scholarly journals Nuclear data activities of the EUROfusion consortium

2020 ◽  
Vol 239 ◽  
pp. 21001
Author(s):  
Ulrich Fischer ◽  
Marilena Avrigeanu ◽  
Vlad Avrigeanu ◽  
Alexander Konobeyev ◽  
Ivo Kodeli ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Data ◽  
Data Bank ◽  
Cross Section Data ◽  
Related Data ◽  
Section Data ◽  
Fusion Applications ◽  
Uncertainty Analyses ◽  
Data Libraries ◽  
Evaluation Work

The activities of the EUROfusion consortiums on the development of high quality nuclear data for fusion applications are presented. The activities, implemented in the Power Plant Physics and Technology (PPPT) programme of EUROfusion, include nuclear data evaluations for neutron and deuteron induced reactions and the production of related data libraries which satisfy the needs for nuclear analyses of the DEMO fusion power plant and the IFMIF-DONES neutron source. The activities are closely linked to the JEFF initiative of the NEA Data Bank. The evaluation work is complemented by extensive benchmark, sensitivity and uncertainty analyses to check the performance of the evaluated cross-section data and libraries against integral experiments.

scholarly journals Comparative survey of evaluated nuclear data libraries for fusion-relevant neutron activation spectrometry

2020 ◽  
Vol 239 ◽  
pp. 21003
Author(s):  
Prasoon Raj ◽  
Ulrich Fischer ◽  
Axel Klix ◽  
JET Contributors
Keyword(s):  
Cross Section ◽  
Nuclear Physics ◽  
Nuclear Data ◽  
Reaction Rates ◽  
Neutron Field ◽  
Physics Institute ◽  
Section Data ◽  
Fusion Applications ◽  
Data Files ◽  
Data Libraries

The neutron flux-spectrum in a fusion device is frequently determined with activation foils and adjustment of a guess-spectrum in unfolding codes. Spectral-adjustment being a rather complex and uncertain procedure, we are carefully streamlining and evaluating it for upcoming experiments. Input nuclear cross-section data holds a vital position in this. This paper presents a survey of common dosimetry reactions and available data files relevant for fusion applications. While the IRDFF v1.05 library is the recommended source, many reactions of our interest are found missing in this. We investigated other standard sources: ENDF/B-VIII.0, EAF-2010, TENDL-2017, JENDL-4.0 etc. And, we analysed two experiments to ascertain the sensitivity of the spectral adjustment to the choice of nuclear data. One was performed with D-D (approx. 2.5 MeV peak) neutrons at the Joint European Torus (JET) machine and another with a white neutron field (approx. 33 MeV endpoint energy) at Nuclear Physics Institute (NPI) of Řež. Choice of cross-section source has affected the integral fluxes (<5%), reaction rates (<10%), total fluxes in some sensitive energy-regions (>20%) and individual group fluxes (<30%). Based on this experience, essential qualitative conclusions are made to improve the fusion activation-spectrometry.

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.

The Development and Validation of Nuclear Cross Section Processing Code for Reactor-RXSP

2013 ◽  
Author(s):  
Jiankai Yu ◽  
Songyang Li ◽  
Kan Wang ◽  
Guanbo Wang ◽  
Ganglin Yu
Keyword(s):  
Cross Section ◽  
Nuclear Data ◽  
Original Data ◽  
Data File ◽  
Cross Section Data ◽  
Reactor Physics ◽  
R Matrix ◽  
Section Data ◽  
Nuclear Cross Section ◽  
Long Time

The accuracy of the nuclear cross section data is a prerequisite for the accuracy of reactor physics calculations. The RXSP(Reactor Cross Section Processing Code) which is developed by REAL (Reactor Engineering Analysis Laboratory) of Department of Engineering Physics in Tsinghua University, has changed the situation in China that nuclear cross section processing has been dependent of NJOY for a long time. The key methods such as fast Doppler broadening, thermal libraries interpolation, and OpenMP parallel acceleration, can be achieved with RXSP. This code is able to process the original data of ENDF/B (Evaluated Nuclear Data File/B) efficiently and accurately to produce the continuous energy point cross section data which is necessary for RMC. By comparing with NJOY, The microscopic and macroscopic verification shows that RXSP has the same accuracy as NJOY while RXSP has saved greatly the processing time to meet the efficient demand in the frequent reactor physics-thermal-hydraulic coupling calculations to solve the complex questions related on a large number of materials and temperature. In addition, RXSP make it available to process the resonance parameters of the R-matrix Limited format.

