scholarly journals MAPPING OF SODIUM VOID EFFECT AND DOPPLER CONSTANT IN ESFR-SMART CORE WITH MONTE CARLO CODE SERPENT AND DETERMINISTIC CODE ERANOS

2021 ◽  
Vol 247 ◽  
pp. 13004
Author(s):  
Jiri Krepel ◽  
Valeria Raffuzzi
Keyword(s):  
Monte Carlo ◽  
Inner Core ◽  
Fuel Cycle ◽  
Cycle Performance ◽  
Monte Carlo Code ◽  
Stochastic Noise ◽  
Sodium Fast Reactor ◽  
The Monte Carlo Method ◽  
Void Effect ◽  
Gen Iv

The Sodium Fast Reactor is one of the most technologically developed Gen-IV reactors, which can close the nuclear fuel cycle. Its criticality safety directly depends on the sodium void effect and Doppler constant. Hence the knowledge of their local distribution is important. These coefficients can be mapped by deterministic or Monte Carlo codes, where the latter provide higher modeling accuracy, but are also strongly computer demanding and subject to stochastic noise issues. In this study, the void effect and Doppler constant have been enumerated for the ESFR core by Serpent2 and ERANOS2 codes, preserving a six-batch operation scheme. The Serpent code was coupled to the Python script BBP to simulate batch-wise operation in a radially infinite inner core configuration; the ERANOS code was applied to the whole core geometry and the batch-wise operation was simulated by the EQL3D routine. Sodium void effect and Doppler constant spatial maps with different levels of refinement were produced, as well as the time evolution of the integral coefficients during the transition from initial cycle to equilibrium cycle. Both codes indicate deterioration of these coefficients during the transition. The equilibrium cycle performance of the inner core zone from the ERANOS calculation was compared with Serpent results and they showed reasonable agreement. For very fine mapping, the Monte Carlo method employed was computationally very demanding and the enumerated effect was lower than the stochastic noise. In general, the Serpent model practically excludes modeling assumptions and produces reliable results for reasonably sized maps, which can be combined if needed with the high spatial resolution results obtained by ERANOS simulations.

X-Ray Microanalysis Combined with Monte Carlo Simulation for the Analysis of Layered Thin Films: The Case of Carbon Contamination

2009 ◽  
Vol 15 (2) ◽  
pp. 99-105 ◽  
Author(s):  
Aldo Armigliato ◽  
Rodolfo Rosa
Keyword(s):  
Monte Carlo ◽  
Carbon Film ◽  
Sample Surface ◽  
Scanning Transmission Electron Microscope ◽  
Carbon Layer ◽  
Monte Carlo Code ◽  
X Ray ◽  
Transmission Electron ◽  
The Monte Carlo Method ◽  
Scanning Transmission

AbstractA previously developed Monte Carlo code has been extended to the X-ray microanalysis in a (scanning) transmission electron microscope of plan sections, consisting of bilayers and triple layers. To test the validity of this method for quantification purposes, a commercially available NiOx (x ∼ 1) thin film, deposited on a carbon layer, has been chosen. The composition and thickness of the NiO film and the thickness of the C support layer are obtained by fitting to the three X-ray intensity ratios I(NiK)/I(OK), I(NiK)/I(CK), and I(OK)/I(CK). Moreover, it has been investigated to what extent the resulting film composition is affected by the presence of a contaminating carbon film at the sample surface. To this end, the sample has been analyzed both in the (recommended) “grid downward” geometry and in the upside/down (“grid upward”) situation. It is found that a carbon contaminating film of few tens of nanometers must be assumed in both cases, in addition to the C support film. Consequently, assuming the proper C/NiOx/C stack in the simulations, the Monte Carlo method yields the correct oxygen concentration and thickness of the NiOx film.

scholarly journals Integration of the differential evolution algorithm and the Monte Carlo code to create a spectrometric detector model

2022 ◽  
Vol 2155 (1) ◽  
pp. 012020
Author(s):  
I V Prozorova
Keyword(s):  
Monte Carlo ◽  
Standard Procedure ◽  
Differential Evolution Algorithm ◽  
Germanium Detector ◽  
Cost Effective ◽  
Monte Carlo Code ◽  
High Purity Germanium ◽  
The Monte Carlo Method ◽  
Evolution Algorithm ◽  
Detector Model

Abstract A standard procedure for characterizing the high-purity germanium detector (HPGe), manufactured by Canberra Industries Inc [1], is performed directly by the company using patented methods. However, the procedure is usually expensive and must be repeated because the characteristics of the HPGe crystal change over time. In this work, the principles of a technique are developed for use in obtaining and optimizing the detector characteristics based on a cost-effective procedure in a standard research laboratory. The technique requires that the detector geometric parameters are determined with maximum accuracy by the Monte Carlo method [2] in parallel with the optimization based on evolutionary algorithms. The development of this approach facilitates modeling of the HPGe detector as a standardized procedure. The results will be also beneficial in the development of gamma spectrometers and/or their calibrations before routine measurements.

