scholarly journals GENERATION OF NODAL CORE SIMULATOR UTILIZING VERA

2021 ◽  
Vol 247 ◽  
pp. 02018
Author(s):  
Paul Turinsky ◽  
Aaron Graham ◽  
Hisham Sarsour ◽  
Benjamin Collins
Keyword(s):  
Cross Sections ◽  
Reactor Core ◽  
Form Factors ◽  
Thermal Conditions ◽  
Diffusion Method ◽  
Fuel Temperature ◽  
Nuclear Core ◽  
Zero Current ◽  
Nodal Diffusion ◽  
Full Core

Nuclear core simulators based upon few-group nodal diffusion method currently are widely used to predict light water reactor core behavior. Nodal parameters’ input, e.g. cross-sections, discontinuity factors, and pin form factors, are typically generated utilizing lattice physics codes. In doing so, a number of approximations are introduced related to using zero current boundary conditions, 2-D radial geometry, and uniform thermal conditions in coolant and fuel. Usage of full core models with prediction fidelity typical of lattice physics to predict nodal parameters would eliminate these approximations. The VERA code can serve as such a full core model and was so utilized in this work. Via subsequent processing of VERA predictions, for a range of state points, nodal parameters and their functionalization in terms of coolant density, fuel temperature, and soluble poison concentration, were obtained and input to the NESTLE nodal code. By processing VERA predictions using consistent nodal methodologies as used in NESTLE, when using nodal parameters after functionalization based upon All-Rods-Out (ARO) VERA state points, the maximum reactivity and pin power differences between VERA and NESTLE were 2 pcm and 0.003 for ARO core simulations. For rodded core simulations, these maximum differences grew to 58 pcm and 0.04. Increases in differences were determined to be attributed to usage of unrodded nodal parameters generated using the VERA ARO state points whether the core is partially rodded or not, consistent with lattice physics practice. Obtaining unrodded nodal parameters using the VERA rodded state points reduced maximum differences to 2 pcm and 0.003 in pin powers.

scholarly journals Evaluation of PBMR-400 Core Design Steady State Condition with Serpent and AGREE

2021 ◽  
Vol 2048 (1) ◽  
pp. 012024
Author(s):  
H Ardiansyah ◽  
V Seker ◽  
T Downar ◽  
S Skutnik ◽  
W Wieselquist
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Reactor Core ◽  
Light Water Reactor ◽  
Steady State Condition ◽  
Light Water ◽  
Water Reactor ◽  
The Core ◽  
Core Simulation ◽  
Full Core

Abstract The significant recent advances in computer speed and memory have made possible an increasing fidelity and accuracy in reactor core simulation with minimal increase in the computational burden. This has been important for modeling some of the smaller advanced reactor designs for which simplified approximations such as few groups homogenized diffusion theory are not as accurate as they were for large light water reactor cores. For narrow cylindrical cores with large surface to volume ratios such the Ped Bed Modular Reactor (PBMR), neutron leakage from the core can be significant, particularly with the harder neutron spectrum and longer mean free path than a light water reactor. In this paper the core from the OECD PBMR-400 benchmark was analyzed using multigroup Monte Carlo cross sections in the HTR reactor core simulation code AGREE. Homogenized cross sections were generated for each of the discrete regions of the AGREE model using a full core SERPENT Monte Carlo model. The cross sections were generated for a variety of group structures in AGREE to assess the importance of finer group discretization on the accuracy of the core eigenvalue and flux predictions compared to the SERPENT full core Monte Carlo solution. A significant increase in the accuracy was observed by increasing the number of energy groups, with as much as a 530 pcm improvement in the eigenvalue calculation when increasing the number of energy groups from 2 to 14. Significant improvements were also observed in the AGREE neutron flux distributions compared to the SERPENT full core calculation.

scholarly journals INTRODUCTION OF OSCAR-4 AT THE HIGH FLUX REACTOR (PETTEN)

2021 ◽  
Vol 247 ◽  
pp. 10029
Author(s):  
B. Erasmus ◽  
J.A. Hendriks ◽  
A. Hogenbirk ◽  
S.C. van der Marck ◽  
N.L. Asquith
Keyword(s):  
Cross Sections ◽  
Neutron Transport ◽  
Reactor Core ◽  
Reaction Rates ◽  
High Flux ◽  
High Flux Reactor ◽  
Support Tool ◽  
Flux Reactor ◽  
Nodal Diffusion ◽  
Reactor Models

