scholarly journals Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems

2015 ◽  
Vol 2015 ◽  
pp. 1-11 ◽  
Author(s):  
Wonkyeong Kim ◽  
Jinsu Park ◽  
Tomasz Kozlowski ◽  
Hyun Chul Lee ◽  
Deokjung Lee
Keyword(s):  
Comparative Analysis ◽  
Cross Sections ◽  
Power Distribution ◽  
Reactor Core ◽  
Computer Code ◽  
Anisotropic Scattering ◽  
Monte Carlo Code ◽  
Core Analysis ◽  
Modeling Approach ◽  
Significant Difference

A high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a two-step homogenization modeling approach, using contemporary code systems developed for reactor core analysis. The focus of this work is a comprehensive comparative analysis of the conventional approaches and codes with a small core design, DIMPLE S06 critical experiment. The calculation procedure for the two approaches is explicitly presented in this paper. Comprehensive comparative analysis is performed by neutronics parameters: multiplication factor and assembly power distribution. Comparison of two-group homogenized cross sections from each lattice physics codes shows that the generated transport cross section has significant difference according to the transport approximation to treat anisotropic scattering effect. The necessity of the ADF to correct the discontinuity at the assembly interfaces is clearly presented by the flux distributions and the result of two-step approach. Finally, the two approaches show consistent results for all codes, while the comparison with the reference generated by MCNP shows significant error except for another Monte Carlo code, SERPENT2.

A Monte Carlo study on burnup treatment in sodium-cooled reactor with Th fuel

Kerntechnik ◽  
2021 ◽  
Vol 86 (4) ◽  
pp. 302-311
Author(s):  
M. E. Korkmaz ◽  
N. K. Arslan
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Power Distribution ◽  
Reactor Core ◽  
Monte Carlo Study ◽  
Minor Actinides ◽  
Monte Carlo Code ◽  
Effective Multiplication Factor ◽  
Conceptual Core ◽  
Effective Multiplication

Abstract Sodium Cooled Reactors is one of the Generation-IV plants selected to manage the long-lived minor actinides and to transmute the long-life radioactive elements. This study presents the comparison between two-designed SFR cores with 600 and 800 MWth total heating power. We have analyzed a conceptual core design and nuclear characteristic of SFR. Monte Carlo depletion calculations have been performed to investigate essential characteristics of the SFR core. The core calculations were performed by using the Serpent Monte Carlo code for determining the burnup behavior of the SFR, the power distribution and the effective multiplication factor. The neutronic and burn-up calculations were done by means of Serpent-2 Code with the ENDF-7 cross-sections library. Sodium Cooled Fast Reactor core was taken as the reference core for Th-232 burnup calculations. The results showed that SFR is an important option to deplete the minor actinides as well as for transmutation from Th-232 to U-233.

Characterization of Radioactive Waste From Side Structural Components of a CANDU Reactor for Decommissioning Applications in Korea

2010 ◽  
Author(s):  
Rizwan Ahmed ◽  
Gyunyoung Heo ◽  
Dong-Keun Cho ◽  
Jongwon Choi
Keyword(s):  
Radioactive Waste ◽  
Cross Sections ◽  
Reactor Core ◽  
International Standards ◽  
Monte Carlo Code ◽  
Capture Cross ◽  
Structural Components ◽  
The Core ◽  
Waste Classification ◽  
Capture Cross Sections

Reactor core components and structural materials of nuclear power plants to be decommissioned have been irradiated by neutrons of various intensities and spectrum. This long term irradiation results in the production of large number of radioactive isotopes that serve as a source of radioactivity for thousands of years for future. Decommissioning of a nuclear reactor is a costly program comprising of dismantling, demolishing of structures and waste classification for disposal applications. The estimate of radio-nuclides and radiation levels forms the essential part of the whole decommissioning program. It can help establishing guidelines for the waste classification, dismantling and demolishing activities. ORIGEN2 code has long been in use for computing radionuclide concentrations in reactor cores and near core materials for various burn-up-decay cycles, using one-group collapsed cross sections. Since ORIGEN2 assumes a constant flux and nuclide capture cross-sections in all regions of the core, uncertainty in its results could increase as region of interest goes away from the core. This uncertainty can be removed by using a Monte Carlo Code, like MCNP, for the correct calculations of flux and capture cross-sections inside the reactor core and in far core regions. MCNP has greater capability to model the reactor problems in much realistic way that is to incorporate geometrical, compositional and spectrum information. In this paper the classification of radioactive waste from the side structural components of a CANDU reactor is presented. MCNP model of full core was established because of asymmetric structure of the reactor. Side structural components of total length 240 cm and radius 16.122 cm were modeled as twelve (12) homogenized cells of 20 cm length each along the axial direction. The neutron flux and one-group collapsed cross-sections were calculated by MCNP simulation for each cell, and then those results were applied to ORIGEN2 simulation to estimate nuclide inventory in the wastes. After retrieving the radiation level of side structural components of in- and ex-core, the radioactive wastes were classified according to the international standards of waste classification. The wastes from first and second cell of the side structural components were found to exhibit characteristics of class C and Class B wastes respectively. However, the rest of the waste was found to have activity levels as that of Class A radio-active waste. The waste is therefore suitable for land disposal in accordance with the international standards of waste classification and disposal.

