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DOE OSTI2018

Rossi and Feynman-alpha formulas including prompt neutron decay and detector die away time constants - 25396

A simple Rossi-alpha formula and the corresponding Feynman-alpha formula have been derived that take into account two time constants; one characterizing the prompt neutron decay in the fissile source domain and the other the detector die-away time. The derivation of the formula is an extension of the classic heuristic method. A test problem is defined and simulated using the Monte Carlo MCNPX code in order to illustrate the applicability of the formulas. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2022

Neutronics Modeling of the pulsed plasma rocket reactor using rattlesnake

In the pulsed plasma rocket (PPR) reactor, neutron induced fission processes are utilized to implement a series of pulsed micro-explosions of very high power and eject plasma as a propellant. More specifically, projectiles (bullets) composed of moderated uranium are sent through the chamber of an unmoderated uranium barrel. By inducing rapid fission within the bullets, a plasma can be generated with appropriate delivery of neutrons. This study aims to assess the neutronics performance of the PPR reactor during normal operations using the MOOSE-based Rattlesnake code through the evaluation of the impact of the movement of the fuel bullet and the rotation of control drums on the criticality of the system. The Monte Carlo (MC) code Serpent 2 was employed to generate material-based cross sections for use in Rattlesnake and the reference neutronics solution. Cubit was used to generate the mesh for the Rattlesnake model. Parametric studies were conducted to evaluate the best approaches for cross section and mesh generation to ensure accurate results from Rattlesnake. As part of the verification process, the eigenvalue results of the system at various fuel bullet positions were obtained using Rattlesnake and compared with the reference solutions. The acceptable differences show that the Rattlesnake model with appropriate cross section and mesh generation procedure is a sufficiently accurate approximation of the continuous energy (CE) MC model for micro-sized reactors like the PPR reactor. Next, the impact of the rotation of the control drums on the system criticality was evaluated using the verified model. It was shown that the high-fidelity simulation using the deterministic code Rattlesnake can produce sufficiently accurate results for the evaluation of the reactor's neutronics performance in different phases of the normal operation with acceptable computational cost.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2011

Monte Carlo simulation of the neutron characteristics of VVER-1000 core using the MCU-PD program and comparison of the results with calculations by the BIPR-7A program and experimental data

The Monte Carlo method has been used to simulate the neutron transport in nuclear reactors for over fifty years. Fast progress in computer power and development of more and more robust and reliable algorithms, codes, and nuclear databases allow solving more challenging problems, including three-dimensional (3D) simulations of full-scale reactor cores. Short descriptions of a full-scale 3D model of the VVER-1000 core and algorithms and methods implemented in the MCU-PD and BIPR-7A codes and a comparison of the calculations by each program as well as a comparison with experimental data are given in this paper.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Coupled reactors analysis: New needs and advances using Monte Carlo methodology

Coupled reactors and the coupling features of large or heterogeneous core reactors can be investigated with the Avery theory that allows a physics understanding of the main features of these systems. However, the complex geometries that are often encountered in association with coupled reactors, require a detailed geometry description that can be easily provided by modern Monte Carlo (MC) codes. This implies a MC calculation of the coupling parameters defined by Avery and of the sensitivity coefficients that allow further detailed physics analysis. The results presented in this paper show that the MC code SERPENT has been successfully modifed to meet the required capabilities.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Using Green's Functions to Compute 3 Dimensional Neutron Flux Maps - 16016

The knowledge of the radionuclide content of radioactive waste is of utmost importance for safety and waste management reasons. Numerical simulations are used by EDF-DP2D to anticipate dismantling and radioactive waste management. The main numerical activation schemes include the computing of a 3 dimensional neutron flux map, obtained on the basis of a neutron propagation calculation. The codes used solve the transport equation called the Boltzmann equation without simplification. The Monte Carlo method is used to solve the Boltzmann equation. Random series of numbers are used to simulate the lives of millions of neutrons. The codes follow each neutron individually, from its birth to its disappearance by leakage, absorption, or fission, and then the neutron flux map is calculated at the nominal power rating conditions. The main drawback of the Monte Carlo codes is the large computing time needed to converge within acceptable statistical criteria. This extended time period is incompatible with the multiple fuel managements used by EDF to operate its NPPs because it is necessary to make a Monte Carlo calculation for each fuel loading pattern. The solution developed by EDF-DP2D is to use the importance factors of Green's Functions to simplify the computing of the 3 dimensional neutron flux maps. Physicists talk about Green's Functions where mathematicians talk about elementary solutions of linear equations with constant coefficients. With an activation scheme based on the Green's Functions approach the 3D neutron flux is obtained on the basis of the convolution of the normalized source contributions (i.e. normalized response matrices) with the neutrons emitted by the fuel assemblies. The response matrices to unitary sources are computed with the Monte Carlo code MCNP for different tallies of interest and the neutron sources emitted by the fuel are computed with a core code for different fuel loading patterns. The principal interest of this approach is the calculation of the response matrices to unitary sources that can be used for all power distributions, thus limiting the Monte Carlo computing time. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Lecture Notes on Sensitivity-Uncertainty Based Nuclear Criticality Safety Validation

