A technique for estimating neutron fluxes in a large number of in-core detectors by the Monte Carlo code PRIZMA for a full-scale model of a water-moderated reactor core

2014 ◽  
Author(s):  
Y.Z. Kandiev ◽  
E.A. Kashaeva ◽  
G.N. Malyshkin ◽  
K.E. Khatuntsev
Keyword(s):  
Monte Carlo ◽  
Reactor Core ◽  
Full Scale ◽  
Scale Model ◽  
Monte Carlo Code

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

Core Seismic Experiment and Analysis of Full Scale Single Model for Fast Reactor

2017 ◽  
Author(s):  
Akihisa Iwasaki ◽  
Shinichiro Matsubara ◽  
Tomohiko Yamamoto ◽  
Seiji Kitamura ◽  
Hidenori Harada
Keyword(s):  
Fast Reactor ◽  
Structural Integrity ◽  
Impact Force ◽  
Seismic Analysis ◽  
Reactor Core ◽  
Horizontal Displacement ◽  
Full Scale ◽  
Three Dimensions ◽  
Scale Model ◽  
Core Element

To design fast reactor (FR) core components, seismic response must be evaluated in order to ensure structural integrity. Generally, the fast reactor core is made of several hundred core elements in hexagonal arrangement. When a big earthquake occurs, large horizontal displacement, vertical displacement (raising) and impact force of each core element may cause a trouble for control rod insertability, reactivity insertion and core element intensity. Therefore, a seismic analysis method of a fast reactor core considering three-dimensional nonlinear behavior, such as bouncing, impact, fluid-structure interaction, etc. was developed. This paper presents a validation of the core element vibration analysis code in three dimensions (REVIAN-3D) for a full scale model. In this validation, the vertical behavior (rising displacement) and horizontal behavior (Impact force, horizontal response) as a single core element of the analysis result agreed very well with the experiments.

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.

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.

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.

A Method of Optimized Utilization of Point-Wise Data Format in Monte Carlo Code

2010 ◽  
Author(s):  
Yuxuan Liu ◽  
Ganglin Yu ◽  
Kan Wang
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Cross Sections ◽  
Search Algorithm ◽  
Reactor Core ◽  
Hash Table ◽  
Monte Carlo Code ◽  
Data Format ◽  
Section Data ◽  
Grid Interval

Monte Carlo codes are powerful and accurate tools for reactor core calculation. Most Monte Carlo codes use the point-wise data format, in which the data are given as tables of energy-cross section pairs. When calculating the cross sections at an incident energy value, it should be determined which grid interval the energy falls in. This procedure is repeated so frequently in Monte Carlo codes that its contribution in the overall calculation time can become quite significant. In this paper, the time distribution of Monte Carlo method is analyzed to illustrate the time consuming of cross section calculation. By investigation on searching and calculating cross section data in Monte Carlo code, a new search algorithm called hash table is elaborately designed to substitute the traditional binary search method in locating the energy grid interval. The results indicate that in the criticality calculation, hash table can save 5%∼17% CPU time, depending on the number of nuclides in the material, as well as complexity of geometry for particles tracking.

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