scholarly journals VERIFICATION OF THE OpenMC HOMOGENIZED MYRRHA-1.6 CORE MODEL

2021 ◽  
Vol 247 ◽  
pp. 04002
Author(s):  
Augusto Hernandez-Solis ◽  
Yohannes Molla ◽  
Edoaurd Malambu ◽  
Alexey Stankovskiy ◽  
Gert Van den Eynde
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Cross Sections ◽  
Deterministic Model ◽  
Reaction Rates ◽  
Relative Difference ◽  
Reactor Physics ◽  
Heterogeneous Model ◽  
Macroscopic Cross Section ◽  
Nodal Model

The OpenMC code is being employed both as a multi-group nodal macroscopic cross-section generator and a reference multi-group Monte Carlo (MGMC) solution. The aim is to do a neutronic benchmark verification study versus a deterministic model (based on the MYRRHA-1.6 core) performed by the PHISICS simulator. MYRRHA, a novel research accelerator driven system concept that is also foreseen to work as a critical configuration, offers a rich opportunity of testing state-of-the art methods for reactor physics analysis due to its strong heterogeneous configuration utilized for both thermal and fast spectra irradiation purposes. The original core configuration representing MYRRHA-1.6 and formed by 169 assemblies, was launched in OpenMC for producing a homogenous nodal model that, when executed in its multi-group Monte Carlo mode, it produced a keff that differs in almost 500 pcm from the original case. This means that in the future, such approximation should correct the nodal cross-sections to preserve the reaction rates in order to match those ones from the heterogeneous model. Nevertheless, such MGMC mode of operation offered by OpenMC could be exploited in order to verify deterministic core simulators. By inputting the same nodal multi-group cross-section model into the transport solver of the PHISICS toolkit, the neutronic benchmark showed a difference of 171 pcm in eigenvalue while comparing it to its OpenMC MGMC counterpart. Also, local multi-group and energy-integrated nodal profiles of the neutron flux showed a maximum relative difference between methodologies of 15% and 1%, respectively. This means that the MGMC capabilities offered by OpenMC can be employed to verify other deterministic methodologies.

The Super Equivalence Method in Monte Carlo Based Homogenization

2014 ◽  
Author(s):  
Mancang Li ◽  
Kan Wang ◽  
Dong Yao
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Reactor Core ◽  
Surface Current ◽  
Reaction Rates ◽  
Reactor Physics ◽  
Sph Method ◽  
Equivalence Theory ◽  
Spe Method ◽  
Improved Accuracy

The general equivalence theory (GET) and the superhomogenization method (SPH) are widely used for equivalence in the standard two-step reactor physics calculation. GET has behaved well in light water reactor calculation via nodal reactor analysis methods. The SPH was brought up again lately to satisfy the need of accurate pin-by-pin core calculations. However, both of the classical methods have their limitations. The super equivalence method (SPE) is proposed in the paper as an attempt to preserve the surface current, the reaction rates and the reactivity. It enhances the good property of the SPH method through reaction rates based normalization. The concept of pin discontinuity factors are utilized to preserve the surface current, which is the basic idea in the GET technique. However, the pin discontinuity factors are merged into the homogenized cross sections and diffusion coefficients, thus no additional homogenization parameters are needed in the succedent reactor core calculation. The eigenvalue preservation is performed after the reaction rate and surface current have been preserved, resulting in reduced errors of reactivity. The SPE has been implemented into the Monte Carlo method based homogenization code MCMC, as part of RMC Program, under developed in Tsinghua University. The C5G7 benchmark problem have been carried out to test the SPE. The results show that the SPE method not only suits for the equivalence in Monte Carlo based homogenization but also provides improved accuracy compared to the traditional GET or SPH method.

scholarly journals IMPROVEMENT OF SCALE-XSProc MULTIGROUP CROSS SECTION PROCESSING BASED ON THE CENTRM POINTWISE SLOWING DOWN CALCULATION

