Research on Coupling Scheme of Monte Carlo Burnup Calculation in RMC

2018 ◽  
Author(s):  
Wanlin Li ◽  
Kan Wang ◽  
Ganglin Yu ◽  
Yaodong Li
Keyword(s):  
Monte Carlo ◽  
Neutron Flux ◽  
Cross Section ◽  
Neutron Transport ◽  
Effective Cross Section ◽  
High Order ◽  
Step Method ◽  
Step Approximation ◽  
Predictor Corrector ◽  
Burnup Calculation

Monte Carlo (MC) burnup calculation method, implemented through coupling neutron transport and point depletion solvers, is widely used in design and analysis of nuclear reactor. Burnup calculation is generally solved by dividing reactor lifetime into steps and modeling geometry into numbers of burnup areas where neutron flux and one group effective cross sections are treated as constant during each burnup step. Such constant approximation for neutron flux and effective cross section will lead to obvious error unless using fairly short step. To yield accuracy and efficiency improvement, coupling schemes have been researched in series of MC codes. In this study, four coupling schemes, beginning of step approximation, predictor-corrector methods by correcting nuclide density and flux-cross section as well as high order predictor-corrector with sub-step method were researched and implemented in RMC. Verification and comparison were performed by adopting assembly problem from VERA international benchmark. Results illustrate that high order coupled with sub-step method is with notable accuracy compared to beginning of step approximation and traditional predictor-corrector, especially for calculation in which step length is fairly long.

scholarly journals A Coordinate-Free Approach to Tracing the Coupler Curves of Pin-Jointed Linkages

2011 ◽  
Author(s):  
Nicola´s Rojas ◽  
Federico Thomas
Keyword(s):  
High Order ◽  
Step Method ◽  
Surjective Mapping ◽  
Predictor Corrector

In general, high-order coupler curves of plane mechanisms cannot be properly traced by standard predictor-corrector algorithms due to drifting problems and the presence of singularities. Instead of focusing on finding better algorithms for tracing curves, a simple coordinate-free method that first traces these curves in a distance space and then maps them onto the mechanism workspace is proposed. Tracing a coupler curve in the proposed distance space is much simpler because (a) the equation of this curve in this space can be straightforwardly obtained from a sequence of bilaterations; and (b) the curve in this space naturally decomposes into branches in which the signs of the oriented areas of the triangles involved in the aforementioned bilaterations remain constant. A surjective mapping permits to map the thus traced curves onto the workspace of the mechanism. The advantages of this two-step method are exemplified by tracing the coupler curves of a double butterfly linkage, curves that can reach order 48.

A Universal Adjoint-Weighted Algorithm for Geometric Sensitivity Analysis of K-Eigenvalue Based on Continuous-Energy Monte Carlo Method

2018 ◽  
Author(s):  
Hao Li ◽  
Ganglin Yu ◽  
Shanfang Huang ◽  
Kan Wang
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Geometric Parameter ◽  
Neutron Transport ◽  
Delta Function ◽  
Step Function ◽  
Universal Method ◽  
Analytic Geometry ◽  
Heaviside Step Function ◽  
Dirac Delta

There exists a typical problem in Monte Carlo neutron transport: the effective multiplication factor sensitivity to geometric parameter. In several methods attempting to solve it, Monte Carlo adjoint-weighted theory has been proven to be quite effective. The major obstacle of adjoint-weighted theory is calculating derivative of cross section with respect to geometric parameter. In order to fix this problem, Heaviside step function and Dirac delta function are introduced to describe cross section and its derivative. This technique is crucial, and it establishes the foundation of further research. Based on above work, adjoint-weighted method is developed to solve geometric sensitivity. However, this method is limited to surfaces which are uniformly expanded or contracted with respect to its origin, such as vertical movement of plane or expansion of sphere. Rotation and translation are not allowed, while these two transformation types are more common and more important in engineering projects. In this paper, a more universal method, Cell Constraint Condition Perturbation (CCCP) method, is developed and validated. Different from traditional method, CCCP method for the first time explicitly articulates that the perturbed quantity is the parameter of spatial analytic geometry equations that used to describe surface. Thus, the CCCP can treat arbitrary one-parameter geometric perturbation of arbitrary surface as long as this surface can be described by spatial analytic geometry equation. Furthermore, CCCP can treat the perturbation of the whole cell, such as translation, rotation, expansion and constriction. Several examples are calculated to confirm the validity of CCCP method.

