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.

Neutron Field Shaping Using Graphite for Reaction Rate Measurements

2021 ◽  
Vol 7 (2) ◽  
Author(s):  
Mikita Sobaleu ◽  
Michal Košťál ◽  
Jan Šimon ◽  
Evžen Losa
Keyword(s):  
Cross Section ◽  
Reaction Rate ◽  
Nuclear Data ◽  
Data File ◽  
Reaction Rates ◽  
Reaction Cross ◽  
Neutron Field ◽  
Pure Graphite ◽  
24Na Reaction ◽  
Data Libraries

Abstract Neutron field shaping is the suitable method for validation of cross section in various energy regions. By increasing the share of neutrons of a certain energy interval and decreasing the share of other, a reaction becomes more sensitive to selected neutrons. As a result, reaction cross section can be validated in selected energy regions more precisely. The shaping can be carried out by both neutron filters which are materials with high absorption in some energy region, or by diffusion material changing the shape of neutron spectra by means of slowing down process. In the presented experiments, the neutron field of the light reactor 0 (LR-0) research reactor was shaped by both using graphite blocks inserted into the core and Cd cladding for increasing the epithermal reaction rate share in total reaction rates. The calculations were carried out with the Monte Carlo N-Particle Transport Code 6 (MCNP6) code and the most recent nuclear data libraries. The results in the pure graphite neutron field are in good agreement; in case of Cd cladding, significant discrepancies were reported. In case of the 23Na(n,γ)24Na reaction, overestimation by about 14% was reached in International Reactor Dosimetry and Fusion File (IRDFF-II), results in other libraries are comparable. In case of 58Fe(n,γ)59Fe, the overestimation as high as 18% is reported in IRDFF-II. For 64Zn(n,γ)65Zn reasonable agreement was reached in evaluated nuclear data file (ENDF/B-VIII), where discrepancies in pure graphite neutron field or in case of Cd cladding are about 10–15%.

scholarly journals Yalina-thermal facility neutron characteristic computational study 129I, 237Np and 243Am transmutation reaction rates calculations

2020 ◽  
Vol 239 ◽  
pp. 22013
Author(s):  
Tamara Korbut ◽  
Maksim Kravchenko ◽  
Ivan Edchik ◽  
Sergey Korneev
Keyword(s):  
Monte Carlo ◽  
Computational Study ◽  
Nuclear Research ◽  
Nuclear Data ◽  
Reaction Rates ◽  
Neutron Field ◽  
Minor Actinide ◽  
National Academy Of Sciences ◽  
Data Libraries ◽  
Academy Of Sciences

Present work describes Monte-Carlo calculations of the neutron field and minor actinide transmutation reaction rates within the Yalina-Thermal sub-critical assembly of the Joint Institute for Power and Nuclear Research – Sosny of the National Academy of Sciences of Belarus. The computer model of the facility was prepared for the corresponding calculations via MCU-PD and MCNP Monte-Carlo codes. The model neutron characteristics estimations were performed as well as the nuclear safety analysis. The up-to-date ENDF B/VIII, JEFF 3.3 and JENDL 4.0 nuclear data libraries were used during research.

scholarly journals Measurements of the neutron activation cross sections for Bi at 287 and 370 MeV

2011 ◽  
Vol 1 (1) ◽  
pp. 135-139
Author(s):  
H. Yashima ◽  
S. Sekimoto ◽  
T. Utsunomiya ◽  
K. Ninomiya ◽  
T. Omoto ◽  
...  
Keyword(s):  
Neutron Activation ◽  
Proton Beam ◽  
Cross Section ◽  
Cross Sections ◽  
Nuclear Data ◽  
Neutron Field ◽  
Low Energy ◽  
Energy Neutron ◽  
Data Files ◽  
Activation Cross Sections

AbstractWe present cross section measurements for neutron-induced activation of Bi, at 287 and 370 MeV. These values were derived from the activation method using a quasi-monoenergetic neutron field based on the 7Li(p,n) reaction. In separate experiments, samples were irradiated with neutrons derived from 7Li(p,n) reaction at either 0º or 30º for proton beam axis. This approach allows the subtraction of the low energy neutron components. The measured cross sections are compared with the findings of other studies, and evaluated in relation to nuclear data files.

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.

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 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 High-power TR-24 cyclotron-based p-n convertor cooled by submerged orifice jet

2020 ◽  
Vol 231 ◽  
pp. 03005
Author(s):  
Bém Pavel ◽  
Běhal Radomír ◽  
Gӧtz Miloslav ◽  
Plíhal Petr ◽  
Poklop Dušan ◽  
...  
Keyword(s):  
Nuclear Physics ◽  
Cooling System ◽  
Forward Direction ◽  
Beam Current ◽  
Neutron Field ◽  
Physics Institute ◽  
Jet Cooling ◽  
Submerged Orifice ◽  
Target Output ◽  
Nuclear Physics Institute

The TR-24 cyclotron (Advanced Cyclotron Systems Inc., Canada) of the Nuclear Physics Institute in Řež provides protons with variable energies from 18 MeV up to 24 MeV and beam current of 0.3 mA. For such parameters, the p +Be source reaction on thick Be target can produce a white-spectrum neutron field (En ≤ 22 MeV) with the intensity of 5×10 12 n/s/sr in forward direction. Present paper outlines the development of Be-target cooling system, devoted to remove the heat load of 7 kW (density up to 4 kW/cm2) from the target. Due to novel “orifice-form“ of jet cooling (resulting in the shortest source-to- sample distance of 20 mm) with extremely high cooling efficiency, the TR-24 p-n convertor can achieve neutron-flux up to 2×1012n/cm2/s nearby the target output.

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.

Nuclear Data and the Effect of Gadolinium in the Moderator

2012 ◽  
Vol 1 (1) ◽  
pp. 21-25 ◽  
Author(s):  
J.C. Chow ◽  
F.P. Adams ◽  
D. Roubstov ◽  
R.D. Singh ◽  
M.B. Zeller
Keyword(s):  
Cross Section ◽  
Neutron Capture ◽  
Capture Cross Section ◽  
Neutron Transport ◽  
Nuclear Data ◽  
Calibration Data ◽  
Capture Cross ◽  
Neutron Capture Cross Section ◽  
Reactivity Effect ◽  
Section Data

Recent cross-section measurements on gadolinium have raised concerns over the accuracy of moderator poison reactivity coefficient calculations. Measurements have been made at the ZED-2 (Zero Energy Deuterium) critical facility, Chalk River Laboratories, AECL, to study the reactivity effect of gadolinium in the moderator. Since the neutron capture cross-section of boron is well known, measurements were also made with boron to provide calibration data for measurements with gadolinium. The measurements have been used to quantify the bias of the reactivity effect in full-core simulations of ZED-2 using MCNP, a neutron transport code used extensively for simulations of nuclear systems, along with the ENDF/B-VII.0 cross-section data. The results showed a bias of -0.41 ± 0.07 mk/ ppm, or -2.1% ± 0.3%, given a reactivity worth of -20.1 mk/ppm for gadolinium. Additional simulations also show that the gadolinium neutron capture cross-section has been over-corrected, relative to previous evaluations, in a beta version of ENDF/B VII.1, which incorporates the Leinweber data.

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