Research

24 pages, 15249 KiB

Open AccessArticle

Analysis of Rostov-II Benchmark Using Conventional Two-Step Code Systems

by Jaerim Jang, Mathieu Hursin, Woonghee Lee, Andreas Pautz, Marianna Papadionysiou, Hakim Ferroukhi and Deokjung Lee

Energies 2022, 15(9), 3318; https://doi.org/10.3390/en15093318 - 02 May 2022

Cited by 6 | Viewed by 1651

Abstract

This paper presents the steady state analysis of the Rostov-II benchmark using the conventional two-step approach. It involves the STREAM/RAST-K and CASMO-5/PARCS code systems. This paper documents a comprehensive code-to-code comparison between Serpent 2, CASMO-5, and STREAM at the lattice level for the [...] Read more.

This paper presents the steady state analysis of the Rostov-II benchmark using the conventional two-step approach. It involves the STREAM/RAST-K and CASMO-5/PARCS code systems. This paper documents a comprehensive code-to-code comparison between Serpent 2, CASMO-5, and STREAM at the lattice level for the different fuel assemblies (FAs) loaded in the Rostov-II core; and between Serpent 2, PARCS, and RAST-K at the core level in 2D. Finally, the 3D results of both deterministic models are compared to the steady state measurements of the Rostov-II benchmark. With respect to the measurements available in the Rostov-II benchmark, comparable accuracy (30 ppm difference in boron concentration, 2% assembly power) with an industrial calculation scheme (BIPR8) are reported up to 36.73 EFPDs. The calculations reported in the paper showed that the modeling of the resonance self-shielding in the lattice code as well as the geometrical modeling of the reflector are key for an accurate solution (reducing the in-out power tilt). At the core simulator level, a fairly crude 1D reflector model appears to be enough. Overall, this paper provides the detailed models and conditions used in STREAM/RAST-K and CASMO-5/PARCS, and accurate calculation solution for the Rostov-II benchmark with STREAM/RAST-K and CASMO-5/PARCS compared with measurement. Full article

(This article belongs to the Special Issue Computational Techniques of Nuclear Reactor Physics)

► Show Figures

Figure 1

12 pages, 7446 KiB

Open AccessArticle

Reactivity Effect Evaluation of Fast Reactor Based on Angular-Dependent Few-Group Cross Sections Generation

by Xianan Du, Xuewen Wu, Youqi Zheng and Yongping Wang

Energies 2021, 14(13), 4042; https://doi.org/10.3390/en14134042 - 04 Jul 2021

Cited by 1 | Viewed by 1869

Abstract

Among all the possible occurring reactivity effects of a fast reactor, the situations whereby the control rod was inserted, or the coolant was voided could lead to strong anisotropy of neutron flux distribution, therefore the angular dependence on neutron flux should be considered [...] Read more.

Among all the possible occurring reactivity effects of a fast reactor, the situations whereby the control rod was inserted, or the coolant was voided could lead to strong anisotropy of neutron flux distribution, therefore the angular dependence on neutron flux should be considered during the few-group cross-sections generation. Therefore, the purpose of this paper is to compare the influence whether the angular dependence on neutron flux is considered in the calculation of few-group cross sections for the reactivity effect calculation. In the study, the 1-D SN finite difference neutron transport equation solver was implemented in the TULIP of SARAX code system so that the high-order neutron flux could be obtained. Meanwhile, the improved Tone’s method was also applied. The numerical results were obtained based on three experimental FR cores, the JOYO MK-I core, ZPPR-9 core, and ZPPR-10B core. Both control rod worth and sodium void reactivity were calculated and compared with the measurement data. By summarizing and comparing the results of 46 cases, significant differences were found between different consideration of the neutronic analysis. The consideration of angular dependence on neutron flux distribution in the few-group cross-sections generation was beneficial to the neutronic design analysis of FR, especially for the reactivity effect calculation. Full article

(This article belongs to the Special Issue Computational Techniques of Nuclear Reactor Physics)

► Show Figures

Figure 1

28 pages, 4771 KiB

Open AccessArticle

Verification of MPACT for the APR1400 Benchmark

by Kaitlyn Elizabeth Barr, Sooyoung Choi, Junsu Kang and Brendan Kochunas

Energies 2021, 14(13), 3831; https://doi.org/10.3390/en14133831 - 25 Jun 2021

Cited by 1 | Viewed by 2339

Abstract

This paper describes benchmark calculations for the APR1400 nuclear reactor performed using the high-fidelity deterministic whole-core simulator MPACT compared to reference solutions generated by the Monte Carlo code McCARD. The methodology presented in this paper is a common approach in the field of [...] Read more.

