scholarly journals Neutron Transport Simulations of RBMK Fuel Assembly Using Multigroup and Continuous Energy Data Libraries within the SCALE Code

2021 ◽  
Vol 2021 ◽  
pp. 1-11
Author(s):  
Andrius Slavickas ◽  
Raimondas Pabarčius ◽  
Aurimas Tonkūnas ◽  
Sigitas Rimkevičius
Keyword(s):  
Fuel Assembly ◽  
Cross Section ◽  
Neutron Transport ◽  
Accurate Estimation ◽  
Energy Data ◽  
Continuous Energy ◽  
Multigroup Data ◽  
Data Libraries ◽  
Rbmk Fuel ◽  
Data Library

The neutron transport simulations of RBMK-1500 fuel assembly were performed using both multigroup and continuous energy data libraries available within the SCALE code system in order to validate its suitability for the estimation of RBMK neutronic characteristics. The resonance processing of cross section, involved in the preparation of the multigroup data library, has a significant impact on neutron transport calculations. Standard Dancoff factors (DFs) used for the heterogeneous geometry of RBMK fuel assembly are insufficient for the accurate estimation of resonance self-shielding. Thus, the SCALE module MCDancoff was used in this study to determine location-specific DFs. The results of RBMK-1500 fuel assembly simulations using standard and user-defined DFs were compared. In addition, the continuous energy (CE) cross-section data library was applied for the benchmark calculations. The impact of different nuclear data libraries on neutron transport simulations was tested as well. It was found out that the usage of the multigroup data libraries generates some deviation from the reference simulations obtained with CE libraries. The CE library based on the estimated ENDF/B-VII.1 data proved to be the best alternative for neutron transport simulations of RBMK fuel assembly.

scholarly journals Impact of the cross section library on 93mNb activity in VVER-1000 reactor dosimetry

2020 ◽  
Vol 225 ◽  
pp. 03009
Author(s):  
P. Haroková ◽  
M. Lovecký
Keyword(s):  
Cross Section ◽  
Neutron Transport ◽  
Nuclear Data ◽  
Nuclear Data Library ◽  
The Cross ◽  
Materials Used ◽  
The Impact ◽  
Data Library ◽  
Reactor Dosimetry

One of the objectives of reactor dosimetry is determination of activity of irradiated dosimeters, which are placed on reactor pressure vessel surface, and calculation of neutron flux in their position. The uncertainty of calculation depends mainly on the choice of nuclear data library, especially cross section used for neutron transport and cross section used as the response function for neutron activation. Nowadays, number of libraries already exists and can be still used in some applications. In addition, new nuclear data library was recently released. In this paper, we have investigated the impact of the cross section libraries on activity of niobium, one of the popular materials used as neutron fluence monitor. For this purpose, a MCNP6 model of VVER-1000 was made and we have compared the results between 14 commonly used cross section libraries. A possibility of using IRDFF library in activation calculations was also considered. The results show good agreement between the new libraries, with the exception of the most recent ENDF/B-VIII.0, which should be further validated.

scholarly journals Adaptation of HELIOS Models for WWER-440 Fuel Assemblies to Application by the TRACE/PARCS Program

2019 ◽  
pp. 46-51
Author(s):  
I. Ovdiienko ◽  
O. Kuchyn ◽  
M. Ieremenko ◽  
P. Vlasenko
Keyword(s):  
Fuel Assembly ◽  
Cross Section ◽  
Neutron Transport ◽  
Reactor Core ◽  
Neutron Cross Section ◽  
Safety Analysis ◽  
Neutron Transport Equation ◽  
Fuel Assemblies ◽  
Adaptation Procedure

The preparation of a few-group neutron cross-section library is an important step in implementation of the computer packages that are based on solution of the neutron transport equation in the few-group diffusion approximation into the safety analysis practices. The accuracy of modelling the physical neutron kinetic processes in the reactor core directly depends on the quality of few-group cross-section library. It is important to note that such cross-section library should be prepared in the format applied in the computer package and with use of a spectral code that models the fuel assembly quite adequately. The best option for preparing the few-group neutron crosssection library for the PARCS few-group diffusion code, which is being introduced into SSTC NRS safety analysis practices as a part of the TRACE/PARCS coupled neutron kinetic/thermal hydraulic package, is to adapt the previously developed and validated models of fuel assemblies for the HELIOS spectral program. The adaptation procedure for HELIOS models for WWER-440 including the fuel follower and transition part forming the input file structure required for correct work of the GenPMAXS program is presented. The approaches to the choice of reference states and branch parameters in the PARCS code format are presented. The results from correctness analysis of the adaptation of the HELIOS WWER-440 fuel assembly computer models are presented. The results are based on a comparative analysis of the fuel assembly multiplication properties obtained by the HELIOS model that was developed for preparation of the cross-section libraries for the DYN3D program (validated and widely used at SSTC NRS at present), and by the HELIOS model that was adapted for the GENPMAX program.

