• Nuclear Engineering and Technology
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• 제55권2호
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• pp.734-748
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• 2023

A comprehensive investigation of the Ronen method is performed in homogeneous and heterogeneous slab problems from the Sood benchmark, considering isotropic and linearly-anisotropic problems. Three finite differences implementations are exercised and compared. The results are compared to reference solutions using one and two energy groups. The validation is performed for the criticality eigenvalue and the fundamental neutron flux distribution. The results demonstrate the significantly improved accuracy achievable by the Ronen method using a broad set of problems. For standard convergence tolerances, the maximal deviation in criticality eigenvalue is less than ten pcm, and the maximal deviation in the spatial distribution of the flux is less than 2%, always located near sharp interfaces or vacuum boundaries.

• 한국재료학회:학술대회논문집
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• 한국재료학회 2011년도 추계학술발표대회
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• pp.7-7
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• 2011

Quantum beam technology has been expected to develop breakthroughs for nanotechnology during the third basic plan of science and technology (2006~2010). Recently, Green- or Life Innovations has taken over the national interests in the fourth basic science and technology plan (2011~2015). The NIMS (National Institute for Materials Science) has been conducting the corresponding mid-term research plans, as well as other national projects, such as nano-Green project (Global Research for Environment and Energy based on Nanomaterials science). In this lecture, the research trends in Japan and NIMS are firstly reviewed, and the typical achievements are highlighted over key nanotechnology fields. As one of the key nanotechnologies, the quantum beam research in NIMS focused on synchrotron radiation, neutron beams and ion/atom beams, having complementary attributes. The facilities used are SPring-8, nuclear reactor JRR-3, pulsed neutron source J-PARC and ion-laser-combined beams as well as excited atomic beams. Materials studied are typically fuel cell materials, superconducting/magnetic/multi-ferroic materials, quasicrystals, thermoelectric materials, precipitation-hardened steels, nanoparticle-dispersed materials. Here, we introduce a few topics of neutron scattering and ion beam nanofabrication. For neutron powder diffraction, the NIMS has developed multi-purpose pattern fitting software, post RIETAN2000. An ionic conductor, doped Pr2NiO4, which is a candidate for fuel-cell material, was analyzed by neutron powder diffraction with the software developed. The nuclear-density distribution derived revealed the two-dimensional network of the diffusion paths of oxygen ions at high temperatures. Using the high sensitivity of neutron beams for light elements, hydrogen states in a precipitation-strengthened steel were successfully evaluated. The small-angle neutron scattering (SANS) demonstrated the sensitive detection of hydrogen atoms trapped at the interfaces of nano-sized NbC. This result provides evidence for hydrogen embrittlement due to trapped hydrogen at precipitates. The ion beam technology can give novel functionality on a nano-scale and is targeting applications in plasmonics, ultra-fast optical communications, high-density recording and bio-patterning. The technologies developed are an ion-and-laser combined irradiation method for spatial control of nanoparticles, and a nano-masked ion irradiation method for patterning. Furthermore, we succeeded in implanting a wide-area nanopattern using nano-masks of anodic porous alumina. The patterning of ion implantation will be further applied for controlling protein adhesivity of biopolymers. It has thus been demonstrated that the quantum beam-based nanotechnology will lead the innovations both for nano-characterization and nano-fabrication.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.151-152
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• 2004

An acceleration technique combined with the discrete ordinates method which has been widely used in the solution of neutron transport phenomena is applied to the solution of radiative transfer equation. The self-adjoint form of the second order radiation intensity equation is used to enhance the stability of the solution, and a new linearization method is developed to avoid the nonlinearity of the material temperature equation. This new acceleration method is applied to the well known Marshak wave problem, and the numerical result is compared with that of a non-accelerated calculation

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.103-108
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• 1995

By applying the P$_1$ and P$_2$ equations to the operator form of a synthetic acceleration, we derive the p,-acceleration (diffusion synthetic acceleration: DSA) and P$_2$-acceleration (p$_2$SA) schemes in one dimensional slab geometry. We Fourier-analyze the derived acceleration schemes with the discrete-ordinates transport equation and showed that the DSA outperforms the P$_2$SA. These results confirm that one cannot simply assume that replacement of the DSA with a higher order approximation will lead to a better acceleration performance.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2635-2649
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• 2022

CEFR is a small core-size sodium-cooled fast reactor (SFR) using high enrichment fuel with stainless-steel reflectors, which brings a significant challenge to the deterministic methodologies due to the strong spectral effect. The neutronic simulation of the start-up experiments conducted at the CEFR have been performed with a deterministic code system RAST-F, which is based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo (MC) code and a multi-group nodal diffusion solver. The RAST-F results were compared against the measurement data. Moreover, the characteristic of neutron spectrum in the fuel rings, and adjacent reflectors was evaluated using different models for generation of accurate nuclear libraries. The numerical solution of RAST-F system was verified against the full core MC solution MCS at all control rods fully inserted and withdrawn states. A good agreement between RAST-F and MCS solutions was observed with less than 120 pcm discrepancies and 1.2% root-mean-square error in terms of k_eff and power distribution, respectively. Meanwhile, the RAST-F result agreed well with the experimental values within two-sigma of experimental uncertainty. The good agreement of these results indicating that RAST-F can be used to neutronic steady-state simulations for small core-size SFR, which was challenged to deterministic code system.