This paper describes benchmark calculations for the APR1400 nuclear reactor performed using the high-fidelity deterministic whole-core simulator MPACT compared to reference solutions generated by the Monte Carlo code McCARD. The methodology presented in this paper is a common approach in the field of nuclear reactor analysis, when measured data are not available for comparison, and may be more broadly applied in other simulation applications of energy systems. The benchmark consists of several problems that span the complexity of single pins to a hot full power cycle depletion. Overall, MPACT shows excellent agreement compared to the reference solutions. MPACT effectively predicts the reactivity for different geometries and several temperature and boron conditions. The largest deviation from McCARD occurs for cold zero conditions in which the fuel, moderator, and cladding are all 300 K. Possible reasons for this are discussed. Excluding these cases, the

ρ

reactivity difference from McCARD is consistently below 100 pcm. For single fuel pin problems, the highest error of 151 pcm occurs for the lowest fuel enrichment of 1.71 wt.% UO

_{2}

, indicating possible, albeit small, enrichment bias in MPACT’s cross-section library. Furthermore, MOC and spatial mesh parametric studies indicate that default meshing parameters and options yield results comparable to finely meshed cases. Additionally, there is very good agreement of the radial and axial power distributions. RMS radial pin and assembly power differences for all cases are at or below 0.75%, and all RMS axial power differences are below 1.65%. These results are comparable to previous results from the VERA progression problems benchmark and meet generally accepted accuracy criteria for whole-core transport codes. Full article

(This article belongs to the Special Issue Computational Techniques of Nuclear Reactor Physics)

► Show Figures

Figure 1

21 pages, 1012 KiB

Open AccessArticle

Methodology for Sensitivity Analysis of Homogenized Cross-Sections to Instantaneous and Historical Lattice Conditions with Application to AP1000^® PWR Lattice

by Dean Price, Thomas Folk, Matthew Duschenes, Krishna Garikipati and Brendan Kochunas

Energies 2021, 14(12), 3378; https://doi.org/10.3390/en14123378 - 08 Jun 2021

Cited by 1 | Viewed by 1833

Abstract

In the two-step method for nuclear reactor simulation, lattice physics calculations are performed to compute homogenized cross-sections for a variety of burnups and lattice configurations. A nodal code is then used to perform full-core analysis using the pre-calculated homogenized cross-sections. One source of [...] Read more.

In the two-step method for nuclear reactor simulation, lattice physics calculations are performed to compute homogenized cross-sections for a variety of burnups and lattice configurations. A nodal code is then used to perform full-core analysis using the pre-calculated homogenized cross-sections. One source of uncertainty introduced in this method is that the lattice configuration or depletion conditions typically do not match a pre-calculated one from the lattice physics simulations. Therefore, some interpolation model must be used to estimate the homogenized cross-sections in the nodal code. This current study provides a methodology for sensitivity analysis to quantify the impact of state variables on the homogenized cross-sections. This methodology also allows for analyses of the historical effect that the state variables have on homogenized cross-sections. An application of this methodology on a lattice for the Westinghouse AP1000^® reactor is presented where coolant density, fuel temperature, soluble boron concentration, and control rod insertion are the state variables of interest. The effects of considering the instantaneous values of the state variables, historical values of the state variables, and burnup-averaged values of the state variables are analyzed. Using these methods, it was found that a linear model that only considers the instantaneous and burnup-averaged values of state variables can fail to capture some variations in the homogenized cross-sections. Full article

(This article belongs to the Special Issue Computational Techniques of Nuclear Reactor Physics)