THE ABSORPTION OF HIGH-ENERGY RADIATION, A MONTE CARLO STUDY: I. RECTANGULAR GEOMETRY

1963 ◽  
Vol 41 (4) ◽  
pp. 632-650 ◽  
Author(s):  
N. R. Steenberg ◽  
W. Van Iterson
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Effective Cross Section ◽  
Monte Carlo Study ◽  
High Energy ◽  
Rare Gas ◽  
High Energy Radiation ◽  
Energy Radiation ◽  
Alternative Formula ◽  
Random Array

As part of an investigation of the scattering of high-energy radiation by nuclei a Monte Carlo study has been made of the attenuation of radiation by a rectangular barrier composed of a moderately dense random array of rigid semiopaque spheres. It is found that the attenuation is much more rapid than that predicted by the rare-gas formula usually assumed and is well described by an alternative formula which is derived on probabilistic grounds. A corollary is that an effective cross section which is substantially larger than the free cross section must be assumed inside such a medium.

Cross Section Measurements for the 103Rh(n,n′)103Rhm Reaction from 0.122 to 14.74 MeV

1974 ◽  
Vol 52 (15) ◽  
pp. 1421-1428 ◽  
Author(s):  
D. C. Santry ◽  
J. P. Butler
Keyword(s):  
Neutron Flux ◽  
Cross Section ◽  
Neutron Spectrum ◽  
Cross Sections ◽  
Energy Levels ◽  
Effective Cross Section ◽  
Fission Neutron ◽  
Activation Method ◽  
Excitation Curve ◽  
Fission Neutron Spectrum

Cross sections for the production of 103Rhm were measured by the activation method. At energies below 5.3 MeV the neutron flux was measured with a calibrated neutron long counter, while at higher energies, measurements were made relative to the known cross section for the 32S(n,p)32P reaction. The shape of the Rh excitation curve is discussed in terms of known energy levels in 103Rh. An effective cross section for a 235U fission neutron spectrum calculated from the measured excitation curve is 724 ± 43 mb.

Simulation on Neutron Space-Time Kinetics for Accelerator Driven Sub-Critical System

2017 ◽  
Author(s):  
Yingming Song ◽  
Qingyu Gao ◽  
Ke Wang ◽  
Yaping Guo ◽  
Lu Zhang ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Neutron Flux ◽  
Transport Theory ◽  
Neutron Transport ◽  
Space Time ◽  
Outlet Temperature ◽  
Critical System ◽  
Static Approximation ◽  
Energy Group ◽  
Mc Method

Monte Carlo transport theory was applied to the variables space and time separated framework of neutron space-time kinetics calculation for Accelerator driven sub-critical reactor. The improved quasi-static approximation was combined with Monte Carlo neutron transport code (IQS/MC) for neutron space-time kinetic process of ADS sub-critical system. Besides, the IQS/MC simulation calculation program with visualization operation platform for ADS sub-critical system was developed simultaneously. The beam transient was analysed simulatedly based on the physical model of CIADS. Three-dimensional grid distributions of relative neutron flux of energy group were separated along time can be obtained by computing energy group separated of neutron flux, meanwhile the totally relative power, fuel temperature and outlet temperature of coolant at the core varied as the time were also obtained. The calculated results of IQS/MC method and point reactor method were compared, which agreed well with the relevant physics laws and verify that the IQS/MC method is applicable to the simulation of ADS neutron space-time kinetics and ADS neutronics transient security analysis.

scholarly journals Development of an MCNP5-ORIGEN2 coupling scheme for burnup calculation of VVER-1000 fuel assemblies

2016 ◽  
Vol 6 (3) ◽  
pp. 16-30
Author(s):  
Huy Hiep Nguyen ◽  
Huu Tiep Nguyen ◽  
Viet Phu Tran ◽  
Tuan Khai Nguyen
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Shielding Effect ◽  
Coupling Scheme ◽  
Fuel Rods ◽  
Transport Code ◽  
Fuel Assemblies ◽  
Calculation Code ◽  
Burnup Calculation ◽  
Good Agreement