2021 ◽  
Vol 247 ◽  
pp. 02011
Author(s):  
Seog Kim Kang ◽  
Andrew M. Holcomb ◽  
Friederike Bostelmann ◽  
Dorothea Wiarda ◽  
William Wieselquist
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Cross Sections ◽  
Reaction Rate ◽  
Neutron Transport ◽  
Energy Structure ◽  
Reactor Physics ◽  
Slowing Down ◽  
Monte Carlo Calculations ◽  
Scattering Matrices

The SCALE-XSProc multigroup (MG) cross section processing procedure based on the CENTRM pointwise slowing down calculation is the primary procedure to process problem-dependent self-shielded MG cross sections and scattering matrices for neutron transport calculations. This procedure supports various cell-based geometries including slab, 1-D cylindrical, 1-D spherical and 2-D rectangular configurations and doubly heterogeneous particulate fuels. Recently, this procedure has been significantly improved to be applied to any advanced reactor analysis covering thermal and fast reactor systems, and to be comparable to continuous energy (CE) Monte Carlo calculations. Some reactivity bias and reaction rate differences have been observed compared with CE Monte Carlo calculations, and several areas for improvement have been identified in the SCALE-XSProc MG cross section processing: (1) resonance self-shielding calculations within the unresolved resonance range, (2) 10 eV thermal cut-off energy for the free gas model, (3) on-the-fly adjustments to the thermal scattering matrix, (4) normalization of the pointwise neutron flux, and (5) fine MG energy structure. This procedure ensures very accurate MG cross section processing for high-fidelity deterministic reactor physics analysis for various advanced reactor systems.

scholarly journals Production of radioisotopes within a plasma focus device

2005 ◽  
Vol 20 (1) ◽  
pp. 33-37 ◽  
Author(s):  
Ergisto Angeli ◽  
Agostino Tartari ◽  
Michele Frignani ◽  
Vincenzo Molinari ◽  
Domiziano Mostacci ◽  
...  
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
Plasma Focus ◽  
Reaction Rates ◽  
Experimental Production ◽  
Endogenous Production ◽  
The Us ◽  
Further Development ◽  
Radio Isotopes ◽  
Very High

In recent years, research conducted in the US and in Italy has demonstrated production of radioisotopes in plasma focus devices, and particularly, on what could be termed "endogenous" production, to wit, production within the plasma it self, as opposed to irradiation of tar gets. This technique relies on the formation of localized small plasma zones characterized by very high densities and fairly high temperatures. The conditions prevailing in these zones lead to high nuclear reaction rates, as pointed out in previous work by several authors. Further investigation of the cross sections involved has proven necessary to model the phenomena involved. In this paper, the present status of research in this field is re viewed, both with regards to cross section models and to experimental production of radio isotopes. Possible out comes and further development are discussed.

The Generation of Few-Group Constants for Fast Reactor Analysis

2016 ◽  
Author(s):  
Xianan Du ◽  
Liangzhi Cao ◽  
Youqi Zheng
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Cross Sections ◽  
Fast Reactor ◽  
Monte Carlo Calculation ◽  
Anisotropic Scattering ◽  
Slowing Down ◽  
Nodal Method ◽  
Reactor Calculation ◽  
Broad Energy

A way to generate the few-group cross sections for fast reactor calculation is presented in this paper. It is based on the three steps computational scheme. In the first step, the ultrafine method is used to solve the slowing down equation based on the ultrafine group cross section generated by NJOY. Optional 0D or 1D calculation is used to collapse energy group into broad energy groups. In the second step, the 2D RZ calculation using SN method is performed to obtain the space dependent neutron spectra to collapse broad energy groups into few groups. The anisotropic scattering is well handled by the direct SN calculation. Finally, the full core calculation is performed by using the 3D SN nodal method. The results are compared with continuous energy Monte-Carlo calculation. Both the cross section generated in the first step and the final keff in the last step are compared. The results match well between the three steps calculation and Monte-Carlo calculation.

