Geant4 is a software toolkit for the simulation of the passage of particles through matter. It is used by a large number of experiments and projects in a variety of application domains, including high energy physics, astrophysics and space science, medical physics and radiation protection. Over the past several years, major changes have been made to the toolkit in order to accommodate the needs of these user communities, and to efficiently exploit the growth of computing power made available by advances in technology. The adaptation of Geant4 to multithreading, advances in physics, detector modeling and visualization, extensions to the toolkit, including biasing and reverse Monte Carlo, and tools for physics and release validation are discussed here.

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Recent developments in GEANT4

ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

https://digital.library.unt.edu/ark:/67531/metadc888955/m2/1/high_res_d/900147.pdf

ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology

The fourth version of the Japanese Evaluated Nuclear Data Library has been produced in cooperation with the Japanese Nuclear Data Committee. In the new library, much emphasis is placed on the impro...

https://www.tandfonline.com/doi/pdf/10.1080/18811248.2011.9711675

JENDL-4.0: A New Library for Nuclear Science and Engineering

https://escholarship.org/content/qt2dj9p3h4/qt2dj9p3h4.pdf?t=pmi3bw

ENDF/B-VIII.0: The 8th Major Release of the Nuclear Reaction Data Library with CIELO-project Cross Sections, New Standards and Thermal Scattering Data

Neutron resonance parameters of 232Th were obtained from the Reich-Moore SAMMY analysis of high-resolution neutron transmission measurements performed at the Oak Ridge Electron Linear Accelerator (ORELA) by Olsen in 1981, along with the high-resolution neutron capture measurements performed in 2005 at the Geel Linear Accelerator (GELINA, Belgium) by Schillebeeckx and at the n-TOF facility (CERN, Switzerland) by Aerts. The ORELA data were analyzed previously by Olsen with the Breit-Wigner multilevel code SIOB, and the results were used in the ENDF/B-VI evaluation. In the new analysis of the Olsen neutron transmissions by the modern computer code SAMMY, better accuracy is obtained for the resonance parameters by including in the experimental data base the recent experimental neutron capture data. The experimental data base and the method of analysis are described in the report. The neutron transmissions and the capture cross sections calculated with the resonance parameters are compared to the experimental values. A description is given of the statistical properties of the resonance parameters. The new evaluation results in a decrease in the capture resonance integral and improves the prediction of integral thermal benchmarks.

R-Matrix Analysis of 232Th Neutron Transmissions and Capture Cross Sections in the Energy Range Thermal to 4 keV

We have evaluated 16O neutron cross sections in the resolved resonance region with the multilevel Reich-Moore code SAMMY. Resonance parameters were determined by a consistent analysis, including both Doppler and resolution broadening effects. To properly treat the α particle exit channel, an algorithm to calculate charged particle penetrabilities and shifts was incorporated into SAMMY.

R-Matrix Evaluation of 16O Neutron Cross Sections up to 6.3 MeV

For many Monte Carlo codes cross sections are generally only created at a set of predetermined temperatures. This causes an increase in error as one moves further and further away from these temperatures in the Monte Carlo model. This paper discusses recent progress in the Scale Monte Carlo module KENO to create problem dependent, Doppler broadened, cross sections. Currently only broadening the 1D cross sections and probability tables is addressed. The approach uses a finite difference method to calculate the temperature dependent cross-sections for the 1D data, and a simple linear-logarithmic interpolation in the square root of temperature for the probability tables. Work is also ongoing to address broadening theS (alpha , beta) tables. With the current approach the temperature dependent cross sections are Doppler broadened before transport starts, and, for all but a few isotopes, the impact on cross section loading is negligible. Results can be compared with those obtained by using multigroup libraries, as KENO currently does interpolation on the multigroup cross sections to determine temperature dependent cross-sections. Current results compare favorably with these expected results.

Problem dependent doppler broadening of continuous energy cross sections in the keno monte carlo computer code

The error estimation for calculated quantities relies on nuclear data uncertainty information available in the basic nuclear data libraries such as the U.S. Evaluated Nuclear Data File (ENDF/B). The uncertainty files (covariance matrices) in the ENDF/B library are generally obtained from analysis of experimental data. In the resonance region, the computer code SAMMY is used for analyses of experimental data and generation of resonance parameters. In addition to resonance parameters evaluation, SAMMY also generates resonance parameter covariance matrices (RPCM). SAMMY uses the generalized least-squares formalism (Bayes’ method) together with the resonance formalism (R-matrix theory) for analysis of experimental data. Two approaches are available for creation of resonance-parameter covariance data. (1) During the data-evaluation process, SAMMY generates both a set of resonance parameters that fit the experimental data and the associated resonance-parameter covariance matrix. (2) For existing resonance-parameter evaluations for which no resonance-parameter covariance data are available, SAMMY can retroactively create a resonance-parameter covariance matrix. The retroactive method was used to generate covariance data for 235 U. The resulting 235 U covariance matrix was then used as input to the PUFF-IV code, which processed the covariance data into multigroup form, and to the TSUNAMI code, which calculated the uncertainty in the multiplication factor due to uncertainty in the experimental cross sections. The objective of this work is to demonstrate the use of the 235 U

Impact of the 235U Covariance Data in Benchmark Calculations

The Oak Ridge Electron Linear Accelerator (ORELA) was used to measure neutron total and capture cross sections of aluminum, silicon, chlorine, fluorine, and potassium in the energy range from 100 eV to ~600 keV. These measurements were carried out to support the Nuclear Criticality Safety Program (NCSP). Concerns about the use of existing cross section data in nuclear criticality calculations have been a prime motivator for the new cross section measurements. Our results are substantially different from the evaluated nuclear data files of ENDF/B-VI and JENDL-3.2.

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Luiz C Leal

Papers

R-Matrix Analysis of 232Th Neutron Transmissions and Capture Cross Sections in the Energy Range Thermal to 4 keV

R-Matrix Evaluation of 16O Neutron Cross Sections up to 6.3 MeV

Problem dependent doppler broadening of continuous energy cross sections in the keno monte carlo computer code

Impact of the 235U Covariance Data in Benchmark Calculations

New Neutron Cross Section Measurements at ORELA for Improved Nuclear Data