Monte-Carlo code calculation of 3D reactor core model with usage of burnt fuel isotopic compositions, obtained by engineering codes

Kerntechnik ◽  
2016 ◽  
Vol 81 (4) ◽  
pp. 356-362 ◽  
Author(s):  
S. S. Aleshin ◽  
S. S. Gorodkov ◽  
A. I. Shcherenko
Keyword(s):  
Monte Carlo ◽  
Reactor Core ◽  
Core Model ◽  
Monte Carlo Code ◽  
Isotopic Compositions

Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations

Kerntechnik ◽  
2015 ◽  
Vol 80 (4) ◽  
pp. 394-401 ◽  
Author(s):  
S. S. Aleshin ◽  
S. S. Gorodkov ◽  
A. I. Shcherenko
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Code ◽  
Isotopic Compositions

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.

scholarly journals Generation of XS library for the reflector of VVER reactor core using Monte Carlo code Serpent

2017 ◽  
Vol 781 ◽  
pp. 012029 ◽  
Author(s):  
K I Usheva ◽  
S A Kuten ◽  
A A Khruschinsky ◽  
L F Babichev
Keyword(s):  
Monte Carlo ◽  
Reactor Core ◽  
Vver Reactor ◽  
Monte Carlo Code

scholarly journals EFFECT OF THE UNIFORM FISSION SOURCE METHOD ON LOCAL POWER VARIANCE IN FULL CORE SERPENT CALCULATION

2021 ◽  
Vol 247 ◽  
pp. 04024
Author(s):  
Yurii Bilodid ◽  
Jaakko Leppänen
Keyword(s):  
Monte Carlo ◽  
Power Distribution ◽  
Reactor Core ◽  
Monte Carlo Code ◽  
Local Power ◽  
Axial Distribution ◽  
Source Method ◽  
Power Variance ◽  
Full Core ◽  
Statistical Variance

One of challenges of the Monte Carlo full core simulations is to obtain acceptable statistical variance of local parameters throughout the whole reactor core at a reasonable computation cost. The statistical variance tends to be larger in low-power regions. To tackle this problem, the Uniform-Fission-Site method was implemented in Monte Carlo code MC21 and its effectiveness was demonstrated on NEA Monte Carlo performance benchmark. The very similar method is also implemented in Monte Carlo code Serpent under the name Uniform Fission Source (UFS) method. In this work the effect of UFS method implemented in Serpent is studied on the BEAVRS benchmark which is based on a real PWR core with relatively flat radial power distribution and also on 3x3 PWR mini-core simulated with thermo-hydraulic and thermo-mechanic feedbacks. It is shown that the application of the Uniform Fission Source method has no significant effect on radial power variance but equalizes axial distribution of variance of local power.

Application of a 3D Discrete Ordinates-Monte Carlo Coupling Method on CAP1400 Cavity Streaming Calculation

2017 ◽  
Author(s):  
Zheng Zheng ◽  
Hui Li ◽  
Mengqi Wang
Keyword(s):  
Monte Carlo ◽  
Reactor Core ◽  
Coupling Method ◽  
Normal Operation ◽  
Discrete Ordinates ◽  
Deep Penetration ◽  
Monte Carlo Code ◽  
Surface Source ◽  
Dose Rates ◽  
Mc Method

Neutrons and photons produced from reactor core during operation pass through the pressure vessel, reach the reactor cavity, and form the reactor cavity streaming. Reactor cavity streaming dose rates calculation during normal operation is important for the evaluation and control of the equipment dose rates in the nuclear power plant. Because reactor is great in dimension and complex in geometry, neutrons and photons fluence rates declined by several orders from reactor core to outside. Cavity streaming calculation is a deep penetration calculation with heavy computation load which is difficult to converge. Three dimensional Discrete Ordinates and Monte Carlo (SN-MC) coupling method combines the advantage of the SN method with high efficiency and the MC method with fine geometrical modeling. The SN-MC coupling method decreases the tally errors and increases the efficiency of the MC method effectively by using MC surface source generated by the SN fluence rates. In this paper, the theoretical model of the 3D SN-MC coupling method is presented. In order to fulfill the coupling calculation, a 3D Discrete Ordinates code is modified to output angular fluence rates, a link code DO2MC is developed to calculate cummulative distribution functions of source particle variables on surface source, and a source subroutine is written for a 3D Monte Carlo code. The 3D SN-MC coupling method is applied on the calculation of the CAP1400 cavity streaming neutron and photon dose rates. Numerical results show that the 3D SN-MC coupling codes are correct, the relative errors of the results are less than 20% compared with those of the MC bootstrapping method, and the efficiency is greatly enhanced.

Criticality qualification of a new Monte Carlo code for reactor core analysis

2009 ◽  
Vol 36 (11-12) ◽  
pp. 1689-1693 ◽  
Author(s):  
N. Catsaros ◽  
B. Gaveau ◽  
M. Jaekel ◽  
J. Maillard ◽  
G. Maurel ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Reactor Core ◽  
Monte Carlo Code ◽  
Core Analysis

PWR MOX/UO2 Transient Benchmark Calculation Using Monte Carlo Serpent 2 Code and Open Nodal Core Simulator ADPRES

2020 ◽  
Author(s):  
Muhammad Imron ◽  
Donny Hartanto
Keyword(s):  
Monte Carlo ◽  
Diffusion Coefficients ◽  
Power Distribution ◽  
Reactor Core ◽  
Reactor Power ◽  
Monte Carlo Code ◽  
Step Method ◽  
Control Rod ◽  
Benchmark Calculation ◽  
Transient Solutions

Abstract This paper presents static and transient solutions for the PWR MOX/UO2 transient benchmark by Serpent 2 Monte Carlo code and open nodal core simulator called ADPRES. The presences of MOX fuels and burn-up variation in the benchmark’s reactor core pose challenges for reactor simulators due to severe flux gradient across fuel assemblies. In this work, the two-step method was used, in which the assembly level two-group constants were generated from single assembly calculations with zero net current boundary conditions using Serpent 2 Monte Carlo code, and later the core calculation was performed using ADPRES open nodal core simulator. Two types of diffusion coefficients were generated: the conventional B1 leakage corrected and Cumulative Migration Method (CMM). Finally, the solutions of Serpent 2/ADPRESS, including multiplication factor, power distribution, control rod worth, and critical boron concentration using both diffusion coefficients were compared against solutions from heterogeneous Serpent 2 calculations where the fuel and cladding are explicitly modeled. The reactor power during transients were also compared qualitatively against other nodal core simulators. The results showed that Serpent 2/ADPRES were able to predict the heterogeneous Monte Carlo solutions very well with reasonable differences. The transient solutions were also quite accurate compared to other nodal core simulators. As for the diffusion coefficients comparison, it was found that the CMM diffusion coefficient provide more accurate solutions for the benchmark compared to the B1 leakage corrected diffusion coefficients.

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