• Title/Summary/Keyword: Multi-group cross-sections

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Multigroup cross-sections generated using Monte-Carlo method with flux-moment homogenization technique for fast reactor analysis

  • Yiwei Wu;Qufei Song;Kuaiyuan Feng;Jean-Francois Vidal;Hanyang Gu;Hui Guo
    • Nuclear Engineering and Technology
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    • v.55 no.7
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    • pp.2474-2482
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    • 2023
  • The development of fast reactors with complex designs and operation status requires more accurate and effective simulation. The Monte-Carlo method can generate multi-group cross-sections in arbitrary geometry without approximation on resonances treatment and leads to good results in combination with diffusion codes. However, in previous studies, the coupling of Monte-Carlo generated multi-group cross-sections (MC-MGXS) and transport solvers has shown relatively large biases in fast reactor problems. In this paper, the main contribution to the biases is proved to be the neglect of the angle-dependence of the total cross-sections. The flux-moment homogenization technique (MHT) is proposed to take into account this dependence. In this method, the angular dependence is attributed to the transfer cross-sections, keeping an independent form for the total sections. For the MET-1000 benchmark, the multi-group transport simulation results with MC-MGXS generated with MHT are improved by 700 pcm and an additional 120 pcm with higher order scattering. The factors that cause the residual bias are discussed. The core power distribution bias is also significantly reduced when MHT is used. It proves that the MCMGXS with MHT can be applicable with transport solvers in fast reactor analysis.

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

  • Yiwei Wu;Qufei Song;Ruixiang Wang;Yao Xiao;Hanyang Gu;Hui Guo
    • Nuclear Engineering and Technology
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    • v.55 no.6
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    • pp.2112-2124
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    • 2023
  • 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.

STRAUM-MATXST: A code system for multi-group neutron-gamma coupled transport calculation with unstructured tetrahedral meshes

  • MyeongHyeon Woo;Ser Gi Hong
    • Nuclear Engineering and Technology
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    • v.54 no.11
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    • pp.4280-4295
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    • 2022
  • In this paper, a new multi-group neutron-gamma transport calculation code system STRAUM-MATXST for complicated geometrical problems is introduced and its development status including numerical tests is presented. In this code system, the MATXST (MATXS-based Cross Section Processor for SN Transport) code generates multi-group neutron and gamma cross sections by processing MATXS format libraries generated using NJOY and the STRAUM (SN Transport for Radiation Analysis with Unstructured Meshes) code performs multi-group neutron-gamma coupled transport calculation using tetrahedral meshes. In particular, this work presents the recent implementation and its test results of the Krylov subspace methods (i.e., Bi-CGSTAB and GMRES(m)) with preconditioners using DSA (Diffusion Synthetic Acceleration) and TSA (Transport Synthetic Acceleration). In addition, the Krylov subspace methods for accelerating the energy-group coupling iteration through thermal up-scatterings are implemented with new multi-group block DSA and TSA preconditioners in STRAUM.

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

  • Du, Xianan;Choe, Jiwon;Choi, Sooyoung;Lee, Woonghee;Cherezov, Alexey;Lim, Jaeyong;Lee, Minjae;Lee, Deokjung
    • Nuclear Engineering and Technology
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    • v.51 no.8
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    • pp.1871-1885
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    • 2019
  • 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%.

Analyzing nuclear reactor simulation data and uncertainty with the group method of data handling

  • Radaideh, Majdi I.;Kozlowski, Tomasz
    • Nuclear Engineering and Technology
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    • v.52 no.2
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    • pp.287-295
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    • 2020
  • Group method of data handling (GMDH) is considered one of the earliest deep learning methods. Deep learning gained additional interest in today's applications due to its capability to handle complex and high dimensional problems. In this study, multi-layer GMDH networks are used to perform uncertainty quantification (UQ) and sensitivity analysis (SA) of nuclear reactor simulations. GMDH is utilized as a surrogate/metamodel to replace high fidelity computer models with cheap-to-evaluate surrogate models, which facilitate UQ and SA tasks (e.g. variance decomposition, uncertainty propagation, etc.). GMDH performance is validated through two UQ applications in reactor simulations: (1) low dimensional input space (two-phase flow in a reactor channel), and (2) high dimensional space (8-group homogenized cross-sections). In both applications, GMDH networks show very good performance with small mean absolute and squared errors as well as high accuracy in capturing the target variance. GMDH is utilized afterward to perform UQ tasks such as variance decomposition through Sobol indices, and GMDH-based uncertainty propagation with large number of samples. GMDH performance is also compared to other surrogates including Gaussian processes and polynomial chaos expansions. The comparison shows that GMDH has competitive performance with the other methods for the low dimensional problem, and reliable performance for the high dimensional problem.

