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 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.

scholarly journals Correlated energy uncertainties in reaction rate calculations

2020 ◽  
Vol 642 ◽  
pp. A41
Author(s):  
Richard Longland ◽  
Nicolas de Séréville
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Reaction Cross Section ◽  
Reaction Rate ◽  
Resonance Energy ◽  
Reaction Rates ◽  
Reaction Cross ◽  
Current Interest ◽  
Resonance Energies ◽  
The Impact

Context. Monte Carlo methods can be used to evaluate the uncertainty of a reaction rate that arises from many uncertain nuclear inputs. However, until now no attempt has been made to find the effect of correlated energy uncertainties in input resonance parameters. Aims. Our goal is to investigate the impact of correlated resonance energy uncertainties on reaction rates. Methods. Using a combination of numerical and Monte Carlo variation of resonance energies, the effect of correlations are investigated. Five reactions are considered: two fictional, illustrative cases and three reactions whose rates are of current interest. Results. The effect of correlations in resonance energies depends on the specific reaction cross section and temperatures considered. When several resonances contribute equally to a reaction rate, and when they are located on either side of the Gamow peak, correlations between their energies dilute their effect on reaction rate uncertainties. If they are both located above or below the maximum of the Gamow peak, however, correlations between their resonance energies can increase the reaction rate uncertainties. This effect can be hard to predict for complex reactions with wide and narrow resonances contributing to the reaction rate.

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.

Methodology of 63Cu(n,2n)62Cu Reaction Rate Measurement

2020 ◽  
Vol 7 (2) ◽  
Author(s):  
Martin Schulc ◽  
Michal Košťál ◽  
Jan Šimon ◽  
Evžen Novák
Keyword(s):  
Spontaneous Fission ◽  
Reaction Rate ◽  
Gamma Spectroscopy ◽  
Nuclear Data ◽  
Fission Neutron ◽  
Neutron Field ◽  
Gamma Line ◽  
Annihilation Peak ◽  
Standard Neutron ◽  
Data Libraries

Abstract This paper presents the measurement of the spectrum-averaged cross section (SACS) of 63Cu(n,2n)62Cu reaction in 252Cf spontaneous fission neutron spectrum. The SACS in the 252Cf spectrum was chosen as a validation tool since 252Cf is the only standard neutron field and 62Cu isotope is not easy to measure by gamma spectroscopy since the gamma line of interest is an annihilation peak, which is also produced by 64Cu isotope. Fortunately, contributions to the annihilation peak from these isotopes can be distinguished due to the very different half-lives. SACS was inferred from the experimental reaction rate. The SACS in the 252Cf spontaneous fission neutron field for the 63Cu(n,2n)62Cu reaction was determined as equal to (0.1763 ± 0.0077 mb). This value agrees with value (0.183 ± 0.007) × 10−3 b within uncertainty presented by W. Mannhart. However, it differs by 12.7% from IRDFF-II value, which is equal to (0.19874 ± 8.954 10−3) × 10−3 b. Furthermore, reasonable agreement is not achieved with ENDF/B-VIII.0, JEFF-3.3, CENDL-3.1, ROSFOND-2010, nor JENDL-4.0 nuclear data libraries.

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 Preliminary results for the 19F(ρ,α)16O reaction cross section measured at INFN-LNS

2020 ◽  
Vol 227 ◽  
pp. 02009
Author(s):  
T Petruse ◽  
G. L. Guardo ◽  
M. La Cognata ◽  
D. Lattuada ◽  
C. Spitalieri ◽  
...  
Keyword(s):  
Nuclear Reaction ◽  
Cross Section ◽  
Theoretical Models ◽  
Calibration Procedure ◽  
Reaction Rates ◽  
Reaction Cross ◽  
Asymptotic Giant Branch ◽  
Angular Distributions ◽  
Agb Stars ◽  
Indirect Measurements

The 19F(ρ,α)16O reaction is an important fluorine destruction chan- nel in the proton-rich outer layers of asymptotic giant branch (AGB) stars and it might also play a role in hydrogen-deficient post-AGB star nucleosynthesis. At present, theoretical models overproduce F abundances in AGB stars with re-spect to the observed values, thus calling for further investigation of the nuclear reaction rates involved in the production and destruction of fluorine. In the last years, new direct and indirect measurements improved significantly the knowl- edge of 19F(ρ,α)16O cross section at deeply sub-Coulomb energies (below 0.8 MeV). However, those data are larger by a factor of 1.4 with respect the previ- ous data reported in the NACRE compilation in the energy region 0.6-0.8 MeV. Using the Large High resolution Array of Silicons for Astrophysics (LHASA), we performed a new direct measurement of the 19F(ρ,α)16O. The goal of this experiment is to reduce the uncertainties in the nuclear reaction rate of the 19F(ρ,α)16O reaction. Here, experimental details, the calibration procedure and angular distributions are presented.

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.

scholarly journals 14.8 MeV Neutron Activation Cross Section Measurements for Ge Isotopes

2009 ◽  
Vol 1 (2) ◽  
pp. 173-181 ◽  
Author(s):  
M. M. Haque ◽  
M. T. Islam ◽  
M. A. Hafiz ◽  
R. U. Miah ◽  
M. S. Uddin
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
Gamma Ray ◽  
Reaction Cross ◽  
Activation Cross Section ◽  
Section Data ◽  
Gamma Ray Spectroscopy ◽  
24Na Reaction ◽  
The Cross ◽  
Good Agreement

The cross sections of Ge isotopes were measured with the activation method at 14.8 MeV neutron energy. The quasi-monoenergetic neutron beams were produced via the 3H(d,n)4He reaction at the 150 kV J-25 neutron generator of INST, AERE. The characteristics γ-lines of the product nuclei were measured with a closed end coaxial 17.5 cm2 high purity germanium (HPGe) detector gamma ray spectroscopy. The cross sections were determined with reference to the known 27Al(n,α)24Na reaction. Cross section data are presented for 72Ge(n,p)72Ga, 74Ge(n,α)71mZn and 76Ge(n,2n)75m+gGe reactions. The cross section values obtained for the above reactions were 24.78±1.75 mb, 1.69±0.11 mb and 860±50 mb, respectively. The results obtained were compared with the values reported in literature as well as theoretical calculation performed by the statistical code SINCROS-II. The experimental data were found fairly in good agreement with the calculated and literature data.  Keywords: Activation cross section; Neutron induced reaction; Gamma-ray spectroscopy; 14.8 MeV. © 2009 JSR Publications. ISSN: 2070-0237 (Print); 2070-0245 (Online). All rights reserved. DOI: 10.3329/jsr.v1i2.1532  

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.

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