactor. Full article

(This article belongs to the Special Issue Advanced Numerical Modelling Techniques for Nuclear Reactors)

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28 pages, 15780 KiB

Open AccessArticle

Advanced Couplings and Multiphysics Sensitivity Analysis Supporting the Operation and the Design of Existing and Innovative Reactors

by Barbara Calgaro and Barbara Vezzoni

Energies 2022, 15(9), 3341; https://doi.org/10.3390/en15093341 - 4 May 2022

Cited by 2 | Viewed by 1707

Abstract

Codes and methods are subjected to a continuous update process for answering the regulatory requirements concerning the long-term operation of existing reactors and new concept deployment. In this continuous improvement process, new generation codes are developed for supporting industrial applications and the long-term strategy. In this paper, attention is devoted to selecting codes under development in France for lattice and core steady-state and transient calculations. These codes, APOLLO3^® and CATHARE3, have been selected for carrying out the activities of the H2020 CAMIVVER Project oriented to the 3D-multiphysics couplings improvements. Multiscale and multiphysics solutions are key topics to keep competitiveness and answer to newer industrial needs in plant operations, licensing, and safe operating envelope requirements. The paper presents an overview of the activities performed by Framatome to support the definition of benchmarks exercises proposed in the CAMIVVER Project. Small core configurations subjected to a Reactivity Insertion Accident (RIA) are presented with associated preliminary results. To open the discussions toward the development of Best-Estimate Plus Uncertainties (BEPU) solutions, the URANIE statistical platform was used for sensitivity analysis over different configurations. These preliminary results are also presented. Full article

(This article belongs to the Special Issue Advanced Numerical Modelling Techniques for Nuclear Reactors)

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17 pages, 38185 KiB

Open AccessArticle

Full Core Pin-Level VVER-440 Simulation of a Rod Drop Experiment with the GPU-Based Monte Carlo Code GUARDYAN

by David Legrady, Gabor Tolnai, Tamas Hajas, Elod Pazman, Tamas Parko and Istvan Pos

Energies 2022, 15(8), 2712; https://doi.org/10.3390/en15082712 - 7 Apr 2022

Cited by 2 | Viewed by 2097

Abstract

Targeting ultimate fidelity reactor physics calculations the Dynamic Monte Carlo (DMC) method simulates reactor transients without resorting to static or quasistatic approximations. Due to the capability to harness the computing power of Graphics Processing Units, the GUARDYAN (GpU Assisted Reactor DYnamic ANalysis) code has been recently upscaled to perform pin-by-pin simulations of power plant scale systems as demonstrated in this paper. A recent rod drop experiment at a VVER-440/213 (vodo-vodyanoi enyergeticheskiy reaktor) type power plant at Paks NPP, Hungary, was considered and signals of ex-core detectors placed at three different positions were simulated successfully by GUARDYAN taking realistic fuel loading, including burn-up data into account. Results were also compared to the time-dependent Paks NPP in-house nodal diffusion code VERETINA (VERONA: VVER Online Analysis and RETINA: Reactor Thermo-hydraulics Interactive). Analysis is given of the temporal and spatial variance distribution of GUARDYAN fuel pin node-wise power estimates. We can conclude that full core, pin-wise DMC power plant simulations using realistic isotope concentrations are feasible in reasonable computing times down to 1–2% error of ex-core detector signals using current GPU (Graphics Processing Unit) High Performance Computing architectures, thereby demonstrating a technological breakthrough. Full article

(This article belongs to the Special Issue Advanced Numerical Modelling Techniques for Nuclear Reactors)

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22 pages, 6145 KiB

Open AccessArticle

Improvement and Validation of the System Analysis Model and Code for Heat-Pipe-Cooled Microreactor

by Li Ge, Huaqi Li, Xiaoyan Tian, Zeyu Ouyang, Xiaoya Kang, Da Li, Jianqiang Shan and Xinbiao Jiang