The paper aims to develop an MCNP5-ORIGEN2 coupling scheme for burnup calculation. Specifically, the Monte Carlo neutron transport code (MCNP5) and the nuclides depletion and decay calculation code (ORIGEN2) are combined by data processing and linking files written in the PERL programming language. The validity and applicability of the developed coupling scheme are tested through predicting the neutronic and isotopic behavior of the “VVER-1000 LEU Assembly Computational Benchmark”. The MCNP5-ORIGEN2 coupling results showed a good agreement with the k-inf benchmark values within 600 pcm during the entire burnup history. In addition, the differences of isotopes concentration at the end of the burnup (40 MWd/kgHM) when compared with benchmark values were reasonable and generally within 6.5%. The developed coupling scheme also considered the shielding effect due to gadolinium isotopes and simulated well the depletion of isotopes as a function of the radial position in gadolinium bearing fuel rods.

scholarly journals MPACT VERIFICATION WITH MAGNOX REACTOR NEUTRONICS PROGRESSION PROBLEMS

2021 ◽  
Vol 247 ◽  
pp. 10031
Author(s):  
Nicholas P. Luciano ◽  
Brian J. Ade ◽  
Kang Seog Kim ◽  
Andrew J. Conant
Keyword(s):  
Cross Section ◽  
Neutron Transport ◽  
Three Dimensional ◽  
Light Water Reactors ◽  
Pressurized Water Reactor ◽  
Reactor Physics ◽  
Pressurized Water ◽  
Water Reactors ◽  
Improved Performance ◽  
Burnup Calculation

MPACT is a state-of-the-art core simulator designed to perform high-fidelity analysis using whole-core, three-dimensional, pin-resolved neutron transport calculations on modern parallel computing hardware. MPACT was originally developed to model light water reactors, and its capabilities are being extended to simulate gas-cooled, graphite-moderated cores such as Magnox reactors. To verify MPACT’s performance in this new application, the code is being formally benchmarked using representative problems. Progression problems are a series of example models that increase in complexity designed to test a code’s performance. The progression problems include both beginning-of-cycle and depletion calculations. Reference solutions for each progression problem have been generated using Serpent 2, a continuous-energy Monte Carlo reactor physics burnup calculation code. Using the neutron multiplication eigenvalue ke_ as a metric, MPACT’s performance is assessed on each of the progression problems. Initial results showed that MPACT’s multigroup cross section libraries, originally developed for pressurized water reactor problems, were not sufficient to accurately solve Magnox problems. MPACT’s improved performance on the progression problems is demonstrated using this new optimized cross section library.

Adjoint Monte Carlo neutron transport using cross-section probability table representation

2010 ◽  
Vol 37 (9) ◽  
pp. 1186-1196 ◽  
Author(s):  
Cheikh M’Backé Diop ◽  
Odile Petit ◽  
Cédric Jouanne ◽  
Mireille Coste-Delclaux
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Neutron Transport ◽  
Probability Table

scholarly journals Uncertainty propagation based on correlated sampling technique for nuclear data applications

2020 ◽  
Vol 6 ◽  
pp. 8 ◽  
Author(s):  
Axel Laureau ◽  
Vincent Lamirand ◽  
Dimitri Rochman ◽  
Andreas Pautz
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Neutron Transport ◽  
Uncertainty Propagation ◽  
Computation Time ◽  
Sampling Technique ◽  
Nuclear Data ◽  
Correlated Sampling ◽  
Enriched Uranium ◽  
The Impact

A correlated sampling technique has been implemented to estimate the impact of cross section modifications on the neutron transport and in Monte Carlo simulations in one single calculation. This implementation has been coupled to a Total Monte Carlo approach which consists in propagating nuclear data uncertainties with random cross section files. The TMC-CS (Total Monte Carlo with Correlated Sampling) approach offers an interesting speed-up of the associated computation time. This methodology is detailed in this paper, together with two application cases to validate and illustrate the gain provided by this technique: the highly enriched uranium/iron metal core reflected by a stainless-steel reflector HMI-001 benchmark, and the PETALE experimental programme in the CROCUS zero-power light water reactor.

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