Development and Verification of a New Nuclear Data Processing System NECP-Atlas

2018 ◽  
Author(s):  
Jialong Xu ◽  
Tiejun Zu ◽  
Liangzhi Cao ◽  
Hongchun Wu
Keyword(s):  
Data Processing ◽  
Cross Section ◽  
Processing System ◽  
Nuclear Data ◽  
Data File ◽  
Cross Section Data ◽  
Section Data ◽  
Data Libraries ◽  
Data Library ◽  
Evaluated Nuclear Data

To process the evaluated nuclear data file (ENDF) libraries and generate the cross section data library for neutronics calculations, a new nuclear data processing system NECP-Atlas was developed by Nuclear Engineering Computational Physics Lab. of Xi'an Jiaotong University. Meanwhile, some flaws of the current widely used nuclear data processing systems were made up. Some new methods and techniques were proposed and integrated into NECP-Atlas. NECP-Atlas could process ENDF and generate point-wise evaluated nuclear data file (PENDF) and the multigroup cross section data library in WIMS-D format. Verification of NECP-Atlas was carried out by comparing the keff values for WLUP benchmark cases and benchmark experiments in the ICSBEP handbook using cross section data libraries processed by NECP-Atlas with those by NJOY2016. The results showed that NECP-Atlas processes the ENDF correctly and generates more reliable cross section data libraries.

scholarly journals NUMA-AWARE DATA MANAGEMENT FOR NEUTRON CROSS SECTION DATA IN CONTINUOUS ENERGY MONTE CARLO NEUTRON TRANSPORT SIMULATION

2021 ◽  
Vol 247 ◽  
pp. 04020
Author(s):  
Nicolas Denoyelle ◽  
John Tramm ◽  
Kazutomo Yoshii ◽  
Swann Perarnau ◽  
Pete Beckman
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Neutron Transport ◽  
Data Allocation ◽  
Next Generation ◽  
Cross Section Data ◽  
Section Data ◽  
Continuous Energy ◽  
Computational Performance ◽  
Mc Simulation

The calculation of macroscopic neutron cross-sections is a fundamental part of the continuous-energy Monte Carlo (MC) neutron transport algorithm. MC simulations of full nuclear reactor cores are computationally expensive, making high-accuracy simulations impractical for most routine reactor analysis tasks because of their long time to solution. Thus, preparation of MC simulation algorithms for next generation supercomputers is extremely important as improvements in computational performance and efficiency will directly translate into improvements in achievable simulation accuracy. Due to the stochastic nature of the MC algorithm, cross-section data tables are accessed in a highly randomized manner, resulting in frequent cache misses and latency-bound memory accesses. Furthermore, contemporary and next generation non-uniform memory access (NUMA) computer architectures, featuring very high latencies and less cache space per core, will exacerbate this behaviour. The absence of a topology-aware allocation strategy in existing high-performance computing (HPC) programming models is a major source of performance problems in NUMA systems. Thus, to improve performance of the MC simulation algorithm, we propose a topology-aware data allocation strategies that allow full control over the location of data structures within a memory hierarchy. A new memory management library, known as AML, has recently been created to facilitate this mapping. To evaluate the usefulness of AML in the context of MC reactor simulations, we have converted two existing MC transport cross-section lookup “proxy-applications” (XSBench and RSBench) to utilize the AML allocation library. In this study, we use these proxy-applications to test several continuous-energy cross-section data lookup strategies (the nuclide grid, unionized grid, logarithmic hash grid, and multipole methods) with a number of AML allocation schemes on a variety of node architectures. We find that the AML library speeds up cross-section lookup performance up to 2x on current generation hardware (e.g., a dual-socket Skylake-based NUMA system) as compared with naive allocation. These exciting results also show a path forward for efficient performance on next-generation exascale supercomputer designs that feature even more complex NUMA memory hierarchies.

