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.

U233 Data Evaluation for Criticality Study

2016 ◽  
Vol 2 (3) ◽  
Author(s):  
Mohammad Alrwashdeh ◽  
Wang Kan
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
Computer Code ◽  
Data Fitting ◽  
Multiplication Factor ◽  
Cross Section Data ◽  
Mathematical Methods ◽  
Bayes Method ◽  
Section Data ◽  
Experimental Data Fitting

The aim of this study is to investigate the available U233 cross section data for adequate calculation of critical benchmark experiments, to calculate the multiplication factor Keff for several benchmarks in both fast and thermal energy ranges. The evaluation of the U233 has been investigated using SAMMY code, in order to generate a useful database for criticality calculations; the computer code FITWR for experimental data fitting shows the same results obtained from the Bayes method included within the SAMMY code, with a slight difference in the results in the evaluated cross sections due to different mathematical methods having different results. Excellent results for the calculated Keff values are obtained for several benchmarks in the thermal and fast benchmarks considered in this study.

scholarly journals The Effect of E-Commerce on National Income in 114 Countries

2019 ◽  
pp. 148-155
Author(s):  
Rasi Tamadhika Fajar Ramadhan ◽  
Bambang Juanda ◽  
Muhammad Findi Alexandi
Keyword(s):  
Purchasing Power Parity ◽  
National Income ◽  
Least Square Method ◽  
Positive Influence ◽  
Error Component ◽  
Least Square ◽  
Information Communication ◽  
Power Parity ◽  
Economic Activities ◽  
Section Data

Information, Communication and Technology (ICT) was growing rapidly. E-commerce creates new economic activities through technology, especially in the field of trade. The main objective of the studied was to analyze the effect of e-commerce on national income. The analysis was performed by EC2SLS (Error Component-Two Stage Least Square) method. Analysis using panel data with cross section data of 114 countries in the 2016-2017 period. The data used were GDP data based on purchasing power parity (PPP); previous year's GDP; e-commerce users; ICT services exports and imports; government spending; investment (FDI); country dummy. The analysis shows that e-commerce had a significant effect on national income. E-commerce for 113 other countries had a positive influence of 0.0026% while e-commerce for indonesia had a positive influence of 0.0080%.

scholarly journals The Neutron Facility at NCSR "Demokritos"- Implementation in the Case of the 232Th(n,2n) and 241Am(n,2n) Reactions

2020 ◽  
Vol 13 ◽  
pp. 136
Author(s):  
R. Vlastou ◽  
C. T. Papadopoulos ◽  
G. Perdikakis ◽  
M. Kokkoris ◽  
S. Kossionides ◽  
...  
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
Computer Code ◽  
Reaction Cross ◽  
Cross Section Data ◽  
Model Calculations ◽  
Section Data ◽  
Neutron Facility ◽  
Mev Protons ◽  
Reaction Cross Sections

In the 5.5 MV tandem T11/25 Accelerator Laboratory of NCSR "Demokritos" monoenergetic neutron beams can be produced in the energy ranges 120-650 keV, 4-11.5 MeV and 16-20.5 MeV by using the 7Li(p,n), 2H(d,n) and 3H(d,n) reactions, respectively. The corresponding beam energies and ions delivered by the accelerator, are 1.92-2.37 MeV protons, 0.8-9.6 MeV deuterons and 0.8-3.7 MeV deuterons, for the three reactions, respectively. Experimental results for neutron energies from threshold up to 11.5 MeV and at 17.1 MeV will be given for the 232Th(n,2n)231Th reaction, while for the 241 Am(n,2n)240 Am reaction, preliminary cross section data at 10.4, 10.6 and 17.1 MeV will be discussed. In the framework of the CERN n-TOF collaboration, the cross section of these reactions have been measured relative to the 197Au(n,2n)196Au, 27Al(n,a)24Na and 93Nb(n,2n) reaction cross sections, by using the activation method. In addition to the experimental work, theoretical Statistical model calculations are being carried out using the computer code STAPRE/F. The results are compared to the experimental data.