scholarly journals Comparative Analysis of Severe Accident at WWER-1000 NPP with MELCOR 1.8.5 and 2.1 Code Versions

2020 ◽  
pp. 30-40
Author(s):  
O. Kotsuba ◽  
Yu. Vorobyov ◽  
O. Zhabin ◽  
D. Gumenyuk
Keyword(s):  
Comparative Analysis ◽  
Pressure Vessel ◽  
Reactor Core ◽  
Computer Code ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Bottom Part ◽  
Control Volumes ◽  
Vessel Bottom ◽  
Reactor Pressure

An overview of the main improvements in updated version 2.1 of MELCOR computer code related to more representative mathematical modeling of complex thermohydraulic severe accident processes of core degradation, transfer of molten fragments to the bottom of the reactor, heating and failure of the bottom of the reactor pressure vessel is presented. The elements of WWER-1000 NPP computer model for the MELCOR 1.8.5 (control volumes, thermal structures and structures of the reactor core) that are reproduced for a reactor with the primary side, the secondary side and the containment are described. The changes implemented in WWER-1000 NPP model for MELCOR 1.8.5 to convert it to MELCOR 2.1 version that are mainly related to more detailed modeling of the reactor core and reactor pressure vessel bottom are provided. The paper presents the results of comparative analysis of severe accident scenario of total station blackout at WWER-1000 NPP with MELCOR 1.8.5 and 2.1. The comparison demonstrates good agreement between the main parameters’ results (pressure and temperature in hydraulic elements of the primary, secondary sides and the containment, temperature of core elements, the mass of the generated non-condensed gases and their concentration in the containment) obtained with these code versions for severe accident in-vessel phase. The identified differences in the time of core structures degradation and reactor vessel bottom failure are insignificantly affected by the behavior of the parameters in the primary side and the containment in the in-vessel phase of the severe accident and are related to more detailed modelling of the reactor core and bottom part of the reactor in MELCOR 2.1.

scholarly journals Study on the effect of various burnable poisons on pebble bed reactor with OTTO fuelling schemes using Serpent 2 Monte Carlo code

2021 ◽  
Vol 927 (1) ◽  
pp. 012018
Author(s):  
Nicholas Sidharta ◽  
Almanzo Arjuna
Keyword(s):  
Power Density ◽  
Power Distribution ◽  
Reactor Core ◽  
Monte Carlo Code ◽  
Volumetric Fraction ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Burnable Poison ◽  
Discrete Element Method Simulation ◽  
Upper Level

Abstract Pebble bed reactor with a once-through-then-out fuelling scheme has the advantage of simplifying the refueling system. However, the core upper-level power density is relatively higher than the bottom, producing an asymmetric core axial power distribution. Several burnable poison (BP) configurations are used to flatten the peak power density and improve power distribution while suppressing the excess core reactivity at the beginning of the burnup cycle. This study uses HTR-PM, China’s pebble bed reactor core, to simulate several burnable poison (BP) configurations. Serpent 2 coupled with Octave and a discrete element method simulation is used to model and simulate the pebble bed reactor core. It is found that erbium needs a large volumetric fraction in either QUADRISO or distributed BP to perform well. On the other hand, gadolinium and boron need a smaller volumetric fraction but perform worse in radial power distribution criteria in the fuel sphere. This study aims to verify the effect of BP added fuel pebbles on an OTTO refueling scheme HTR-PM core axial power distribution and excess reactivity.

scholarly journals Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL

2015 ◽  
Vol 2015 ◽  
pp. 1-9 ◽  
Author(s):  
A. Rais ◽  
D. Siefman ◽  
G. Girardin ◽  
M. Hursin ◽  
A. Pautz
Keyword(s):  
Power Distribution ◽  
Reactor Core ◽  
Transient Behavior ◽  
Nuclear Data ◽  
Safety Analysis ◽  
Monte Carlo Code ◽  
Thermal Hydraulics ◽  
Control Rod ◽  
Nuclear Data Library ◽  
And Control

