Coupling of system thermal–hydraulics and Monte-Carlo code: Convergence criteria and quantification of correlation between statistical uncertainty and coupled error

2015 ◽  
Vol 75 ◽  
pp. 377-387 ◽  
Author(s):  
Xu Wu ◽  
Tomasz Kozlowski
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Code ◽  
Statistical Uncertainty ◽  
Thermal Hydraulics ◽  
Convergence Criteria

scholarly journals DEVELOPMENT OF 3D PIN-BY-PIN CORE SOLVER TORTIN AND COUPLING WITH THERMAL-HYDRAULICS

2021 ◽  
Vol 247 ◽  
pp. 02034
Author(s):  
P. Mala ◽  
A. Pautz ◽  
H. Ferroukhi ◽  
A. Vasiliev
Keyword(s):  
Monte Carlo ◽  
State Of The Art ◽  
The State ◽  
Monte Carlo Code ◽  
Channel Code ◽  
Thermal Hydraulics ◽  
Fuel Temperature ◽  
Power Profile ◽  
Assembly Problem ◽  
Good Agreement

Currently, safety analyses mostly rely on codes which solve both the neutronics and the thermal-hydraulics with assembly-wise nodes resolution as multiphysics heterogeneous transport solvers are still too time and memory expensive. The pin-by-pin homogenized codes can be seen as a bridge between the heterogeneous codes and the traditional nodal assembly-wise calculations. In this work, the pin-by-pin simplified transport solver Tortin has been coupled with a sub-channel code COBRA-TF. The verification of the 3D solver of Tortin is presented at first, showing very good agreement in terms of axial and radial power profile with the Monte Carlo code SERPENT for a small minicore and with the state-of-the-art nodal code SIMULATE5 for a quarter core without feedback. Then the results of Tortin+COBRA-TF are compared with SIMULATE5 for one assembly problem with feedback. The axial profiles of power and moderator temperature show good agreement, while the fuel temperature differ by up to 40 K. This is caused mainly by different gap and fuel conductance parameters used in COBRA-TF and in SIMULATE5.

scholarly journals MULTI-PHYSICS TRANSIENT SIMULATIONS WITH TRIPOLI-4®

2021 ◽  
Vol 247 ◽  
pp. 07019 ◽  
Author(s):  
Margaux Faucher ◽  
Davide Mancusi ◽  
Andrea Zoia
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Coolant Temperature ◽  
Test Case ◽  
Monte Carlo Code ◽  
Thermal Hydraulics ◽  
Fuel Temperature ◽  
Control Rod ◽  
Transient Simulations ◽  
Neutron Power

In this work, we present the first dynamic calculations performed with the Monte Carlo neutron transport code TRIPOLI-4R with thermal-hydraulics feedback. For this purpose, the Monte Carlo code was extended for multi-physics capabilities and coupled to the thermal-hydraulics subchannel code SUBCHANFLOW. As a test case for the verification of transient simulation capabilities, a 3x3-assembly mini-core benchmark based on the TMI-1 reactor is considered with a pin-by-pin description. Two reactivity excursion scenarios initiated by control-rod movement are simulated starting from a critical state and compared to analogous simulations performed using the Serpent 2 Monte-Carlo code. The time evolution of the neutron power, fuel temperature, coolant temperature and coolant density are analysed to assess the multi-physics capabilities of TRIPOLI-4. The stabilizing e_ects of thermal-hydraulics on the neutron power appear to be well taken into account. The computational requirements for massively parallel calculations are also discussed.