► Show Figures

Figure 1

14 pages, 2302 KiB

Open AccessArticle

Critical Buckling Generation of TCA Benchmark by the B₁ Theory-Augmented Monte Carlo Calculation and Estimation of Uncertainties

by Ho Jin Park and Jin Young Cho

Energies 2021, 14(9), 2578; https://doi.org/10.3390/en14092578 - 30 Apr 2021

Viewed by 1302

Abstract

The Korea Atomic Energy Research Institute (KAERI) has developed the DeCART2D 2-dimensional (2D) method of characteristics (MOC) transport code and the MASTER nodal diffusion code and has established its own two-step procedure. For design code licensing, KAERI prepared a critical experiment on the [...] Read more.

The Korea Atomic Energy Research Institute (KAERI) has developed the DeCART2D 2-dimensional (2D) method of characteristics (MOC) transport code and the MASTER nodal diffusion code and has established its own two-step procedure. For design code licensing, KAERI prepared a critical experiment on the verification and validation (V&V) of the DeCART2D code. DeCART2D is able to perform the MOC calculation only for 2D nuclear fuel systems, such as the fuel assembly. Therefore, critical buckling in the vertical direction is essential for comparison between the results of experiments and DeCART2D. In this study, the B₁ theory-augmented Monte Carlo (MC) method was adopted for the generation of critical buckling. To examine critical buckling using the B₁ theory-augmented MC method, TCA critical experiment benchmark problems were considered. Based on the TCA benchmark results, it was confirmed that the DeCART2D code with the newly-generated critical buckling predicts the criticality very well. In addition, the critical buckling generated by the B₁ theory-augmented MC method was bound to uncertainties. Therefore, utilizing basic equations (e.g., SNU S/U formulation) linking input uncertainties to output uncertainties, a new formulation to estimate the uncertainties of the newly generated critical buckling was derived. This was then used to compute the uncertainties of the critical buckling for a TCA critical experiment, under the assumption that nuclear cross-section data have uncertainties. Full article

(This article belongs to the Special Issue Computational Techniques of Nuclear Reactor Physics)

► Show Figures

Figure 1

14 pages, 3354 KiB

Open AccessArticle

Extension of Pin-Based Point-Wise Energy Slowing-Down Method for VHTR Fuel with Double Heterogeneity

by Tae-Young Han, Jin-Young Cho, Chang-Keun Jo and Hyun-Chul Lee

Energies 2021, 14(8), 2179; https://doi.org/10.3390/en14082179 - 14 Apr 2021

Cited by 2 | Viewed by 1299

Abstract

For the resonance treatment of a very high temperature reactors (VHTR) fuel with the double heterogeneity, an extension of the pin-based pointwise energy slowing-down method (PSM) was developed and implemented into DeCART. The proposed method, PSM-double heterogeneity (DH), has an improved spherical unit [...] Read more.

For the resonance treatment of a very high temperature reactors (VHTR) fuel with the double heterogeneity, an extension of the pin-based pointwise energy slowing-down method (PSM) was developed and implemented into DeCART. The proposed method, PSM-double heterogeneity (DH), has an improved spherical unit cell model with an explicit tri-structural isotropic (TRISO) model, a matrix layer, and a moderator for reflecting the moderation effect. The moderator volume was analytically derived using the relation of the Dancoff factor and the mean chord length. In the first step, the pointwise homogenized cross-sections for the compact was obtained after solving the slowing down equation for the spherical unit cell. Then, the shielded cross-section for the homogenized fuel compact was generated using the original PSM. The verification calculations were performed for the fuel pins with various packing fractions, compact sizes, TRISO sizes, and fuel temperatures. Additionally, two fuel block problems with very different sizes were examined and the depletion calculation was carried out for investigating the accuracy of the proposed method. They revealed that the PSM-DH has a good performance in the VHTR problems. Full article

(This article belongs to the Special Issue Computational Techniques of Nuclear Reactor Physics)

► Show Figures

Figure 1

17 pages, 1820 KiB

Open AccessArticle

A Multi-Physics Adaptive Time Step Coupling Algorithm for Light-Water Reactor Core Transient and Accident Simulation

by Alexey Cherezov, Jinsu Park, Hanjoo Kim, Jiwon Choe and Deokjung Lee

Energies 2020, 13(23), 6374; https://doi.org/10.3390/en13236374 - 02 Dec 2020

Cited by 7 | Viewed by 2155

Abstract

A new reactor core multi-physics system addresses the pellet-to-cladding heat transfer modeling to improve full-core operational transient and accident simulation used for assessment of reactor core nuclear safety. The rigorous modeling of the heat transfer phenomena involves strong interaction between neutron kinetics, thermal-hydraulics [...] Read more.