Energies 2022, 15(7), 2586; https://doi.org/10.3390/en15072586 - 1 Apr 2022

Cited by 9 | Viewed by 2891

Abstract

Heat-pipe-cooled microreactors (HPMR) use a passive high-temperature alkali metal heat pipe to directly transfer the heat of solid core to the hot end of the intermediate heat exchanger or thermoelectric conversion device, thus avoiding a single point failure. To analyze and evaluate the transient safety characteristics of an HPMR system under accident conditions, such as heat pipe failure in the core or a loss of system heat sink and other accidents, a previously developed model for transient analysis of a heat-pipe-cooled space nuclear reactor power system (HPSR) was improved and validated in this study. The models improved mainly comprise: (1) An entire 2-D solid-core heat transfer model is established to analyze the accident conditions of core heat pipe failure and system heat sink loss. In this model, radial and axial Fourier heat conduction equations are used to divide the core into r-θ direction control volumes. The physical parameters of the material in the control volume are calculated according to the volume-weighted average. (2) By coupling the heat transfer limit model and the two-dimensional thermal resistance network model, the transient model of a heat pipe for HPMR system analysis is improved. (3) Conversion system models are established to simulate the system characteristics of the advanced HPMR concept, such as thermoelectric conversion, Stirling conversion, and the open Brayton conversion analysis model. Based on the improved models, the HPMR system analysis program TAPIRSD was developed, which was verified by experimental data of the separated conversion components and the ground nuclear test device KRUSTY. The maximum deviation of the power output predicted by the energy conversion model is less than 8%. The accident conditions of the KRUSTY tests, such as load change, core heat pipe failure, and heat sink loss accident, were studied by using TAPIRSD. The results show that the simulation results of the TAPIRSD code agree well with the experimental data of the KRUSTY prototype reactor. The maximum error between the TAPIRSD code prediction and the measured value of the core temperature under accident conditions is less than 10 K, and the maximum deviation is less than 2%. The results show that the developed code can predict the transient response process of the HPMR system well. At the same time, the accuracy and reliability of the improved model are proved. The TAPIRSD is suitable for system transient analysis of different types of HPMRs and provides an optional tool for the system safety characteristics analysis of HPMR. Full article

(This article belongs to the Special Issue Advanced Numerical Modelling Techniques for Nuclear Reactors)

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16 pages, 877 KiB

Open AccessArticle

Verification of the Parallel Transport Codes Parafish and AZTRAN with the TAKEDA Benchmarks

by Julian Duran-Gonzalez, Victor Hugo Sanchez-Espinoza, Luigi Mercatali, Armando Gomez-Torres and Edmundo del Valle-Gallegos

Energies 2022, 15(7), 2476; https://doi.org/10.3390/en15072476 - 28 Mar 2022

Viewed by 1488

Abstract

With the increase in computational resources, parallel computation in neutron transport codes is inherent since it allows simulations with high spatial-angular resolution. Among the different methodologies available for the solution of the neutron transport equation, spherical harmonics (

P_{N}

) and discrete-ordinates [...] Read more.

P_{N}

) and discrete-ordinates (

S_{N}

) approximations have been widely used, as they are established classical methods for performing nuclear reactor calculations. This work focuses on describing and verifying two parallel deterministic neutron transport codes under development. The first one is the Parafish code that is based on the finite-element method and

P_{N}

approximation. The second one is the AZTRAN code, based on the RTN-0 nodal method and

S_{N}

approximation. The capabilities of these two codes have been tested on the TAKEDA benchmarks and the results obtained show good behavior and accuracy compared to the Monte Carlo reference solutions. Additionally, the speedup obtained by each code in the parallel execution is acceptable. In general, the results encourage further improvement in the codes to be comparable to other well-validated deterministic transport codes. Full article

(This article belongs to the Special Issue Advanced Numerical Modelling Techniques for Nuclear Reactors)

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20 pages, 11258 KiB

Open AccessArticle

High-Fidelity Steady-State and Transient Simulations of an MTR Research Reactor Using Serpent2/Subchanflow

