New Neutron Cross-Section Measurements at ORELA for Improved Nuclear Data Calculations

2005 ◽  
Author(s):  
K. H. Guber
Keyword(s):  
Cross Section ◽  
Neutron Cross Section ◽  
Nuclear Data

Cross Section and Fission Yields Induced Uncertainty in the Vver-440 Burnup Calculation

2021 ◽  
Author(s):  
Branislav Vrban ◽  
Stefan Cerba ◽  
Jakub Luley ◽  
Filip Osuský ◽  
Vladimir Necas
Keyword(s):  
Cross Section ◽  
Neutron Cross Section ◽  
Nuclear Data ◽  
Reactor Operation ◽  
Precise Estimation ◽  
Stochastic Techniques ◽  
Core Characteristics ◽  
Fission Yields ◽  
Operational Data ◽  
Burnup Calculation

Abstract The properties of nuclear fuel depend on the actual isotopic composition which develops during a reactor operation. In practice, the prediction accuracy of burnup calculations serves as the basis for the future precise estimation of a core lifetime and other safety-based core characteristics. The present study quantifies nuclear data induced uncertainties of nuclide concentrations and multiplication factors in VVER-440 fuel depletion analysis. The well-known SCALE system and the TRITON sequence are used with the NEWT deterministic solver in the SAMPLER module that implements stochastic techniques to assess the uncertainty in computed results. The propagation of uncertainties in neutron cross section and fission yields is studied through the depletion calculation of 2D heterogeneous VVER-440 fuel assembly with an average enrichment of 4.87 wt % of 235U and six gadolinium rods with 3.35 % of Gd2O3. In the paper, fixed nominal depletion conditions are based on the real operational data of the Slovak NPP Bohunice unit 4 during cycle 30. In total 250 cases with uncertain parameters are computed and the results are evaluated by an auxiliary tool.

scholarly journals Effects of different nuclear evaluation data on the RMC keff calculation

2020 ◽  
Vol 239 ◽  
pp. 19005
Author(s):  
Zhang Wenxin ◽  
Qiang shenglong ◽  
Yin qiang ◽  
Cui Xiantao
Keyword(s):  
Cross Section ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Neutron Cross Section ◽  
Nuclear Data ◽  
Section Data ◽  
Calculation Results ◽  
The Impact ◽  
Physical Calculation

Neutron cross section data is the basis of nuclear reactor physical calculation and has a decisive influence on the accuracy of calculation results. AFA3Gassemble is widely used in nuclear power plants. CENACE is an ACE format multiple-temperature continuous energy cross section library that developed by China Nuclear Data Centre. In this paper, we calculated the AFA3G assemble by RMC.We respectively used ENDF6.8/, ENDF/7 and CENACE data for calculation. The impact of nuclear data on RMC calculation is studied by comparing the results of different nuclear data.

scholarly journals Development of a Fast-Spectrum Self-Powered Neutron Detector for Molten Salt Experiments in the Versatile Test Reactor

2021 ◽  
Vol 253 ◽  
pp. 05006
Author(s):  
K. C. Goetz ◽  
S. M. Cetiner ◽  
C. Celik
Keyword(s):  
Cross Section ◽  
Neutron Detector ◽  
Nuclear Reactions ◽  
Neutron Cross Section ◽  
Nuclear Data ◽  
Data File ◽  
Fast Neutrons ◽  
Flux Monitor ◽  
Depth Analysis ◽  
Self Powered

The self-powered neutron detector (SPND) is a widely used flux monitor in thermal nuclear reactors. Although this is a mature technology, the current state of the art is tuned for a thermal neutron spectrum, so many of the devices currently in use lack sensitivity to fast neutrons. Because current in SPNDs is produced through nuclear reactions with the neutron flux inside a reactor, sensitivity in SPNDs is determined by the neutron cross section of the neutron-sensitive portion of the detector, termed the emitter. This neutron cross section drops by orders of magnitude between thermal and fast neutron energies for many emitters in currently used SPNDs, with a corresponding drop in current from the detector. This paper discusses efforts to develop a fast-spectrum self-powered neutron detector (FS-SPND) that is sensitive to neutrons with energies ranging from 0.025 eV up to 1 MeV. An in-depth analysis of Evaluated Nuclear Data File (ENDF)/B-VII.1 neutron-capture cross sections was performed, and four new materials were identified that are suitable emitter candidates for use in measuring fast neutrons. All four materials are stable mid-shell nuclei in the region between doubly magic 132Sn and 208Pb. Each candidate was simulated with the Geant4 Monte Carlo simulation toolkit to optimize overall detector efficiency.