STOPPING CROSS SECTIONS IN CARBON FOR LOW-ENERGY ATOMS WITH

1963 ◽  
Vol 41 (9) ◽  
pp. 1424-1442 ◽  
Author(s):  
J. H. Ormrod ◽  
H. E. Duckworth
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Atomic Number ◽  
Cross Sections ◽  
Monte Carlo Calculation ◽  
Low Energy ◽  
Energy Interval ◽  
Nuclear Scattering

The electronic stopping cross sections in carbon for atomic projectiles with [Formula: see text] have been determined in the energy interval from 10 to 140 kev. In doing so a Monte Carlo calculation was used to subtract from each experimentally observed cross section the contribution which arises from nuclear scattering. The trend of the results thus obtained agrees well with theory. In addition, however, a periodic dependence of Sε on the atomic number of the projectile is observed.

scholarly journals A systematic study of proton capture reactions in the Se-Sn region relevant to p-process nucleosynthesis

2019 ◽  
Vol 11 ◽  
Author(s):  
S. Harissopulos ◽  
P. Demetriou ◽  
S. Galanopoulos ◽  
G. Kriembardis ◽  
M. Kokkoris ◽  
...  
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
Nuclear Physics ◽  
Nuclear Reactions ◽  
Reaction Network ◽  
Reaction Rates ◽  
Compton Suppression ◽  
Experimental Information ◽  
Systematic Investigation ◽  
Model Calculations

The synthesis of the so-called ρ nuclei, i.e. a certain class of proton rich nuclei that are heavier than iron, requires a special mechanism known as ρ process. This process consists of various nucleosynthetic scenaria. In some of them proton and alpha-capture reactions are strongly involved, p-process nucleosynthesis is assumed to occur in the Oxygen/Neon rich layers of type II supernovae during their explosion, ρ nuclei are typically 10-100 times less abundant than the corresponding more neutron-rich isotopes. The prediction of their abundances is one of the major puzzles of all models of p-process nucleosynthesis. Until now all these models are capable of reproducing these abundances within a factor of 3. However, they all fail in the case of the light ρ nuclei with A<100. The observed discrepancies could be attributed to uncertainties in the pure "astrophysical" part of the p-process modelling. However, they could also be the result of uncertainties in the nuclear physics data entering the corresponding abundance calculations. In order to perform these calculations the cross sections of typically 10000 nuclear reactions of an extended reaction network involving almost 1000 nuclei from A=12 to 210 are used as input data. Such a huge amount of experimental cross section data are not available. Hence, all extended network calculations rely almost completely on cross sections predicted by the Hauser-Feshbach (HF) theory. It is therefore of paramount importance, on top of any astrophysical model improvements, to test also the reliability of the HF calculations, i.e. to investigate the uncertainties associated with the evaluation of the nuclear properties, like nuclear level densities and nucleon-nucleus potentials, entering the calculations. Until now, this check has been hindered significantly by the fact that in the Se-Sn region there has been scarce experimental information on cross sections at astrophysically relevant energies. In the present work, a systematic investigation of (p,7) cross sections of nuclei from Se to Sb is presented for the first time. The in-beam cross section measurements reported were carried out at energies relevant to p-process nucleosynthesis, i.e. from 1.4 to 5 MeV. The experiments were performed by using either an array of 4 HPGe detectors of 100% relative efficiency shielded with BGO crustals for Compton suppression, or a 4π Nal summing detector. The resulting cross sections, astrophysical S-factors and reaction rates of more than 10 nuclear reactions are compared with the predictions of various statistical model calculations.

scholarly journals Comprehensive Second-Order Adjoint Sensitivity Analysis Methodology (2nd-ASAM) Applied to a Subcritical Experimental Reactor Physics Benchmark: V. Computation of Mixed 2nd-Order Sensitivities Involving Isotopic Number Densities

