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.

Measurement of Selected Differential Cross Sections in 235U Spectrum

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Nicola Burianová ◽  
Michal Košt´ál ◽  
Martin Schulc ◽  
Jan Šimon ◽  
Martin Mareček ◽  
...  
Keyword(s):  
Cross Sections ◽  
Nuclear Data ◽  
Reaction Rates ◽  
Fission Neutron ◽  
Data Libraries ◽  
Nuclear Data Libraries ◽  
Prompt Fission Neutron Spectrum ◽  
Prompt Fission ◽  
Prompt Fission Neutron ◽  
Good Agreement

This paper describes the measurement of 55Mn(n,2n) and 127I(n,2n) reaction rates in a well-defined reactor field in a special core of LR-0 reactor. The reaction rates were derived using gamma-spectrometry by measuring gamma activities of irradiated MnO2 and NaI samples at a high purity germanium (HPGe) detector. The spectral average cross section (SACS) in 235U prompt fission neutron spectrum (PFNS) was experimentally determined to be 0.2393 ± 0.015 × 10−3 b for 55Mn and 1.2087 ± 0.052 × 10−3 b for 127I. These obtained results were compared with calculations by MCNP6 code using ENDF/B VII.1, ENDF/B VII, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND- 2010, CENDL-3.1, and IRDFF nuclear data libraries. In a case of 55Mn, a good agreement with ENDF/B VII.1, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND, and CENDL 3.1 nuclear data libraries was found, where C/E−1 is 0.1%, while IRDFF underestimated by about 15.8%. In the case of 127I, more significant discrepancies were found, where JENDL 3.3 and JENDL 4 overestimate the result by about 31.3%.

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.

Criticality Calculations and Basic Sensitivity/Uncertainty Investigation of LR-0 Benchmark Core

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Tomáš Czakoj ◽  
Evžen Losa
Keyword(s):  
Lithium Fluoride ◽  
Three Dimensional ◽  
Reactor Core ◽  
Nuclear Data ◽  
Monte Carlo Code ◽  
Multiplication Factor ◽  
Reference Case ◽  
Energy Version ◽  
Data Libraries ◽  
Nuclear Data Libraries

Three-dimensional Monte Carlo code KENO-VI of SCALE-6.2.2 code system was applied for criticality calculation of the LR-0 reactor core. A central module placed in the center of the core was filled by graphite, lithium fluoride-beryllium fluoride (FLIBE), and lithium fluoride-sodium fluoride (FLINA) compounds. The multiplication factor was obtained for all cases using both ENDF/B-VII.0 and ENDF/B-VII.1 nuclear data libraries. Obtained results were compared with benchmark calculations in the MCNP6 using ENDF/B-VII.0 library. The results of KENO-VI calculations are found to be in good agreement with results obtained by the MCNP6. The discrepancies are typically within tens of pcm excluding the case with the FLINA filling. Sensitivities and uncertainties of the reference case with no filling were determined by a continuos-energy version of the TSUNAMI sequence of SCALE-6.2.2. The obtained uncertainty in multiplication factor due to the uncertainties in nuclear data is about 650 pcm with ENDF/B-VII.1.

Sensitivity to Nuclear Data Libraries in the Physics Core Characteristics of Conceptual Thorium Breeders

2011 ◽  
Vol 59 (2(3)) ◽  
pp. 1361-1364
Author(s):  
V. Jagannathan ◽  
U. Pal ◽  
R. Karthikeyan ◽  
A. Srivastava ◽  
S. A. Khan
Keyword(s):  
Nuclear Data ◽  
Data Libraries ◽  
Core Characteristics ◽  
Nuclear Data Libraries

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.

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