Application of a Step-Wise Verification and Validation Procedure to the 3D Neutron Kinetics Code DYN3D Within the European NURESIM Project

2009 ◽  
Author(s):  
U. Rohde ◽  
S. Mittag ◽  
U. Grundmann ◽  
P. Petkov ◽  
J. Ha´dek
Keyword(s):  
Cross Sections ◽  
Power Distribution ◽  
Verification And Validation ◽  
Benchmark Problems ◽  
Mathematical Methods ◽  
Neutron Kinetics ◽  
Operational Conditions ◽  
Section Data ◽  
Nodal Method ◽  
Mathematical Problems

A generic strategy of core physics codes benchmarking was elaborated within the European NURESIM code platform development. In this paper, the application of this step-wise procedure to benchmarking the 3D neutron kinetics code DYN3D for applications to VVER-type reactors is described. Numerical and experimental benchmark problems were considered for code verification and validation. Examples of these benchmarks including benchmark set-up and results obtained by use of DYN3D in comparison with other codes are given. First, mathematical problems with given cross sections are used for the verification of the mathematical methods applied e.g. in nodal codes against finite difference solutions. Discretization errors were quantified. After minimisation of numerical errors, modelling errors have to be considered. Diffusion approximation and homogenisation errors are due to simplified physical approaches and can be estimated by comparing diffusion solutions with more accurate Monte Carlo or deterministic transport solutions. Methods to reduce these errors are outlined. A series of 2D whole core benchmarks for different core loadings and operational conditions for VVER-1000 reactors was defined for this purpose. Reference transport solutions were calculated by the MARIKO and APOLLO codes based on Method of Characteristics. Homogenised two-group and few-group diffusion parameters were derived from the reference solutions and used as cross section data for the nodal diffusion code DYN3D. The DYN3D solutions were compared to the reference solution. It was shown that the homogenisation error can be significantly reduced by using Assembly Discontinuity Factors (ADF) and Reflector Discontinuity Factors (RDF) which are obtained from the transport solution by applying equivalence theory. A study using the multi-group version of DYN3D has shown that increasing the number of groups in the considered cases has only a small effect in comparison with homogenisation error. After reducing modelling errors by choosing appropriate physical approximations, the code have to be validated against reality. Experimental problems are used for code validation. Experimental data for VVER reactors, which were used for the benchmarking of the DYN3D code within NURESIM, are power distribution measurements at the full-size (VVER-1000) experimental facility V-1000, which have been well documented within the EC project VALCO, and kinetic experiments at the LR-0 zero power reactor in NRI Rˇezˇ. The code DYN3D, being one of the NURESIM platform codes, has proved to be an effective tool for steady-state and kinetics core calculations. The high accuracy of the code is based on the advanced nodal method “HEXNEM2”, multi-group approach, applying discontinuity factors, and intra-nodal flux reconstruction.

scholarly journals Covariance Matrices for Differential and Angle-Integrated Neutron-Induced Elastic and Inelastic Scattering Cross Sections of 56Fe

2019 ◽  
Vol 211 ◽  
pp. 07002 ◽  
Author(s):  
Shengli Chen ◽  
David Bernard ◽  
Pascal Archier ◽  
Cyrille De Saint Jean ◽  
Gilles Noguere ◽  
...  
Keyword(s):  
Cross Sections ◽  
Uncertainty Propagation ◽  
Least Square Method ◽  
Nuclear Data ◽  
High Energy ◽  
Least Square ◽  
Angular Distributions ◽  
Model Parameters ◽  
Nuclear Data Library ◽  
Scattering Cross Sections

Correlations between neutron inelastic scatterings angular distributions are not included in the Joint Evaluated Fission and Fusion (JEFF) nuclear data library, while they are key quantities for uncertainty propagation of nuclear data. By reproducing the angle-integrated cross sections and uncertainties of JEFF-3.1.1, the present work obtains covariance matrix between high energy model parameters using the least square method implemented in the CONRAD code. With this matrix, it is possible to generate correlations between angle-integrated cross sections and angular distributions, which are usually presented by Legendre coefficients. As expected, strong correlations are found, for example, between the Legendre coefficients of elastic and first-level-inelastic scatterings and the angle-integrated total, elastic, total inelastic cross sections.