Benchmarking of the FENDL-3 Neutron Cross-section Data Starter Library for Fusion Applications

2014 ◽  
Vol 120 ◽  
pp. 230-234 ◽  
Author(s):  
U. Fischer ◽  
M. Angelone ◽  
T. Bohm ◽  
K. Kondo ◽  
C. Konno ◽  
...  
Keyword(s):  
Cross Section ◽  
Neutron Cross Section ◽  
Cross Section Data ◽  
Section Data ◽  
Fusion Applications

scholarly journals Nuclear data sensitivity and uncertainty analyses on the first core criticality of the RSG GAS multipurpose research reactor

2020 ◽  
Vol 239 ◽  
pp. 22011 ◽  
Author(s):  
Peng Hong Liem ◽  
Zuhair ◽  
Donny Hartanto
Keyword(s):  
Nuclear Data ◽  
Worst Case ◽  
Control Rod ◽  
Nuclear Data Library ◽  
Uncertainty Analyses ◽  
Effective Neutron ◽  
Data Libraries ◽  
And Control ◽  
Analytical Tools ◽  
Nuclear Data Libraries

The results of criticality, sensitivity and uncertainty (S\U) analyses on the first core criticality of the Indonesian 30 MWth Multipurpose Reactor RSG GAS (MPR-30) using the recent nuclear data libraries (ENDF/B-VII.1 and JENDL-4.0) and analytical tools available at present (WHISPER-1.1) are presented. Two groups of criticality benchmark cases were carefully selected from the experiments conducted during the first criticality approach and control rod calibrations. The C/E values of effective neutron multiplication factor (k) for the worst case was found around 1.005. Large negative sensitivities were found in (n,e-mail:γ) reaction of H-1, U-235, Al-27, U-238 and Be-9 while large positive sensitivities were found in U-235 (total nu and fission), H-1 (elastic), Be-9 (free gas, elastic) and H-1 S(α,β) (lwtr.20t, inelastic). The S\U analysis results concluded that the uncertainties of k originated from the nuclear data were found around 0.6% which covered well the [C/E-1] values. Differences in the sensitivities amongst the two nuclear data libraries were also identified, and recommendation for improving the nuclear data library was given.

scholarly journals A COMPARISON OF OXIDE DECAY HEAT SIMULATIONS AND NUCLEAR DATA LIBRARIES WITH FUSION IRRADIATION EXPERIMENTS

2021 ◽  
Vol 247 ◽  
pp. 09011
Author(s):  
L W Packer ◽  
M Angelone ◽  
M R Gilbert ◽  
S Loreti ◽  
C R Nobs ◽  
...  
Keyword(s):  
Nuclear Data ◽  
Data Sets ◽  
Section Data ◽  
Fusion Neutron ◽  
Decay Heat ◽  
Decay Profile ◽  
Data Points ◽  
Fast Extraction ◽  
Data Libraries ◽  
Nuclear Data Libraries

We compare existing experimental decay heat data sets measured at the JAEA fusion neutron source (FNS), which employed a fast extraction rabbit system that in certain cases allowed the measurements to capture, at short cooling times, the decay profile of 16N in a range of oxides. Focussing on those data points and timescales that can be attributed to 16N, we compare measurements to simulations performed using the FISPACT-II inventory code together with evaluated nuclear data libraries. Making small corrections for other contributions at these short times, we derive integral cross section data estimates for 16O(n,p)16N from 12 oxide sample measurements and compare with previously obtained measurements in the IAEA EXFOR database and evaluations in the nuclear data libraries themselves.

233U Evaluation Comparison Study

2014 ◽  
Author(s):  
Mohammad Alrwashdeh ◽  
Aniseh Abdalla ◽  
Kan Wang
Keyword(s):  
Cross Sections ◽  
Least Square Method ◽  
Computer Code ◽  
Data Bank ◽  
Least Square ◽  
Mathematical Methods ◽  
Bayes Method ◽  
Critical System ◽  
Section Data ◽  
Data Libraries

The aim of this study is to investigate the availability and accuracy of the cross section data for 233U to perform the calculations of the critical system. Two evaluated data libraries are available, U.S. data bank (ENDF) and the Japanese data bank (JENDL), by using BAYES method for resonance parameters available in SAMMY code and weighted least square method with nonlinear regression by using FITWR computer code. Evaluation of the 233U has been investigated by using of SAMMY code, in order to generate a useful data base for critical calculations, the computer code FITWR for experimental - experimental data fitting show same results obtained from Bayes method included within SAMMY code, with a slight deference in the results at the evaluated cross sections due to different mathematical methods have different results.

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.

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