• Bulletin of the Korean Chemical Society
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• 제25권3호
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• pp.382-388
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• 2004

The sizes and structures of micelles formed in aqueous solutions of cationic octadecyl trimethyl ammonium chloride (OTAC) and anionic ammonium dodecyl sulfate (ADS) surfactants were investigated using smallangle neutron scattering (SANS), self-diffusion coefficients by pulsed-gradient spin-echo (PGSE) NMR, and dynamic light scattering (DLS) methods. SANS and DLS data indicate that their structures are spherical at concentrations as high as 300 mM. As the total surfactant concentration increases, the peaks of SANS spectra shift to higher scattering vector and become sharper, indicating that the intermicellar distance decreases and its distribution becomes narrower. This is due to more compact packing of surfactant molecules at high concentrations. The intermicellar distance of around 100 ${\AA}$ above 200 mM corresponds approximately to the diameter of one micelle. The sizes of spherical micelles are 61 ${\AA}$ and 41 ${\AA}$ for 9 mM OTAC and 10 mM ADS, respectively. Also the self-diffusion coefficients by PGSE-NMR yield the apparent sizes 96 ${\AA}$ and 31 ${\AA}$ for micelles of 1 mM OTAC and 10 mM ADS, respectively. For ADS solutions of high concentrations (100-300 mM), DLS data show that the micelle size remains constant at $25{\pm}2{\AA}$. This indicates that the transition in micellar shape does not take place up to 300 mM, which is consistent with the SANS results.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4491-4503
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• 2023

This study aimed to verify and validate the transient simulation capability of the hybrid code system RAST-F for fast reactor analysis. For this purpose, control rod (CR) drop experiments involving eight separate CRs and six CR groups in the China Experimental Fast Reactor (CEFR) start-up tests were utilized to simulate the CR drop transient. The RAST-F numerical solution, including the neutron population, time-dependent reactivity, and CR worth, was compared against the measurement values obtained from two out-of-core detectors. Moreover, the time-dependent reactivity and CR worth from RAST-F were verified against the results obtained by the Monte Carlo code Serpent using continuous energy nuclear data. A code-to-code comparison between Serpent and RAST-F showed good agreement in terms of time-dependent reactivity and CR worth. The discrepancy was less than 160 pcm for reactivity and less than 110 pcm for CR worth. RAST-F solution was almost identical to the measurement data in terms of neutron population and reactivity. All the calculated CR worth results agreed with experimental results within two standard deviations of experimental uncertainty for all CRs and CR groups. This work demonstrates that the RAST-F code system can be a potential tool for analyzing time-dependent phenomena in fast reactors.

• Nuclear Engineering and Technology
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• 제52권2호
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• pp.223-229
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• 2020

The present work proposes a solution to the static Boltzmann transport equation approximated by the simplified P₃ (SP₃) on angular, and the analytic coarse mesh finite difference (ACMFD) for spatial variables. Multi-group SP₃-ACMFD equations in 3D rectangular geometry are solved using the GMRES solution technique. As the core time dependent analysis necessitates the solution of an eigenvalue problem for an initial condition, this work is hence devoted to development and verification of the proposed static SP₃-ACMFD solver. A 3D multi-group static diffusion solver is also developed as a byproduct of this work to assess the improvement achieved using the SP₃ technique. Static results are then compared against transport benchmarks to assess the proximity of SP₃-ACMFD solutions to their full transport peers. Results prove that the approach can be considered as an acceptable interim approximation with outputs superior to the diffusion method, close to the transport results, and with the computational costs less than the full transport approach. The work would be further generalized to time dependent solutions in Part II.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3543-3562
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• 2021

The methods and performance of a 3D pin-by-pin neutronics code based on the 2D/1D decoupling method are presented. The code was newly developed as an effort to achieve enhanced accuracy and high calculation performance that are sufficient for the use in practical nuclear design analyses. From the 3D diffusion-based finite difference method (FDM) formulation, decoupled planar formulations are established by treating pre-determined axial leakage as a source term. The decoupled axial problems are formulated with the radial leakage source term. To accelerate the pin-by-pin calculation, the two-level coarse mesh finite difference (CMFD) formulation, which consists of the multigroup node-wise CMFD and the two-group assembly-wise CMFD is implemented. To enhance the accuracy, both the discontinuity factor method and the super-homogenization (SPH) factor method are examined for pin-wise cross-section homogenization. The parallelization is achieved with the OpenMP package. The accuracy and performance of the pin-by-pin calculations are assessed with the VERA and APR1400 benchmark problems. It is demonstrated that pin-by-pin 2D/1D alternating calculations within the two-level 3D CMFD framework yield accurate solutions in about 30 s for the typical commercial core problems, on a parallel platform employing 32 threads.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 2000년도 추계 학술발표회 논문집
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• pp.115-120
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• 2000

The analysis of flux distribution under stead-state in large power reactors with assymetry reactivity insertions requires the use of three-dimensional diffusion calculations. For the purpose, consistently formulated modern nodal methods based on higher order interface techniques have become popular tools for flux distributions in large commercial nuclear reactors. Among the earlier developments, the nodal Green's function method obtains its nodal interface equation from the transverse-integrated integral diffusion equation using a finite-medium Green's function. In this method, the outgoing current from a node surface is formulated as a response of the incoming currents and the spatially integrated neutron source within the same node. The well-known nodal expansion method is also based on an interface partial current formulation. Nodal methods high-level interface variables, i.e., interface net current and flux, may be more computationally efficient than the nodal Green's function method because they have one fewer unknown per interface. The Analytic Nodal Method(ANM), which can be classified as an interface net current technique and, was faster in solving some standard benchmark problems than the other two methods.(omitted)