A new reactor core multi-physics system addresses the pellet-to-cladding heat transfer modeling to improve full-core operational transient and accident simulation used for assessment of reactor core nuclear safety. The rigorous modeling of the heat transfer phenomena involves strong interaction between neutron kinetics, thermal-hydraulics and nuclear fuel performance, as well as consideration of the pellet-to-cladding mechanical contact leading to dramatic increase in the gap thermal conductance coefficient. In contrast to core depletion where parameters smoothly depend on fuel burn-up, the core transient is driven by stiff equation associated with rapid variation in the solution and vulnerable to numerical instability for large time step sizes. Therefore, the coupling algorithm dedicated for multi-physics transient must implement adaptive time step and restart capability to achieve prescribed tolerance and to maintain stability of numerical simulation. This requirement is met in the MPCORE (Multi-Physics Core) multi-physics system employing external loose coupling approach to facilitate the coupling procedure due to little modification of constituent modules and due to high transparency of coupling interfaces. The paper investigates the coupling algorithm performance and evaluates the pellet-to-cladding heat transfer effect for the rod ejection accident of a light water reactor core benchmark. Full article

(This article belongs to the Special Issue Computational Techniques of Nuclear Reactor Physics)

► Show Figures

Figure 1

21 pages, 3353 KiB

Open AccessArticle

RAST-K v2—Three-Dimensional Nodal Diffusion Code for Pressurized Water Reactor Core Analysis

by Jinsu Park, Jaerim Jang, Hanjoo Kim, Jiwon Choe, Dongmin Yun, Peng Zhang, Alexey Cherezov and Deokjung Lee

Energies 2020, 13(23), 6324; https://doi.org/10.3390/en13236324 - 30 Nov 2020

Cited by 20 | Viewed by 2604

Abstract

The RAST-K v2, a novel nodal diffusion code, was developed at the Ulsan National Institute of Science and Technology (UNIST) for designing the cores of pressurized water reactors (PWR) and performing analyses with high accuracy and computational performance by adopting state-of-the-art calculation models [...] Read more.

The RAST-K v2, a novel nodal diffusion code, was developed at the Ulsan National Institute of Science and Technology (UNIST) for designing the cores of pressurized water reactors (PWR) and performing analyses with high accuracy and computational performance by adopting state-of-the-art calculation models and various engineering features. It is a three-dimensional multi-group nodal diffusion code developed for the steady and transient states using microscopic cross-sections generated by the STREAM code for 37 isotopes. A depletion chain containing 22 actinides and 15 fission products and burnable absorbers was solved using the Chebyshev rational approximation method. A simplified one-dimensional single-channel thermal-hydraulic calculation was performed with various values for the thermal conductivity. Advanced features such as burnup adaptation and CRUD modeling capabilities are implemented for the multi-cycle analysis of commercial reactor power plants. The performance of RAST-K v2 has been validated with the measured data of PWRs operating in Korea. Furthermore, RAST-K v2 has been coupled with a sub-channel code (CTF), fuel performance code (FRAPCON), and water chemistry code for multiphysics analyses. In this paper, the calculation models and engineering features implemented in RAST-K v2 are described, and then the application status of RAST-K v2 is presented. Full article

(This article belongs to the Special Issue Computational Techniques of Nuclear Reactor Physics)

► Show Figures

Figure 1

Journal Menu

Journal Browser

Computational Techniques of Nuclear Reactor Physics

Share This Special Issue

Special Issue Editor

Special Issue Information

Keywords

Published Papers (8 papers)

Research

Further Information

Guidelines

MDPI Initiatives

Follow MDPI