Newly Developed Coupling Scheme in RMC With Internal Thermal Feedback and OTF Doppler-Broadening Method

2017 ◽  
Author(s):  
Ouwen Yexin ◽  
Shanfang Huang ◽  
Kan Wang
Keyword(s):  
Monte Carlo ◽  
Steady State ◽  
Cross Section ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Simulation Analysis ◽  
Coupling Scheme ◽  
Monte Carlo Code ◽  
Doppler Broadening ◽  
Thermal Hydraulics

RMC (Reactor Monte Carlo)[1] is a self-developed Monte Carlo code for nuclear reactor analysis by Reactor Engineering Analysis Lab (REAL), Tsinghua University. On the basis of the self-developed subchannel module (RMC-TH) and Monte Carlo Cell Tally, the internal coupling interface is developed, which combines both input files to one and realizes the fast mesh correspondence process using the cell expansion technology for repeated structure with thermal-hydraulics feedback. It breaks through the bottleneck of geometrical extensibility for coupled code. On-the-fly Doppler broadening method is adopted as the way to consider the temperature effect on microscopic cross section, which only needs the 0 K cross section library so that the memory cost can be apparently reduced. Steady state simulation analysis are performed on PWR fuel pin and 17×17 assembly model, and the results show the feasibility, accuracy and efficiency of the coupling methodology. Therefore, a promising technology roadmap for the large-scale and geometrically universal nuclear reactor in both steady-state and transient conditions with thermal-hydraulic feedback are established. The roadmap can be further applied to neutronics-thermal-hydraulics-depletion coupling in multi-physics simulation process.

Validation of GATE Monte Carlo code for simulation of proton therapy using National Institute of Standards and Technology library data

2018 ◽  
Vol 18 (1) ◽  
pp. 38-45 ◽  
Author(s):  
Shiva Zarifi ◽  
Hadi Taleshi Ahangari ◽  
Sayyed Bijan Jia ◽  
Mohammad Ali Tajik-Mansoury
Keyword(s):  
Monte Carlo ◽  
Energy Range ◽  
Water Medium ◽  
Monte Carlo Code ◽  
Statistical Uncertainty ◽  
Simulation Code ◽  
Simulation Efficiency ◽  
Optimal Value ◽  
Beam Range ◽  
Range Calculation

AbstractAimTo validate the Geant4 Application for Tomographic Emission (GATE) Monte Carlo simulation code by calculating the proton beam range in the therapeutic energy range.Materials and methodsIn this study, the GATE code which is based on Geant4 was used for simulation. The proton beams in the therapeutic energy range (5–250 MeV) were simulated in a water medium, and then compared with the data from National Institute of Standards and Technology (NIST) in order to investigate the accuracy of different physics list available in the GATE code. In addition, the optimal value of SetCut was assessed.ResultsIn all energy ranges, the QBBC physics had a greater deviation in the ranges relative to the NIST data. With respect to the range calculation accuracy, the QGSP_BIC_EMY and QGSP_BERT_HP_EMY physics were in the range of statistical uncertainty; however, QGSP_BIC_EMY produced better results using the least squares. Based on an investigation into the range calculation precision and simulation efficiency, the optimal SetCut was set at 0·1 mm.FindingsBased on an investigation into the range calculation precision and simulation yield, the QGSP_BIC_EMY physics and the optimal SetCut was recommended to be 0·1 mm.

Dose calculations in heterogeneous volumes with the GATE Monte Carlo software for radiological protection

2019 ◽  
Vol 54 (2) ◽  
pp. 125-132
Author(s):  
T. Deschler ◽  
N. Arbor ◽  
F. Carbillet ◽  
A. Nourreddine
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Methods ◽  
Medical Physics ◽  
Computation Time ◽  
Radiological Protection ◽  
Monte Carlo Code ◽  
Statistical Uncertainty ◽  
Dose Calculations ◽  
The Subject ◽  
Reduce Computation Time