scholarly journals p-Wave Strength in the Low-energy 7Li(p,?0)8Be Cross Section

1995 ◽  
Vol 48 (5) ◽  
pp. 813 ◽  
Author(s):  
FC Barker
Keyword(s):  
Angular Distribution ◽  
Shell Model ◽  
Cross Section ◽  
Low Energy ◽  
P Wave ◽  
Model Calculations ◽  
Energy Data

Recent fits to low-energy 7Li(p, "Yo)8Be angular distribution and analysing power data suggested a large p-wave strength. It is shown that acceptable fits to the data can be obtained by attributing the p-wave Ml contributions to the tails of the 17 �64 and 18 �15 MeV 1+ levels of 8Be, with p-wave strengths much less than those obtained previously, but only if some of the spectroscopic amplitudes have signs opposite to those suggested by shell model calculations and/or a fit to higher-energy data.

Estimating discharge in rivers through the combined use of dimensionless isovels and point velocity measurements

2017 ◽  
Vol 48 (3) ◽  
pp. 616-633 ◽  
Author(s):  
G. Farina ◽  
S. Alvisi ◽  
M. Franchini
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
Accurate Estimation ◽  
Velocity Measurements ◽  
Combined Use ◽  
Point Velocity ◽  
New Formulation ◽  
The Relationship ◽  
River Cross Section

This paper presents a procedure for estimating discharge in a river cross-section based on the combined use of dimensionless isovels and point velocity measurements. Specifically, taking the Biot–Savart law on the magnetic field induced by an electric current in a wire as their basis as already done by other researchers, the authors propose a new formulation of the relationship characterizing the effect of the wetted perimeter on the range of velocities in a cross-section in order to take explicit account of roughness, expressed by means of Manning's coefficient. Once appropriately nondimensionalized, the isoeffect contours can be read as dimensionless isovels. Assuming in situ velocity measurements are available, discharge at a cross-section can be computed using two different methods. The proposed procedure was applied to six case studies characterized by river cross-sections which differed greatly from one another. The results show that the two methods proposed for estimating discharge lead to equivalent outcomes, and in all the cases the procedure as a whole enables a sufficiently accurate estimation of discharge, even when it is based on a limited number of velocity measurements or on the measurement of maximum surface-water velocity alone.

Study on SCWR Fuel Assembly Design With Thermal-Hydraulic and Neutron-Physical Characteristics

2010 ◽  
Author(s):  
Zhenyang Li ◽  
Tao Zhou ◽  
Canhui Sun ◽  
Xiaozhuang Liu
Keyword(s):  
Fuel Assembly ◽  
Cross Section ◽  
Power Distribution ◽  
Neutron Cross Section ◽  
Physical Characteristics ◽  
Supercritical Pressure ◽  
Light Water ◽  
Assembly Design ◽  
Temperature Distributions ◽  
Parallel Channels

Physical characteristics of the coolant in the Supercritical-pressure Light Water Cooled Reactor (SCWR) vary greatly near the pseudo-critical point, which will cause variations of core neutron cross section and then bring about power perturbation. Further it will prompt the corresponding thermal parameters of supercritical water changed, and form feedback action, finally resulting in intensely coupled thermal-hydraulics and neutron-physical. Proper fuel assembly has been selected as research object, and the model of multiple parallel channels has been established. In view of this model, using DRAGON code for neutron-physical calculations and developing corresponding thermal-hydraulic programs, and then achieve coupling them through appropriate data interface, the calculation platform established. Finally the power distribution and the corresponding parameters temperature distributions in the model have been predicted. On account of deficiencies reflected in calculations, such as the heterogeneous power distribution, fuel assembly geometry has been changed, for instance the proper peripheral moderator wall has been added based on the preceding assembly, then do the coupling calculations and analyze the results. Comparisons between different results have been made, and the expected aim has been reached, which can prove the rationality of assembly modifications and meanwhile prove the usability of the calculation platform. Thus, modified assembly and the calculation platform could be the calculation foundation in future designs of SCWR.

Continuous-energy Monte Carlo neutron transport on GPUs in the Shift code

2019 ◽  
Vol 128 ◽  
pp. 236-247 ◽  
Author(s):  
Steven P. Hamilton ◽  
Thomas M. Evans
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Continuous Energy
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