Since 2005 the nodal diffusion based code system, OSCAR-3, was used for reactor support calculations of operational cycles of the High Flux Reactor in Petten, The Netherlands. OSCAR uses a two-step deterministic calculation, in which homogenized cross sections are generated in lattice environments using neutron transport simulations, and then passed to a nodal diffusion core simulator to model the full reactor. Limitations in OSCAR-3 led to the need for improved modelling capabilities and better physics models for components present in the reactor core. OSCAR-4 offers improvements over OSCAR-3 in its approach to homogenization, and the new version of the diffusion core simulator allows for better modelling of movable components such as control rods. Fuel inventories calculated using OSCAR-4 can also easily be exported to MCNP, which allows the calculation of individual plate powers and local reaction rates amongst others. For these reasons OSCAR-4 is currently being introduced as a core support tool at the High Flux Reactor. In this work the steps that were followed to validate the reactor models are presented, and include results of validation calculations from both OSCAR-4 and MCNP6 over multiple reactor cycles. In addition differences in cross section library evaluations and their impact on the results are presented for the MCNP model.

New analysis of the low-energy π±p differential cross-sections of the CHAOS Collaboration

2015 ◽  
Vol 24 (07) ◽  
pp. 1550050 ◽  
Author(s):  
E. Matsinos ◽  
G. Rasche
Keyword(s):  
Experimental Data ◽  
Cross Sections ◽  
Differential Cross ◽  
Form Factors ◽  
Partial Wave Analysis ◽  
Meson Mass ◽  
Data Sets ◽  
Differential Cross Sections ◽  
Pion Nucleon ◽  
Meson Factory

In a previous paper, we reported the results of a partial-wave analysis (PWA) of the pion–nucleon (πN) differential cross-sections (DCSs) of the CHAOS Collaboration and came to the conclusion that the angular distribution of their π+p data sets is incompatible with the rest of the modern (meson factory) database. The present work, re-addressing this issue, has been instigated by a number of recent improvements in our analysis, namely regarding the inclusion of the theoretical uncertainties when investigating the reproduction of experimental data sets on the basis of a given "theoretical" solution, modifications in the parametrization of the form factors of the proton and of the pion entering the electromagnetic part of the πN amplitude, and the inclusion of the effects of the variation of the σ-meson mass when fitting the ETH model of the πN interaction to the experimental data. The new analysis of the CHAOS DCSs confirms our earlier conclusions and casts doubt on the value for the πN Σ term, which Stahov, Clement and Wagner have extracted from these data.

Variational Formulation of a Higher Order Nodal Diffusion Method

1990 ◽  
Vol 105 (3) ◽  
pp. 256-270 ◽  
Author(s):  
D. V. Altiparmakov ◽  
Dj. Tomašević
Keyword(s):  
Variational Formulation ◽  
Higher Order ◽  
Diffusion Method ◽  
Nodal Diffusion

CFD Investigation of Thermal-Hydraulic Behaviors in Full Reactor Core for Sodium-Cooled Fast Reactor

2018 ◽  
Author(s):  
Jing Chen ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
Kui Zhang ◽  
Mingjun Wang ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Three Dimensional ◽  
Reactor Core ◽  
Heat Removal ◽  
The Core ◽  
Computational Fluid Dynamics Cfd ◽  
Heat Removal System ◽  
China Experimental Fast Reactor ◽  
Full Core

As the first developmental step of the sodium-cooled fast reactor (SFR) in China, the pool-type China Experimental Fast Reactor (CEFR) is equipped with the openings and inter-wrapper space in the core, which act as an important part of the decay heat removal system. The accurate prediction of coolant flow in the reactor core calls for complete three-dimensional calculations. In the present study, an investigation of thermal-hydraulic behaviors in a 180° full core model similar to that of CEFR was carried out using commercial Computational Fluid Dynamics (CFD) software. The actual geometries of the peripheral core baffle, fluid channels and narrow inter-wrapper gap were built up, and numerous subassemblies (SAs) were modeled as the porous medium with appropriate resistance and radial power distribution. First, the three-dimensional flow and temperature distributions in the full core under normal operating condition are obtained and quantitatively analyzed. And then the effect of inter-wrapper flow (IWF) on heat transfer performance is evaluated. In addition, the detailed flow path and direction in local inter-wrapper space including the internal and outlet regions are captured. This work can provide some valuable understanding of the core thermal-hydraulic phenomena for the research and design of SFRs.