Monte Carlo methods have become widespread in the field of radiation protection and in particular in medical physics where the use of voxelized volumes for the reconstruction of dosimetric quantities is increasing. Changing the resolution of a dose map can be useful to compare dosimetric results coming from voxelized volumes with different resolutions, or to reduce computation time. This can be done by superimposing a dosel grid with a different resolution than that of the voxelized volume. In this case, each dosel will cover several voxels, leading the Monte Carlo code to calculate the dose in heterogeneous volumes. Two algorithms are available in GATE to perform these calculations, the Volume-Weighting (V-W) and the Mass-Weighting (M-W) algorithms, the latter being the subject of this work. In a general way, the M-W algorithm tends to reconstruct a higher dose than that the V-W one. In dosels involving heavy and lightweight materials (air-skin, bone-tissue), the M-W reconstructed dose is better estimated than the V-W one (up to 10% better at the air-skin interface). Moreover, the statistical uncertainty of the M-W dose can be up to 80% lower than the V-W one at air-skin interfaces. These results show that the M-W algorithm is more suitable for radiological protection applications and must be preferentially used in GATE for dose calculations in heterogeneous volumes.

scholarly journals EFFECT OF ENERGY DEPOSITION MODELLING IN COUPLED STEADY STATE MONTE CARLO NEUTRONICS/THERMAL HYDRAULICS CALCULATIONS

2021 ◽  
Vol 247 ◽  
pp. 06001
Author(s):  
Riku Tuominen ◽  
Ville Valtavirta ◽  
Manuel García ◽  
Diego Ferraro ◽  
Jaakko Leppänen
Keyword(s):  
Monte Carlo ◽  
Steady State ◽  
Power Distribution ◽  
Energy Deposition ◽  
Computational Time ◽  
Test Case ◽  
Monte Carlo Code ◽  
Thermal Hydraulics ◽  
New Treatment ◽  
Effective Multiplication

In coupled calculations with Monte Carlo neutronics and thermal hydraulics the Monte Carlo code is used to produce a power distribution which in practice means tallying the energy deposition. Usually the energy deposition is estimated by making a simple approximation that energy is deposited only in fission reactions. The goal of this work is to study how the accuracy of energy deposition modelling affects the results of steady state coupled calculations. For this task an internal coupling between Monte Carlo transport code Serpent 2 and subchannel code SUBCHANFLOW is used along with a recently implemented energy deposition treatment of Serpent 2. The new treatment offers four energy deposition modes each of which offers a different combination of accuracy and required computational time. As a test case, a 3D PWR fuel assembly is modelled with different energy deposition modes. The resulting effective multiplication factors are within 30 pcm. Differences of up to 100K are observed in the fuel temperatures.

scholarly journals COUPLED MODELLING OF THE EBR-II SHRT-45R INCLUDING PHOTON HEAT DEPOSITION

2021 ◽  
Vol 247 ◽  
pp. 07014
Author(s):  
Domenico Valerio ◽  
Nicolò Abrate ◽  
Sandra Dulla ◽  
Giuseppe Francesco Nallo ◽  
Ravetto Piero
Keyword(s):  
Monte Carlo ◽  
Transient Analysis ◽  
Nuclear Data ◽  
Monte Carlo Code ◽  
Thermal Hydraulics ◽  
Fast Breeder Reactor ◽  
Photon Transport ◽  
Coupled Modelling ◽  
Heat Deposition ◽  
Full Core

The Fast REactor NEutronics/Thermal-hydraulICs (FRENETIC) code has been developed during the last years at Politecnico di Torino, implementing a full-core coupled neutronic/thermal-hydraulics model for steady-state and transient analysis of liquid-metal cooled fast breeder reactor (LMFBR). In the framework of the validation activities for the code, an analysis of the sodium-cooled reactor EBR-II, previously carried out in the frame of a IAEA Coordinated Research Project, is performed with FRENETIC including the most recent physics models. In particular, photon transport and heat deposition are taken into account, a feature which has been proved in previous studies to be relevant to the correct study of the EBR-II core. To this purpose, a set of nuclear data for photons has been generated by means of the Monte Carlo code Serpent-2, and it is demonstrated that the code is able to take into account the photon heat deposition in the EBR-II.

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