Energies ◽  
2020 ◽  
Vol 13 (10) ◽  
pp. 2580 ◽  
Author(s):  
Ruixian Fang ◽  
Dan G. Cacuci
Keyword(s):  
Sensitivity Analysis ◽  
Cross Section ◽  
Cross Sections ◽  
Second Order ◽  
Adjoint Sensitivity Analysis ◽  
Number Density ◽  
Reactor Physics ◽  
Adjoint Sensitivity ◽  
Energy Group ◽  
Analysis Methodology

This work applies the Second-Order Adjoint Sensitivity Analysis Methodology (2nd-ASAM) to compute the mixed 2nd-order sensitivities of a polyethylene-reflected plutonium (PERP) benchmark’s leakage response with respect to the benchmark’s imprecisely known isotopic number densities and the other benchmark imprecisely known parameters, including: (i) the 6 × 180 mixed 2nd-order sensitivities involving the total microscopic cross sections; (ii) the 6 × 21,600 mixed 2nd-order sensitivities involving the scattering microscopic cross sections; (iii) the 6 × 60 mixed 2nd-order sensitivities involving the fission microscopic cross sections; and (iv) the 6 × 60 mixed 2nd-order sensitivities involving the average number of neutrons produced per fission. It is shown that many of these mixed 2nd-order sensitivities involving the isotopic number densities have very large values. Most of the large sensitivities involve the isotopic number density of 239Pu, and the microscopic total, scattering or fission cross sections for the 12th or 30th energy groups of 239Pu or 1H, respectively. The 2nd-order mixed sensitivity of the PERP leakage response with respect to the isotopic number density of 239Pu and the microscopic total cross section for the 30th energy group of 1H is the largest of the above mentioned sensitivities, attaining the value −94.91.

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 Ratio of spectral averaged cross sections measured in standard 252Cf(sf) and 235U(nth,f) neutron fields

2020 ◽  
Vol 239 ◽  
pp. 19004
Author(s):  
Martin Schulc ◽  
Michal Kostal ◽  
Roberto Capote ◽  
Evzen Novak ◽  
Nicola Burianova ◽  
...  
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
High Purity ◽  
Hpge Detector ◽  
High Volume ◽  
Reaction Rates ◽  
Fission Neutron ◽  
Neutron Field ◽  
High Threshold ◽  
High Purity Germanium

The results of systematic evaluations of spectrum averaged cross section (SACS) measurements in the fission neutron fields of 252Cf and 235U are presented. The data form a complete database of high-threshold experimental SACS measured in the same installation under the same conditions and using the same high purity germanium gamma spectrometer. This is crucial to reduce the uncertainty of the ratio and the data scattering and therefore, to minimize discrepancies compared to cross section measured under different conditions in different laboratories. This new dataset complements and extends earlier experimental evaluations. The total emission of the 252Cf neutron source during the experiments varied from 9.5E8 to 4.5E8 neutrons per second. The emission was derived in accordance to the data in the Certificate of Calibration involving absolute flux measurements in a manganese sulphate bath. Concerning 235U fission neutron field, the irradiations were carried out in a specifically designed core assembled in the zero power light water LR-0 reactor. This special core has a well described neutron field. After the irradiation, the low volume irradiated samples to be measured by gamma spectrometry were placed directly on the upper cap of a coaxial high purity germanium (HPGe) detector in a vertical configuration (ORTEC GEM35P4). High volume samples were homogenized and strewn into the Marinelli beaker. The HPGe detector is surrounded by the lead shielding box with a thin inner copper cladding and covered with rubber for suppression of background signal and bremsstrahlung. The experimental reaction rates were derived for irradiated samples from the Net Peak Areas (NPA) measured using the semiconductor HPGe detector. The measured reaction rates are used to derive the spectrum-averaged cross sections. Furthermore, measured reaction rates are also compared with MCNP6 calculations using various nuclear data libraries, in particular IRDFF evaluations.

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