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 Burnup calculation of the OECD VVER-1000 LEU benchmark assembly using MCNP6 and SRAC2006

2021 ◽  
Vol 8 (4) ◽  
pp. 10-19
Author(s):  
Tiep Nguyen Huu ◽  
Dung Nguyen Thi ◽  
Phu Tran Viet ◽  
Thanh Tran Vinh ◽  
Ha Pham Nhu Viet
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Wide Temperature Range ◽  
Collision Probability ◽  
Shielding Effect ◽  
Probability Method ◽  
Monte Carlo Code ◽  
Benchmark Data ◽  
Fuel Assemblies ◽  
Burnup Calculation

The present work aims to perform burnup calculation of the OECD VVER-1000 LEU (lowenriched uranium) computational benchmark assembly using the Monte Carlo code MCNP6 and the deterministic code SRAC2006. The new depletion capability of MCNP6 was applied in the burnup calculation of the VVER-1000 LEU benchmark assembly. The OTF (on-the-fly) methodology of MCNP6, which involves high precision fitting of Doppler broadened cross sections over a wide temperature range, was utilized to handle temperature variation for heavy isotopes. The collision probability method based PIJ module of SRAC2006 was also used in this burnup calculation. The reactivity of the fuel assembly, the isotopic concentrations and the shielding effect due to the presenceof the gadolinium isotopes were determined with burnup using MCNP6 and SRAC2006 incomparison with the available published benchmark data. This study is therefore expected to reveal the capabilities of MCNP6 and SRAC2006 in burnup calculation of VVER-1000 fuel assemblies.

scholarly journals Predicting Ex-core Detector Response in a PWR with Monte Carlo Neutron Transport Methods

2020 ◽  
Vol 225 ◽  
pp. 03007
Author(s):  
Tanja Goričanec ◽  
Domen Kotnik ◽  
Žiga Štancar ◽  
Luka Snoj ◽  
Marjan Kromar
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Reactor Core ◽  
Response Prediction ◽  
Detector Response ◽  
Design Calculation ◽  
Monte Carlo Code ◽  
Control Rod ◽  
The Core ◽  
Fuel Assemblies

An approach for calculating ex-core detector response using Monte Carlo code MCNP was developed. As a first step towards ex-core detector response prediction a detailed MCNP model of the reactor core was made. A script called McCord was developed as a link between deterministic program package CORD-2 and Monte Carlo code MCNP. It automatically generates an MCNP input from the CORD-2 data. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krško nuclea power plant. For the hot zero power configuration, the deviations are within 3 % for majority of fuel assemblies and slightly higher for fuel assemblies located at the core periphery. The computational model was further verified by comparing the calculated control rod worth to the CORD-2 results. The deviations were within 50 pcm and considered acceptable. The research will in future be supplemented with the in-core and ex-core detector signal calculations and neutron transport outside the reactor core.

Performance, Accuracy and Efficiency Evaluation of a Three-Dimensional Whole-Core Neutron Transport Code AGENT

2006 ◽  
Author(s):  
Tatjana Jevremovic ◽  
Mathieu Hursin ◽  
Nader Satvat ◽  
John Hopkins ◽  
Shanjie Xiao ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Three Dimensional ◽  
Open Architecture ◽  
Coarse Mesh ◽  
Research Reactors ◽  
Acceleration Techniques ◽  
Performance Accuracy ◽  
Transport Code ◽  
Neutron Propagation

The AGENT (Arbitrary GEometry Neutron Transport) an open-architecture reactor modeling tool is deterministic neutron transport code for two or three-dimensional heterogeneous neutronic design and analysis of the whole reactor cores regardless of geometry types and material configurations. The AGENT neutron transport methodology is applicable to all generations of nuclear power and research reactors. It combines three theories: (1) the theory of R-functions used to generate real three-dimensional whole-cores of square, hexagonal or triangular cross sections, (2) the planar method of characteristics used to solve isotropic neutron transport in non-homogenized 2D) reactor slices, and (3) the one-dimensional diffusion theory used to couple the planar and axial neutron tracks through the transverse leakage and angular mesh-wise flux values. The R-function-geometrical module allows a sequential building of the layers of geometry and automatic submeshing based on the network of domain functions. The simplicity of geometry description and selection of parameters for accurate treatment of neutron propagation is achieved through the Boolean algebraic hierarchically organized simple primitives into complex domains (both being represented with corresponding domain functions). The accuracy is comparable to Monte Carlo codes and is obtained by following neutron propagation through real geometrical domains that does not require homogenization or simplifications. The efficiency is maintained through a set of acceleration techniques introduced at all important calculation levels. The flux solution incorporates power iteration with two different acceleration techniques: Coarse Mesh Rebalancing (CMR) and Coarse Mesh Finite Difference (CMFD). The stand-alone originally developed graphical user interface of the AGENT code design environment allows the user to view and verify input data by displaying the geometry and material distribution. The user can also view the output data such as three-dimensional maps of the energy-dependent mesh-wise scalar flux, reaction rate and power peaking factor. The AGENT code is in a process of an extensive and rigorous testing for various reactor types through the evaluation of its performance (ability to model any reactor geometry type), accuracy (in comparison with Monte Carlo results and other deterministic solutions or experimental data) and efficiency (computational speed that is directly determined by the mathematical and numerical solution to the iterative approach of the flux convergence). This paper outlines main aspects of the theories unified into the AGENT code formalism and demonstrates the code performance, accuracy and efficiency using few representative examples. The AGENT code is a main part of the so called virtual reactor system developed for numerical simulations of research reactors. Few illustrative examples of the web interface are briefly outlined.

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.

Representative neutronic characteristics calculations for the VVER-1000 reactors using SRAC and MCNP5

2016 ◽  
Vol 6 (2) ◽  
pp. 21-30
Author(s):  
Huu Tiep Nguyen ◽  
Viet Phu Tran ◽  
Tuan Khai Nguyen ◽  
Vinh Thanh Tran ◽  
Minh Tuan Nguyen
Keyword(s):  
Monte Carlo ◽  
Power Distribution ◽  
Reactor Core ◽  
Multiplication Factor ◽  
Benchmark Problem ◽  
Fuel Assemblies ◽  
Effective Multiplication Factor ◽  
Neutronic Characteristics ◽  
Effective Multiplication ◽  
Good Agreement

This paper presents the results of neutronic calculations using the deterministic and Monte-Carlo methods (the SRAC and MCNP5codes) for the VVER MOX Core Computational Benchmark Specification and the VVER-1000/V392 reactor core. The power distribution and keff value have been calculated for a benchmark problem of VVER core. The results show a good agreement between the SRAC and MCNP5 calculations. Then, neutronic characteristics of VVER-1000/V392 such as power distribution, infinite multiplication factor (k-inf) of the fuel assemblies, effective multiplication factor keff, peaking factor and Doppler coefficient were calculated using the two codes.

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