The Testing and Application of Continuous-Energy Neutron Cross Section Library HENDL-ADS/MC

2013 ◽  
Author(s):  
Chong Chen ◽  
Jun Zou ◽  
Dezheng Xu ◽  
Qin Zeng ◽  
Minghuang Wang
Keyword(s):  
Cross Section ◽  
Neutron Energy ◽  
Fission Products ◽  
Neutron Cross Section ◽  
Nuclear Data ◽  
Neutron Shielding ◽  
Section Data ◽  
Nuclear Data Library ◽  
Data Library ◽  
High Neutron

A point-wise cross-section data library HENDL-ADS/MC (Hybrid Evaluated Nuclear Data Library) has been produced by FDS team to do the nuclear analysis for the ADS system. The HENDL-ADS/MC library contained 408 nuclide cross-section files including actinides, fission products and structural materials for neutron energy up to 150 MeV. The nuclear library also contained several sub-libraries with different temperatures. A series of neutron integral experiments and critical safety benchmarks have been performed to test the availability and reliability of the HENDL-ADS/MC data library. To validate and qualify the reliability of the high neutron energy cross section for HENDL-ADS/MC library further, a series of high neutron shielding experiments have been performed using MCNP. The testing results indicated the accuracy and reliability of HENDL-ADS/MC library.

Neutron cross section measurements at ORELA for improved nuclear data and their application

2005 ◽  
Vol 116 (1-4) ◽  
pp. 579-581 ◽  
Author(s):  
K. H. Guber ◽  
L. C. Leal ◽  
R. O. Sayer ◽  
P. E. Koehler ◽  
T. E. Valentine ◽  
...  
Keyword(s):  
Cross Section ◽  
Neutron Cross Section ◽  
Nuclear Data

Treatment of Neutron Cross-Section With Interpolation

2010 ◽  
Author(s):  
Xiang Zhang ◽  
Ganglin Yu ◽  
Guangwen Bi ◽  
Kang Wang
Keyword(s):  
Cross Section ◽  
Neutron Cross Section ◽  
Nuclear Data ◽  
Temperature Dependent ◽  
Benchmark Calculation ◽  
Different Temperatures ◽  
Data Libraries ◽  
Relative Differences ◽  
Nuclear Data Libraries ◽  
Good Agreement

Using NJOY to generate the temperature dependent neutron cross-section is too time-consuming in practice, especially for many nuclides. So an approach involving interpolation between nuclear data libraries at different temperatures is investigated. Based on the ACE data at different temperatures, we used ITND — an neutron cross-section interpolation program, to generate the target temperature ACE data, then we compared it with the ACE data which generated by NJOY at the same temperature. We focused on the interpolation result of 238U, 235U, 232Th, Zr, 16O, 10B and 1H at the temperature of 575K. To that nuclides, several interpolate schemes were studied, and we demonstrated the relative differences, and explain their reasons. Finally we applied these ACE data to benchmark calculation, and good agreement was observed with the benchmark results.

Investigation of the total neutron cross section for natural hafnium on the filtred neutron beam with energy 59 keV

2011 ◽  
Vol 30 (0) ◽  
pp. 142-147
Author(s):  
О. О. Грицай ◽  
А. К. Гримало ◽  
В. В. Колотий ◽  
В. М. Венедиктов ◽  
С. П. Волковецький ◽  
...  
Keyword(s):  
Cross Section ◽  
Neutron Beam ◽  
Neutron Cross Section ◽  
Total Neutron ◽  
Total Neutron Cross Section
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