In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.

scholarly journals Temperature field in gas-cooled reactor core in transient conditions under different approaches to mass flow profiling

2019 ◽  
Vol 5 (4) ◽  
pp. 297-303
Author(s):  
Vyacheslav S. Kuzevanov ◽  
Sergey K. Podgorny
Keyword(s):  
Temperature Field ◽  
Mass Flow ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Computer Code ◽  
Cfd Modeling ◽  
Problem Solution ◽  
Conductivity Equation ◽  
The Core ◽  
Significant Difference

Positive effect of profiling the gas-cooled reactor core within the framework of the GT-MHR project was investigated in (Podgorny and Kuzevanov 2017, Kuzevanov and Podgorny 2017, 2018). The necessity arises to supplement already implemented analysis of equilibrium conditions of core operation with investigation of effects of profiling on the temperature field in transient modes of reactor core operation. The present paper is dedicated to the investigation of development of transients in gas-cooled nuclear reactor core subject to the implementation of different principles of core profiling. Investigation of transients in reactor core represents complex problem, solution of which by conducting direct measurements is beyond the resources available to the authors. Besides the above, numerical simulation based on advanced CFD software complexes (ANSYS 2016, 2016a, 2016b, Shaw 1992, Anderson et al. 2009, Petrila and Trif 2005, Mohammadi and Pironneau 1994) is also fairly demanding in terms of required computer resources. The algorithm for calculating temperature fields using the model where the reactor core is represented as the solid medium with gas voids was developed by the authors and the assumption was made that heat transfer due to molecular heat conductivity can be described by thermal conductivity equation written for continuous medium with thermal physics parameters equivalent to respective parameters of porous object in order to get the possibility of obtaining prompt solutions of this type of problems. Computer code for calculating temperature field in gas-cooled reactor in transient operation modes was developed based on the suggested algorithm. Proprietary computation code was verified by comparing the results of numerous calculations with results of CFD-modeling of respective transients in the object imitating the core of gas-cooled nuclear reactor. The advantage of the developed computer code is the possibility of real-time calculation of evolution of conditions in complex configurations of gas-cooled reactor cores with different channel diameters. This allows using the computer code in the calculations of transients in loops of reactor facility as a whole, in particular for developing reactor simulators. Results are provided of calculations of transients for reactor core imitating the core of gas-cooled nuclear reactor within the framework of GT-MHR project performed for different approaches to profiling coolant mass flow. Results of calculations unambiguously indicate the significant difference of temperature regimes during transients in the reactor core with and without profiling and undeniable enhancement of reliability of nuclear reactor (Design of the Reactor Core 2005, International Safeguards 2014, Safety of Nuclear Power Plants 2014) with profiling of coolant mass flow in the reactor core as a whole.

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.

scholarly journals SMALL MODULAR PWR DESIGN FOR TRU RECYCLING WITH McCARD-MASTER TWO-STEP PROCEDURE

2021 ◽  
Vol 247 ◽  
pp. 01003
Author(s):  
Dae Hee Hwang ◽  
Ser Gi Hong
Keyword(s):  
Mass Flow ◽  
Cross Sections ◽  
Treatment Method ◽  
Monte Carlo Code ◽  
Core Analysis ◽  
Code System ◽  
Lattice Calculation ◽  
Group Constant ◽  
Step Procedure ◽  
Neutronic Characteristics

In our previous study, a small modular PWR core was designed for TRU (Transuranics) recycling with multi-recycling scheme with a typical two-step procedure using DeCART2D/MASTER code system in which the lattice analysis for producing homogenized group constant was performed by DeCART2D while whole core analysis was conducted by MASTER code. However, the neutron spectrum hardening of the LWR core loaded with TRU requires validating the multi-group cross section library and resonance self-shielding treatment method in lattice calculation. In this study, a new procedure using McCARD/MASTER was used to analyze the SMR core, in which the lattice calculation was performed by a Monte Carlo code called McCARD with a continuous energy library to generate homogenized two-group assembly cross sections. The SMR core analysis was performed to show neutronic characteristics and TRU mass flow in the SMR core with TRU multi-recycling. The result shows that the analyses on the neutronic characteristics and TRU mass flow using the McCARD/MASTER code system showed good agreement with the previous ones using the DeCART2D/MASTER code system. The neutronic characteristics of each cycle of the core satisfied the typical limit of a commercial PWR core and the SMR core consumes effectively TRU with net TRU consumption rates of 8.46~14.33 %.

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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