scholarly journals Nuclear policy-nuclear fuels for peaceful, safer co-generation and environmentally benign applications

2020 ◽  
Vol 9 (3) ◽  
pp. 724
Author(s):  
Syazwani Mohd Fadzil ◽  
Shafi Qureshi ◽  
Sekhar Basu ◽  
K. Kasturirangan ◽  
Anil Kakodkar ◽  
...  
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Reactor Core ◽  
Negative Temperature ◽  
Hydrogen Gas ◽  
Hydraulic Diameter ◽  
Nuclear Fuels ◽  
Spent Fuel ◽  
Fuel Temperature ◽  
Nuclear Policy

Here, safer nuclear fuels which can sustain in the high temperature and fluence environment of the reactor core are investigated to utilize nuclear energy peacefully. At Nuclear Fuel Complex in Hyderabad, nuclear fuels are being manufactured which are best suited for high temperature and fluence environment of the reactor core even in accidental scenarios. In this paper, nuclear fuels manufactured at NFC, Hyderabad are presented. The developed nuclear fuels have higher equivalent hydraulic diameter and breeding capability to produce U^233. Nuclear fuels having higher equivalent hydraulic diameter reduce the reactor core temperature substantially. These fuels have negative temperature coefficient of reactivity. Thus, in case of an accident, the fuel temperature never exceeds the safety limit. Therefore, the thermal heat available across the secondary of a heat exchanger can be utilized for different industrial processes. This allows the development of key technologies, such as safer co-generation of electricity and Hydrogen. The Three-Stage Indian Nuclear Power Program has been explained for nuclear disarmament. The product Hydrogen gas has been utilized in many ways for different applications. Moreover, the processing of iron ore with the energy obtained from the IHX secondary side, eliminates the burning of coals and CO2 emissions into the environment. Several radioisotopes have been developed for medical applications from spent fuel.  

scholarly journals Investigation of Coolant Local Hydrodynamics in the Mixed Core of the VVER Reactor

2020 ◽  
Vol 63 (2) ◽  
pp. 151-162
Author(s):  
S. M. Dmitriev ◽  
A. V. Gerasimov ◽  
A. A. Dobrov ◽  
D. V. Doronkov ◽  
A. N. Pronin ◽  
...  
Keyword(s):  
Cross Sections ◽  
Experimental Studies ◽  
Reactor Core ◽  
Cross Flow ◽  
Coolant Flow ◽  
Vver Reactor ◽  
Computational Hydrodynamics ◽  
Cross Section Area ◽  
Fuel Assemblies ◽  
Flow Pressure

The article presents the results of experimental studies of the local hydrodynamics of the coolant flow in the mixed core of the VVER reactor, consisting of the TVSA-T and TVSA-T mod.2 fuel assemblies. Modeling of the flow of the coolant flow in the fuel rod bundle was carried out on an aerodynamic test stand. The research was carried out on a model of a fragment of a mixed core of a VVER reactor consisting of one TVSA-T segment and two segments of the TVSA-T.mod2. The flow pressure fields were measured with a five-channel pneumometric probe. The flow pressure field was converted to the direction and value of the coolant velocity vector according to the dependencies obtained during calibration. To obtain a detailed data of the flow, a characteristic cross-section area of the model was selected, including the space cross flow between fuel assemblies and four rows of fuel rods of each of the TVSA fuel assemblies. In the framework of this study the analysis of the spatial distribution of the projections of the velocity of the coolant flow was fulfilled that has made it possible to pinpoint regularities that are intrinsic to the coolant flowing around spacing, mixing and combined spacing grates of the TVSA. Also, the values of the transverse flow of the coolant caused by the flow along hydraulically nonidentical grates were determined and their localization in the longitudinal and cross sections of the experimental model was revealed. Besides, the effect of accumulation of hydrodynamic flow disturbances in the longitudinal and cross sections of the model caused by the staggered arrangement of hydraulically non-identical grates was determined. The results of the study of the coolant cross flow between fuel assemblies interaction, i.e. between the adjacent TVSA-T and TVSA-T mod.2 fuel assemblies were adopted for practical use in the JSC of “Afrikantov OKB Mechanical Engineering” for assessing the heat engineering reliability of VVER reactor cores; also, they were included in the database for verification of computational hydrodynamics programs (CFD codes) and for detailed cell-based calculation of the reactor core.