This document is a collection of lecture notes for sensitivity-uncertainty analysis of nuclear criticality safety validation. The use of and results from MCNP and Whisper are included.

42 ENGINEERING↗

DOE OSTI2019

Synthetic Spectra of Pair-instability Supernovae in 3D

Pair-instability supernovae (PISNe) may signal the deaths of extremely massive stars in the local Universe or massive primordial stars after the end of the Cosmic Dark Ages. Hydrodynamic simulations of these explosions, performed in 1D, 2D, and 3D geometry, have revealed the strong dependence of mixing in the PISN ejecta on dimensionality. This chemical rearrangement is mainly driven by Rayleigh–Taylor instabilities that start to grow shortly after the collapse of the carbon–oxygen core. We report on the effects of such mixing on the spectroscopic evolution of PISNe by post-processing explosion profiles with the radiation diffusion-equilibrium code SNEC and the implicit Monte Carlo–discrete diffusion Monte Carlo radiation transport code SuperNu . The first 3D radiation transport calculation of a PISN explosion is presented, yielding viewing-angle-dependent synthetic spectra and light curves. Our findings show that while 2D and 3D mixing does not significantly affect the light curves of PISNe, their spectroscopic and color evolution are impacted. Strong features of intermediate-mass elements dominated by silicon, magnesium, and oxygen appear at different phases and reach different intensities depending on the extent of mixing in the silicon/oxygen interface of the PISN ejecta. In contrast, we do not find a significant dependence of PISN light curves and spectra on viewing angle. These results showcase the capabilities of SuperNu to handle 3D radiation transport and highlight the importance of modeling time series of spectra in identifying PISNe with future missions.

79 ASTRONOMY AND ASTROPHYSICS↗

DOE OSTI2012

TRIPOLI-4 criticality calculations for MOX fuelled SNEAK 7A and 7B fast critical assemblies

A prototype Generation IV fast neutron reactor is under design and development in France. The MOX fuel will be introduced into this self-generating core in order to demonstrate low net plutonium production. To support the TRIPOLI-4 Monte Carlo transport code in criticality calculations of fast reactors, the effective delayed neutron fraction {beta}eff estimation and the Probability Tables (PT) option to treat the unresolved resonance region of cross-sections are two essentials. In this study, TRIPOLI-4 calculations have been made using current nuclear data libraries JEFF-3.1.1 and ENDF/B-VII.0 to benchmark the reactor physics parameters of the MOX fuelled SNEAK 7A and 7B fast critical assemblies. TRIPOLI-4 calculated K{sub eff} and {beta}eff of the homogeneous R-Z models and the 3D multi-cell models have been validated against the measured ones. The impact of the PT option on K{sub eff} is 340 {+-} 10 pcm for SNEAK 7A core and 410 {+-} 12 pcm for 7B. Four-group spectra and energy spectral indices, f8/f5, f9/f5, and c8/f5 in the two SNEAK cores have also been calculated with the TRIPOLI-4 mesh tally. Calculated spectrum-hardening index f8/f5 is 0.0418 for SNEAK 7A and 0.0315 for 7B. From this study the SNEAK 3D models have been verified for the next revision of IRPhE (International Handbook of Evaluated Reactor Physics Benchmark Experiments). (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Development of MGMC: A proxy Multi-Group Monte Carlo Particle Transport Application