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 The Reliability Evaluation of Antiskid Valve of CRH2 EMUs Based on Bayes Method

2018 ◽  
Vol 47 (4) ◽  
pp. 268-276
Author(s):  
Jianwei Yang ◽  
Xinghui Qiu ◽  
Zhan Chen
Keyword(s):  
Weibull Distribution ◽  
Theoretical Basis ◽  
Least Square Method ◽  
Least Square ◽  
The Other ◽  
Rail Transit ◽  
Bayes Method ◽  
Braking System ◽  
The Mathematical Model ◽  
Two Parameter

This paper takes the antiskid valve of braking system as the object of the research. Two parameter Weibull distribution is established through analyzing the operational principle of antiskid valve. According to the Bayesian Reliability, the cumulative failure probability of antiskid valve is calculated. Meanwhile, the posterior distribution is obtained through Markov Chain Monte Carlo. By contrast with these two methods that the reliability of antiskid valve is obtained. According to the least square method, the undetermined parameters of the two parameter Weibull distribution were solved. Then, through the sensitivity analysis, the mathematical model is proved more robust. This paper provides a certain theoretical basis for the assessment of antiskid valve’s reliability, and it provides the reference to the other valve in braking system of rail transit vehicle. At the end of the paper, the reliability of the whole antiskid system is calculated through GO method on the basis of the parts' reliability is obtained by Bayes method.

scholarly journals Use of integral data assimilation and differential measurements as a contribution to improve 235U and 238U cross sections evaluations in the fast and epithermal energy range

2018 ◽  
Vol 4 ◽  
pp. 41
Author(s):  
Virginie Huy ◽  
Gilles Noguère ◽  
Gérald Rimpault
Keyword(s):  
Cross Sections ◽  
Critical Mass ◽  
A Priori ◽  
Least Square Method ◽  
Nuclear Data ◽  
Least Square ◽  
Specific Information ◽  
Experimental Database ◽  
Integral Experiment ◽  
Nuclear Data Library

Critical mass calculations of various HEU-fueled fast reactors result in large discrepancies in C/E values, depending on the nuclear data library used and the configuration modeled. Thus, it seems relevant to use integral experiments to try to reassess cross sections that might be responsible for such a dispersion in critical mass results. This work makes use of the Generalized Least Square method to solve Bayes equation, as implemented in the CONRAD code. Experimental database used includes ICSBEP Uranium based critical experiments and benefits from recent re-analyses of MASURCA and FCA-IX criticality experiments (with Monte-Carlo calculations) and of PROFIL irradiation experiments. These last ones provide very specific information on 235U and 238U capture cross sections. Due to high experimental uncertainties associated to fission spectra, we chose to consider either fitting these data or set them to JEFF-3.1.1 evaluations. The work focused on JEFF-3.1.1 235U and 238U evaluations and results presented in this paper for 235U capture and 238U capture, and inelastic cross sections are compared to recent differential experiment or recent evaluations. Our integral experiment assimilation work notably suggests a 30% decrease for 235U capture around 1–2.25 keV, a 10% increase in the unresolved resonance range when using JEFF-3.1.1 as “a priori” data. These results are in agreement with recent microscopic measurements from Danon et al. [Nucl. Sci. Eng. 187, 291 (2017)] and Jandel et al. [Phys. Rev. Lett. 109, 202506 (2012)]. For 238U cross sections, results are highly dependent on fission spectra.

Direct Assessment of Effective Renal Plasma Flow from Renoscintigraphy with a Gamma Camera and an On-Line Computer

1981 ◽  
Vol 20 (06) ◽  
pp. 274-278
Author(s):  
J. Liniecki ◽  
J. Bialobrzeski ◽  
Ewa Mlodkowska ◽  
M. J. Surma
Keyword(s):  
Count Rate ◽  
Renal Plasma Flow ◽  
Early Period ◽  
Least Square Method ◽  
Least Square ◽  
Uptake Coefficient ◽  
Positive Linear Correlation ◽  
On Line ◽  
Best Fit ◽  
Plasma Activity

A concept of a kidney uptake coefficient (UC) of 131I-o-hippurate was developed by analogy from the corresponding kidney clearance of blood plasma in the early period after injection of the hippurate. The UC for each kidney was defined as the count-rate over its ROI at a time shorter than the peak in the renoscintigraphic curve divided by the integral of the count-rate curve over the "blood"-ROI. A procedure for normalization of both curves against each other was also developed. The total kidney clearance of the hippurate was determined from the function of plasma activity concentration vs. time after a single injection; the determinations were made at 5, 10, 15, 20, 30, 45, 60, 75 and 90 min after intravenous administration of 131I-o-hippurate and the best-fit curve was obtained by means of the least-square method. When the UC was related to the absolute value of the clearance a positive linear correlation was found (r = 0.922, ρ > 0.99). Using this regression equation the clearance could be estimated in reverse from the uptake coefficient calculated solely on the basis of the renoscintigraphic curves without blood sampling. The errors of the estimate are compatible with the requirement of a fast appraisal of renal function for purposes of clinical diagknosis.

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