by Juan Carlos Almachi, Víctor Hugo Sánchez-Espinoza and Uwe Imke

Energies 2022, 15(4), 1554; https://doi.org/10.3390/en15041554 - 19 Feb 2022

Cited by 8 | Viewed by 2125

Abstract

In order to join efforts to develop high-fidelity multi-physics tools for research reactor analysis, the KIT is conducting studies to modify the coupled multi-physics codes developed for power reactors. The coupled system uses the Monte Carlo Serpent 2 code for neutron analysis and the Subchanflow code for thermo-hydraulic analysis. Serpent treats temperature dependence using the target motion sampling method and Subchanflow was previously extended and validated with experimental data for plate-type reactor analysis. This work present for the first time the steady-state and transient neutron and thermo-hydraulic analysis of an MTR core defined in the IAEA 10 MW benchmark using Serpent2/Subchanflow. Important global and local parameters for nominal steady-state conditions were obtained, e.g., the lowest and highest core plate/channel power/temperature, the radial and axial core power profile at the plate level, and the core coolant temperature distribution at the subchannel level. The capabilities of Serpent2/Subchanflow to perform transient analysis with on-the-fly motion of the control plates were tested, namely with fast and slow reactivity insertion. Based on the unique results obtained for the first time at the subchannel and plate level, it can be stated that the coupled Serpent2/Subchanflow code is a very promising tool for research reactor safety-related investigations. Full article

(This article belongs to the Special Issue Advanced Numerical Modelling Techniques for Nuclear Reactors)

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13 pages, 9667 KiB

Open AccessArticle

Overview of SERMA’s Graphical User Interfaces for Lattice Transport Calculations

by Daniele Tomatis, François Bidault, Adrien Bruneton and Zarko Stankovski

Energies 2022, 15(4), 1417; https://doi.org/10.3390/en15041417 - 15 Feb 2022

Cited by 6 | Viewed by 1480

Abstract

This article presents an overview of the graphical user interfaces (GUIs) developed at CEA/SERMA (Service d’Études des Réacteurs et de Mathématiques Appliquées) in Saclay, France, which have been used for over forty years by engineers and scientists to build geometries and meshes for general-purpose lattice transport calculations (neutrons and photons). Several applications make use of these calculations, from fuel assembly to full core design, criticality and safety, needing consistency check of the geometry and input properties before starting any lattice calculation. The software pattern design of the GUIs is briefly discussed, showing also the rationale behind the two interfaces for the construction of the geometries for simple fuel assemblies and complex motifs including the reflector (colorsets). The new GUI, ALAMOS, specifically developed for APOLLO3^® with a Python Application Programming Interface (API), is here presented as the successor of Silène, which was the first GUI released in the 1990s to serve APOLLO2 calculations. The considerable experience gained by Silène over the years with plenty of various applications has provided a crucial support for the development of ALAMOS. Full article

(This article belongs to the Special Issue Advanced Numerical Modelling Techniques for Nuclear Reactors)

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17 pages, 1940 KiB

Open AccessArticle

Methodology for Discontinuity Factors Generation for Simplified P₃ Solver Based on Nodal Expansion Formulation

by Yuchao Xu, Jason Hou and Kostadin Ivanov

Energies 2021, 14(20), 6478; https://doi.org/10.3390/en14206478 - 10 Oct 2021

Viewed by 1454

Abstract

The Simplified Spherical Harmonic (SP_N) approximation was first introduced as a three-dimensional (3D) extension of the plane-geometry Spherical Harmonic (P_N) equations. A third order SP_N (SP₃) solver, recently implemented in the Nodal Expansion Method (NEM), has shown promising performance in the reactor core neutronics simulations. This work is focused on the development and implementation of the transport-corrected interface and boundary conditions in an NEM SP₃ solver, following recent published work on the rigorous SP_N theory for piecewise homogeneous regions. A streamlined procedure has been developed to generate the flux zero and second order/moment discontinuity factors (DFs) of the generalized equivalence theory to minimize the error introduced by pin-wise homogenization. Moreover, several colorset models with varying sizes and configurations are later explored for their capability of generating DFs that can produce results equivalent to that using the whole-core homogenization model for more practical implementations. The new developments are tested and demonstrated on the C5G7 benchmark. The results show that the transport-corrected SP₃ solver shows general improvements to power distribution prediction compared to the basic SP₃ solver with no DFs or with only the zeroth moment DF. The complete equivalent calculations using the DFs can almost reproduce transport solutions with high accuracy. The use of equivalent parameters from larger size colorset models show a slightly reduced prediction error than that using smaller colorset models in the whole-core calculations. Full article

(This article belongs to the Special Issue Advanced Numerical Modelling Techniques for Nuclear Reactors)

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