scholarly journals Evaluation of the Compton (Incoherent) and Rayleigh (Coherent) Differential Cross Sections of Scattering for Rhodium 103Rh45 and Tantalum181Ta73 by Employing CSC model

2012 ◽  
Vol 9 (3) ◽  
pp. 554-558 ◽  
Author(s):  
Baghdad Science Journal
Keyword(s):  
Differential Cross Section ◽  
Cross Section ◽  
Cross Sections ◽  
Differential Cross ◽  
Form Factors ◽  
Differential Cross Sections ◽  
The Cross

The differential cross section for the Rhodium and Tantalum has been calculated by using the Cross Section Calculations (CSC) in range of energy(1keV-1MeV) . This calculations based on the programming of the Klein-Nashina and Rayleigh Equations. Atomic form factors as well as the coherent functions in Fortran90 language Machine proved very fast an accurate results and the possibility of application of such model to obtain the total coefficient for any elements or compounds.

scholarly journals ω AND η (η') MESONS FROM NN AND ND COLLISIONS AT INTERMEDIATE ENERGIES

2010 ◽  
Vol 19 (05n06) ◽  
pp. 894-902 ◽  
Author(s):  
L. P. KAPTARI ◽  
B. KÄMPFER
Keyword(s):  
Cross Sections ◽  
Good Description ◽  
Meson Production ◽  
Form Factors ◽  
Angular Distributions ◽  
Transition Form ◽  
Vector Meson Production ◽  
Intermediate Energies ◽  
Electron Production ◽  
Pseudo Scalar

The production of pseudo scalar, η, η′, and vector, ω, ρ, ϕ, mesons in NN collisions at threshold-near energies is analyzed within a covariant effective meson-nucleon theory. It is shown that a good description of cross sections and angular distributions, for vector meson production, can be accomplished by considering meson and nucleon currents only, while for pseudo scalar production an inclusion of nucleon resonances is needed. The di-electron production from subsequent Dalitz decay of the produced mesons, η′ → γγ* → γe+e- and ω → πγ* → πe+e- is also considered and numerical results are presented for intermediate energies and kinematics of possible experiments with HADES, CLAS and KEK-PS. We argue that the transition form factor ω → γ*π as well as η′ → γ*γ can be defined in a fairly model independent way and the feasibility of an experimental access to transition form factors is discussed.

Performance Analysis of Nek5000 for Single-Assembly Calculations

2018 ◽  
Author(s):  
Elia Merzari ◽  
Ronald Rahaman ◽  
Misun Min ◽  
Paul Fischer
Keyword(s):  
Degrees Of Freedom ◽  
Hybrid Approach ◽  
Reactor Core ◽  
Navier Stokes ◽  
Benchmark Problems ◽  
Application Development ◽  
Pressurized Water ◽  
Eddy Simulation ◽  
Common Framework ◽  
Full Core

The ExasSMR project focuses on the exascale application of single and coupled Monte Carlo (MC) and computational fluid dynamics (CFD) physics. Work is based on the Shift MC depletion, OpenMC temperature-dependent MC, and Nek5000 CFD codes. The application development objective is to optimize these applications for exascale execution of full-core simulations and to modularize and integrate them into a common framework for coupled and individual execution. Given the sheer scale of nuclear systems, the main algorithmic driver on the CFD side is weak scaling. The focus for the first four years of the project is on demonstrating scaling up to a full reactor core for high-fidelity simulations of turbulence. Full-core fluid calculations aimed at better predicting the steady-state performance will be conducted with a hybrid approach in which large eddy simulation is used to simulate a portion of a core and unsteady Reynolds-averaged Navier-Stokes handles the rest. This zonal hybrid approach provides an additional scaling dimension besides the number of assemblies. The present manuscript focuses on performance assessment using assembly-level simulations with Nek5000. We discuss the development of two benchmark problems: a subchannel (single-rod) problem to assess internode performance and a larger full-assembly problem representative of a small modular reactor (SMR). We note that current SMR assemblies are considerably simpler than pressurized water reactor assemblies since they contain no mixing vanes. This feature allows for considerable reduction in the degrees of freedom required to simulate the full core. We discuss profiling and scaling results with Nek5000, describe current bottlenecks and potential limitations of the approach, and suggest optimizations for future investigation.

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