OECD/NEA BENCHMARK FOR UNCERTAINTY ANALYSIS IN MODELING (UAM) FOR LWRS - SUMMARY AND DISCUSSION OF NEUTRONICS CASES (PHASE I)

  • Bratton, Ryan N.;Avramova, M.;Ivanov, K.
    • Nuclear Engineering and Technology
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    • v.46 no.3
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    • pp.313-342
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    • 2014
  • A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for Uncertainty Analysis in Modeling (UAM) is defined in order to facilitate the development and validation of available uncertainty analysis and sensitivity analysis methods for best-estimate Light water Reactor (LWR) design and safety calculations. The benchmark has been named the OECD/NEA UAM-LWR benchmark, and has been divided into three phases each of which focuses on a different portion of the uncertainty propagation in LWR multi-physics and multi-scale analysis. Several different reactor cases are modeled at various phases of a reactor calculation. This paper discusses Phase I, known as the "Neutronics Phase", which is devoted mostly to the propagation of nuclear data (cross-section) uncertainty throughout steady-state stand-alone neutronics core calculations. Three reactor systems (for which design, operation and measured data are available) are rigorously studied in this benchmark: Peach Bottom Unit 2 BWR, Three Mile Island Unit 1 PWR, and VVER-1000 Kozloduy-6/Kalinin-3. Additional measured data is analyzed such as the KRITZ LEU criticality experiments and the SNEAK-7A and 7B experiments of the Karlsruhe Fast Critical Facility. Analyzed results include the top five neutron-nuclide reactions, which contribute the most to the prediction uncertainty in keff, as well as the uncertainty in key parameters of neutronics analysis such as microscopic and macroscopic cross-sections, six-group decay constants, assembly discontinuity factors, and axial and radial core power distributions. Conclusions are drawn regarding where further studies should be done to reduce uncertainties in key nuclide reaction uncertainties (i.e.: $^{238}U$ radiative capture and inelastic scattering (n, n') as well as the average number of neutrons released per fission event of $^{239}Pu$).

COMPUTATIONAL EFFICIENCY OF A MODIFIED SCATTERING KERNEL FOR FULL-COUPLED PHOTON-ELECTRON TRANSPORT PARALLEL COMPUTING WITH UNSTRUCTURED TETRAHEDRAL MESHES

  • Kim, Jong Woon;Hong, Ser Gi;Lee, Young-Ouk
    • Nuclear Engineering and Technology
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    • v.46 no.2
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    • pp.263-272
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    • 2014
  • Scattering source calculations using conventional spherical harmonic expansion may require lots of computation time to treat full-coupled three-dimensional photon-electron transport in a highly anisotropic scattering medium where their scattering cross sections should be expanded with very high order (e.g., $P_7$ or higher) Legendre expansions. In this paper, we introduce a modified scattering kernel approach to avoid the unnecessarily repeated calculations involved with the scattering source calculation, and used it with parallel computing to effectively reduce the computation time. Its computational efficiency was tested for three-dimensional full-coupled photon-electron transport problems using our computer program which solves the multi-group discrete ordinates transport equation by using the discontinuous finite element method with unstructured tetrahedral meshes for complicated geometrical problems. The numerical tests show that we can improve speed up to 17~42 times for the elapsed time per iteration using the modified scattering kernel, not only in the single CPU calculation but also in the parallel computing with several CPUs.

