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.

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 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.

New empirical formula for calculating (n,p) reaction cross-sections at 14.5MeV neutrons

2020 ◽  
Vol 29 (08) ◽  
pp. 2050052
Author(s):  
Dashty T. Akrawy ◽  
Ali H. Ahmed ◽  
E. Tel ◽  
A. Aydin ◽  
L. Sihver
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
Empirical Formula ◽  
Nuclear Data ◽  
Reaction Cross ◽  
Section Data ◽  
Nuclear Data Library ◽  
Number Formula ◽  
New Formula ◽  
Reaction Cross Sections

An empirical formula to calculate the ([Formula: see text], [Formula: see text] reaction cross-sections for 14.5[Formula: see text]MeV neutrons for 183 target nuclei in the range [Formula: see text] is presented. Evaluated cross-section data from TENDL nuclear data library were used to test and benchmark the formula. In this new formula, the nonelastic cross-section term is replaced by the atomic number [Formula: see text], while the asymmetry parameter-dependent exponential term has been retained. The calculated results are presented in comparison with the seven previously published formulae. We show that the new formula is significantly in better agreement with the measured values compared to previously published formulae.

Temperature Dependence of the Westcott g-factor for Neutron Reactions in Activation Analysis

1999 ◽  
Vol 71 (12) ◽  
pp. 2309-2315 ◽  
Author(s):  
N. E. Holden
Keyword(s):  
Cross Section ◽  
Neutron Capture ◽  
Nuclear Reactions ◽  
Nuclear Data ◽  
Reaction Rates ◽  
Nuclear Data Library ◽  
Neutron Reactions ◽  
Using Data ◽  
Data Library ◽  
G Factors

The Westcott g-factors, which allow the user to determine reaction rates for nuclear reactions taking place at various temperatures, have been calculated using data from the Evaluated Neutron Nuclear Data Library, ENDF/B-VI. Nuclides chosen have g-factors which are significantly different from unity and result in different reaction rates compared to nuclides whose neutron capture cross section varies as the reciprocal of the neutron velocity. Values are presented as a function of temperature up to 673.16 K (400 °C).

JENDL/ImPACT-2018: a new nuclear data library for innovative studies on transmutation of long-lived fission products

2019 ◽  
Vol 56 (12) ◽  
pp. 1073-1091 ◽  
Author(s):  
Satoshi Kunieda ◽  
Naoya Furutachi ◽  
Futoshi Minato ◽  
Nobuyuki Iwamoto ◽  
Osamu Iwamoto ◽  
...  
Keyword(s):  
Fission Products ◽  
Nuclear Data ◽  
Nuclear Data Library ◽  
Data Library

The Consistency of WIMS-D Library and ACE Format Library

2010 ◽  
Author(s):  
Zhiyan Liu ◽  
Hongchun Wu ◽  
Liangzhi Cao ◽  
Qingjie Liu
Keyword(s):  
Cross Section ◽  
Scaling Function ◽  
Processing System ◽  
Expansion Method ◽  
Nuclear Data ◽  
Calculation Procedure ◽  
Function Expansion ◽  
Nuclear Data Library ◽  
Data Library

Nuclear data library is the cornerstone in the nuclear reactor’s design and calculation. The WIMS-D multi-group library and ACE format library (mainly used in MCNP) is applied frequently in the nuclear calculation. We have developed a new self-shielding calculation procedure based on Wavelets scaling function expansion method. This procedure needs several parts in both WIMS-D and ACE format library. So the consistency of two libraries becomes a very serious problem. This may bring in large errors. In this paper, NJOY cross section processing system is used to produce new WIMS-D and ACE format library from the same ENDF/B data. We compute some homogenous problems using new and old libraries in WIMS-D and ACE format. The results of the two new libraries and the old libraries are compared respectively. It is found that there are consistency problems between the two libraries. The newly produced libraries are more compatible than the old ones.

scholarly journals Generation of Proton- and Alpha-Induced Nuclear Cross-Section Data via Random Forest Algorithm: Production of Radionuclide 111In

2021 ◽  
Vol 11 (15) ◽  
pp. 6969
Author(s):  
Mohamad Amin Bin Hamid ◽  
Hoe Guan Beh ◽  
Yusuff Afeez Oluwatobi ◽  
Xiao Yan Chew ◽  
Saba Ayub
Keyword(s):  
Random Forest ◽  
Cross Section ◽  
Nuclear Data ◽  
Cross Section Data ◽  
Random Forest Algorithm ◽  
Section Data ◽  
Regression Curves ◽  
Nuclear Data Library ◽  
Nuclear Cross Section ◽  
Indium 111

We investigated the generation of proton- and alpha-induced nuclear cross-section data in the production of Indium-111 (111In) for application in nuclear medicine. Here, we are interested in three reaction channels, which are 109Ag (α, 2n), 111Cd (p, n) and 112Cd (p, 2n), in the production of 111In. A random forest algorithm was used to generate nuclear cross-section data by using an experimental nuclear cross-section from the Experimental Nuclear Reaction Data (EXFOR) database as input. Hence, reasonably accurate regression curves of nuclear cross-section data could be produced with the evaluated nuclear data library ENDF/B-VII.0 set as the benchmark.

scholarly journals Verification of CENDL-3.2 Nuclear Data on VENUS-3 Shielding Benchmark by ARES Transport Code

2021 ◽  
Vol 2021 ◽  
pp. 1-13
Author(s):  
Jiaju Hu ◽  
Bin Zhang ◽  
Zhiwei Zong ◽  
Cong Liu ◽  
Yixue Chen
Keyword(s):  
Neutron Flux ◽  
Cross Section ◽  
Cross Sections ◽  
Nuclear Data ◽  
Transport Code ◽  
Nuclear Data Library ◽  
Transport Calculation ◽  
Calculation Results ◽  
Almost All ◽  
Data Library

The recently released CENDL-3.2 nuclear data library is deemed as an important achievement in the field of nuclear data research in China. To verify the applicability of the library to the shielding calculation of PWR and analyze the influence of multigroup cross-section parameters on the shielding calculation, ARES-MACXS module is used to process the MATXS format multigroup library based on CENDL-3.2 to generate multigroup working cross sections for PWR shielding calculation. VENUS-3 experimental facility has a clear and complete geometry. It is often used to test the ability of the advanced transport calculation method of calculating RPV fast neutron flux and to evaluate the accuracy of cross-section library. Different cross-section parameters are chosen for ARES to calculate VENUS-3 benchmark, and equivalent neutron flux of 58Ni(n,p)58Co, 115In(n,n′)115mIn and 27Al(n,α)24Na detectors is calculated according to the data provided by the benchmark report. The numerical results demonstrate that almost all the relative deviations between the calculated results and the experimental results are within 20%, which satisfies the requirement of shielding calculation. CENDL-3.2 is suitable for PWR shielding calculation. The comparison of various cross-section parameters results indicates that multigroup cross-section parameters have large effects on the transport calculation results.

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.

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