Using the Few-Group Approximation for Calculating Some Neutron-Physical Characteristics of VVER-1000 Core by Means of the MCU Monte-Carlo Code

2021 ◽  
Author(s):  
Artem S. Bikeev ◽  
Yulia S. Daichenkova ◽  
Mikhail A. Kalugin ◽  
Denis Shkarovsky ◽  
Vladislav V. Shkityr
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Method ◽  
Cross Sections ◽  
Optimal Number ◽  
Physical Characteristics ◽  
Monte Carlo Code ◽  
Special Version ◽  
The Monte Carlo Method ◽  
Scattering Anisotropy ◽  
Neutron Cross Sections

Abstract The main purpose of this work is to study the possibility of using the few-group approximation for calculation of some neutron-physical characteristics of VVER-1000 core by means of special version of MCU code. The Monte-Carlo method for VVER-1000 core neutron-physical characteristics calculation using the few-group approximation with an estimate of neutron cross sections “by location“ was provided and tested in this research. The reduction of calculation time due to the transition from a pointwise model of representation of cross sections to the few-group approximation and methodical error of this approach were evaluated. Optimal number of energy groups was determined. It was found that consideration of the scattering anisotropy leads to a significant decrease in methodical error. Ways of further reduction of methodical error were worked out.

scholarly journals MACROSCOPIC CROSS SECTIONS GENERATION BY MONTE CARLO CODE MCS FOR FAST REACTOR ANALYSIS

2021 ◽  
Vol 247 ◽  
pp. 02007
Author(s):  
Tung Dong Cao Nguyen ◽  
Hyunsuk Lee ◽  
Xianan Du ◽  
Vutheam Dos ◽  
Tuan Quoc Tran ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Fast Reactor ◽  
Three Dimensional ◽  
Monte Carlo Code ◽  
Metallic Fuel ◽  
Sodium Fast Reactor ◽  
The Difference ◽  
Nodal Diffusion ◽  
Core Problem

Recent researches have become more interested in the feasibility of using Monte Carlo (MC) code to generate multi-group (MG) cross sections (XSs) for fast reactor analysis using nodal diffusion codes. The current study, therefore, presents a brief methodology for MG XSs generation by the in-house UNIST MC code MCS, which can be compatibly utilized in nodal diffusion codes, PARCS and RAST-K. The applicability of the methodology is quantified on the sodium fast reactor (SFR) ABR-1000 design with a metallic fuel from the OECD/NEA SRF benchmark. The few-group XSs generated by MCS with a two-dimensional (2D) fuel assembly geometry are well consistent with those of SERPENT 2. Furthermore, the simulation of beginning-of-cycle (BOC) steady-state three-dimensional (3D) whole-core problem with PARCS and RAST-K is conducted using the generated 24-group XSs by MCS. The nodal diffusion solutions, including the core keff, power profiles and various of reactivity parameters, are compared to reference whole-core results obtained by MC code MCS. Overall, the code-to-code comparison indicates a reasonable agreement between deterministic and stochastic codes, with the difference in keff less than 100 pcm and the root-mean-square (RMS) error in assembly power less than 1.15%. Therefore, it is successfully demonstrated that the employment of the MG XSs generation by MCS for nodal diffusion codes is feasible to accurately perform analyses for fast reactors.

Extended development of a Monte Carlo code OpenMC for fuel cycle simulation of molten salt reactor

2020 ◽  
Vol 118 ◽  
pp. 103115
Author(s):  
Kun Zhuang ◽  
Ting Li ◽  
Qian Zhang ◽  
Qinghua He ◽  
Tengfei Zhang
Keyword(s):  
Monte Carlo ◽  
Molten Salt ◽  
Fuel Cycle ◽  
Monte Carlo Code ◽  
Molten Salt Reactor ◽  
Cycle Simulation

Radiology and mammography standard X-ray spectra simulated with the Monte Carlo method

2021 ◽  
Vol 9 (2C) ◽  
Author(s):  
Joel Marques Xavier Filho ◽  
Iury Santos Silveira ◽  
Linda Viola Ehlin Caldas
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Method ◽  
Quality Parameters ◽  
Monte Carlo Code ◽  
X Ray ◽  
The Monte Carlo Method ◽  
The Mean

Six standard beams described in the TRS-457 (IAEA): RQR 5, 8, M1, M2, M3, M4 were simulated using the EGSnrc Monte Carlo code. Each spectrum was created by an X-ray tube simulated in BEAMnrc, and attenuation curves were obtained using the application egs_kerma. The quality of each beam was evaluated by the 1st and 2nd half-value layers, the homogeneity coefficients and the mean energies. All beams presented quality parameters compatible with those described in TRS-457 (IAEA).

Application of Monte Carlo to PWR Cavity Radiation Streaming Calculation

2017 ◽  
Author(s):  
Han Jingru ◽  
Liu Qiaofeng ◽  
Chen Haiying ◽  
Zhang Chunming
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Method ◽  
Variance Reduction ◽  
Complex Geometry ◽  
Deep Penetration ◽  
Monte Carlo Code ◽  
Discrete Ordinate Method ◽  
Discrete Ordinate ◽  
The Monte Carlo Method ◽  
Cavity Radiation

The cavity streaming is the neutron beam from the reactor core through the tunnel, which is between the external surface of the pressure vessel and the shield inner surface. Reactor cavity streaming calculation is a typical deep penetration problem with complex geometry. The accurate calculation of neutron radiation streaming is a key problem to the reactor shielding calculation, for which the Monte Carlo method and the discrete ordinate method are two popular methods. The speed of discrete ordinate method calculation is fast, but it is hard to describe the complex pile of cavity; the Monte Carlo method can accurately describe the complex geometry, it has a high calculation precision, but with a low direct simulation efficiency. Based on a pressurized water reactor nuclear power plant, this paper presents a detailed model realized by Monte Carlo code, with continuous energy points cross section libraries. The neutron flux density distribution of PWR reactor cavity streaming can directly be calculated by a three-dimensional simulation. For such an actual deep penetration problem, a variety of variance reduction techniques are studied, an effective variance reduction technique is used to obtain results with small statistic errors for a Monte Carlo simulation, which effectively solves the problem of large-scale deep penetrating convergence difficulty, the cavity radiation streaming calculation and analysis are completed. The result shows that the Monte Carlo method can be used as a powerful tool to solve the problem of cavity streaming leakage.

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