Sensitivity and uncertainty quantification of neutronic integral data in the TRIGA Mark II research reactor

  • Makhloul, M.;Boukhal, H.;Chakir, E.;El Bardouni, T.;Lahdour, M.;Kaddour, M.;Ahmed, Abdulaziz;Arectout, A.;El Yaakoubi, H.
    • Nuclear Engineering and Technology
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    • v.54 no.2
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    • pp.523-531
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    • 2022
  • In order to study the sensitivity and the uncertainty of the Moroccan research reactor TRIGA Mark II, a model of this reactor has been developed in our ERSN laboratory for use with the N-Particle MCNP Monte Carlo transport codes (version 6). In this article, the sensitivities of the effective multiplication factor of this reactor are evaluated using the ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0 libraries and in 44 energy groups, for the cross sections of the fuel (U-235 and U-238) and the moderator (H-1 and O-16). However, the quantification of the uncertainty of the nuclear data is performed using the nuclear code NJOY99 for the generation and processing of covariance matrices. On the one hand, the highest uncertainty deviations, calculated using the ENDFB-VII.1 and JENDL4.0 evaluations, are 2275, 386 and 330 pcm respectively for the reactions U235(n, f), $ U_{235}(n\bar{\nu})$ and H1(n, γ). On the other hand, these differences are very small for the neutron reactions of O-16 and U-238. Regarding the neutron spectra, in CT-mid plane, they are very close for the three evaluations (ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0). These spectra present two peaks (thermal and fission) around the energies 0.05 eV and 1 MeV.

Space grid analysis method in modelling shear lag of cable-stayed bridge with corrugated steel webs

  • Ma, Ye;Ni, Ying-Sheng;Xu, Dong;Li, Jin-Kai
    • Steel and Composite Structures
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    • v.24 no.5
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    • pp.549-559
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    • 2017
  • As few multi-tower single-box multi-cell cable-stayed bridges with corrugated steel webs have been built, analysis is mostly achieved by combining single-girder model, beam grillage model and solid model in support of the design. However, such analysis methods usually suffer from major limitations in terms of the engineering applications: single-girder model fails to account for spatial effect such as shear lag effect of the box girder and the relevant effective girder width and eccentric load coefficient; owing to the approximation in the principle equivalence, the plane grillage model cannot accurately capture shear stress distribution and local stress state in both top and bottom flange of composite box girder; and solid model is difficult to be practically combined with the overall calculation. The usual effective width method fails to provide a uniform and accurate "effective length" (and the codes fail to provide a unified design approach at those circumstance) considering different shear lag effects resulting from dead load, prestress and cable tension in the construction. Therefore, a novel spatial grid model has been developed to account for shear lag effect. The theoretical principle of the proposed spatial grid model has been elaborated along with the relevant illustrations of modeling parameters of composite box girder with corrugated steel webs. Then typical transverse and longitudinal shear lag coefficient distribution pattern at the side-span and mid-span key cross sections have been analyzed and summarized to provide reference for similar bridges. The effectiveness and accuracy of spatial grid analysis methods has been finally validated through a practical cable-stayed bridge.

BOND STRENGTH OF BONDED AMALGAM USING DENIAL ADHESIVES (치과용 접착제를 사용한 접착 아말감의 결합 강도)

  • Kam, Dong-Hoon;Lee, Sang-Dae;Hahn, Se-Hyun
    • Journal of the korean academy of Pediatric Dentistry
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    • v.26 no.2
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    • pp.284-295
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    • 1999
  • The purpose of this study was to measure and analyze the bond strength of bonded amalgam using dental adhesives and to compare this with light-curing composite resin. Sections 8mm in diameter were punched out from the labial surface of bovine anterior teeth. These were embedded in clear acrylic resin blocks with labial surface facing out. 55 specimens were made for enamel and dentin each. After dividing these into 5 groups, group 1: Superbond C&B, group 2: Panavia 21, group 3: All-Bond 2, group 4: Fuji I Glass Ionomer Luting Cement, group 5: Scotchbond Multi-Purpose(Restorative Z-100), molds with holes of 6.3mm in diameter and 1.5mm in depth were placed over the specimens. The exposed tooth surfaces were treated with adhesives and the molds were filled with amalgam. In group 5, the mold was filled with composite resin and light-cured for 40 seconds. The author measured all specimens for bond strength 24 hours after amalgam filing and analyzed fracture surfaces. The following results were obtained: 1. Among the dentin groups, groups 1, 2 and 4 showed significantly lower bond strength compared with group 5(P<0.05). 2. Among the enamel groups, group 4 showed significantly lower bond strength compared with group 5(P<0.05). 3. In group 2, 2D showed significantly lower bond strength compared with group 2E(P<0.05). Other adhesives showed no such differences in bond strength between dentin and enamel(P>0.05). 4. Cohesive failure was observed in groups 1E and 5D, while mixed failure was seen in groups 1 and 5. Only adhesive failures were noted in groups 2, 3, 4.

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