This document details the development and current state of MGMC: a proxy Multi-Group Monte Carlo transport application. The goal of this work was to develop a light-weight Monte Carlo transport solver that could be easily reconfigured to test parallelization strategies targeting advanced architectures and heterogeneous computing environments. Additionally, MGMC is able to act as a proxy-app for testing the development of algorithms and the integration of other libraries and into Monte Carlo transport codes. MGMC is currently able to run in parallel using OpenMP for CPU threads or CUDA for NVIDIA GPUs. MGMC was also used to investigate the using SYCL for portable parallelization targeting either CPU threads, vender-agnostic GPUs, and ARM chips, detailed in Section II. In the process of developing MGMC, several other C++ libraries have been created with the intent for re-use in future research projects and potential inclusion in production-level codes, detailed in Section III. The physics capabilities in MGMC are detailed in Section IV, code verification is detailed in Section V, and performance results are detailed in Section VI.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2011

Simulation Studies of Hydrogen Ion reflection from Tungsten for the Surface Production of Negative Hydrogen Ions

The production efficiency of negative ions at tungsten surface by particle reflection has been investigated. Angular distributions and energy spectra of reflected hydrogen ions from tungsten surface are calculated with a Monte Carlo simulation code ACAT. The results obtained with ACAT have indicated that angular distributions of reflected hydrogen ions show narrow distributions for low-energy incidence such as 50 eV, and energy spectra of reflected ions show sharp peaks around 90% of incident energy. These narrow angular distributions and sharp peaks are favorable for the efficient extraction of negative ions from an ion source equipped with tungsten surface as negative ionization converter. The retained hydrogen atoms in tungsten lead to the reduction in extraction efficiency due to boarded angular distributions.

43 PARTICLE ACCELERATORS↗

DOE OSTI2014

Trapping of electrons in troughs of self generated electromagnetic standing waves in a bounded plasma column

Observations and measurements are reported on electron trapping in troughs of self-generated electromagnetic standing waves in a bounded plasma column confined in a minimum-B field. The boundaries are smaller than the free space wavelength of the waves. Earlier work of researchers primarily focused upon electron localization effects induced by purely electrostatic perturbation. We demonstrate the possibility in the presence of electromagnetic standing waves generated in the bounded plasma column. The electron trapping is verified with electrostatic measurements of the plasma floating potential, electromagnetic measurements of the wave field profile, and optical intensity measurements of Argon ionic line at 488 nm. The experimental results show a reasonably good agreement with predictions of a Monte Carlo simulation code that takes into account all kinematical and dynamical effects in the plasma in the presence of bounded waves and external fields.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2015

A comparison of Monte Carlo generators

A comparison of GENIE, NEUT, NUANCE, and NuWro Monte Carlo neutrino event generators is presented using a set of four observables: protons multiplicity, total visible energy, most energetic proton momentum, and π{sup +} two-dimensional energy vs cosine distribution.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2025

COG11.3 Abstract

COG is a high-resolution code for the Monte Carlo simulation of coupled particle transport in arbitrary3-D geometry. COG will transport neutrons, protons, deuterons, alpha particles with energies up to hundreds of GeV, and photons with energy ranges limited by the available cross section sets and physics models. Electrons can be transported via the EGS5 electron transport kernel, electrons can also be transported. The COG code is a significant upgrade from earlier Monte Carlo transport codes and has been written specifically to makes it more versatile, accurate, and easy to use. COG has provisions for calculating deep penetration (shielding) problems, criticality problems, and neutron activation problems.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2020

S/U Comparison Study with a Focus on USLs

Under a DOE Nuclear Criticality Safety Program (NCSP) task involving Analytical Methods, three Laboratories collaborated in a comparison of results obtained from Sensitivity/Uncertainty (S/U) packages relevant to validation of transport codes. The task involves Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Los Alamos National Laboratory (LANL), and Oak Ridge National Laboratory (ORNL) comparing results of MORET 5/MACSENS V3.0, MCNP6.2/Whisper-1.1, and SCALE 6.2.3/TSUNAMI/USLSTATS respectively. All Monte Carlo transport code results utilize ENDF/B-VII.1. Four cases from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) were selected as application models: HEU-MET-FAST-013-001, HEU-SOL-THERM 001-008, PU-MET-FAST-022-001, and PU-SOL-THERM 001-001. Ultimately, comparison is made between Upper Subcritical Limits (USLs) obtained using each code package for each application case. Since differences exist in whether packages take into account margin of subcriticality (MOS), the USL may be computed using bias and bias uncertainty, also known as the calculational margin (CM) in ANSI/ANS 8.24. Application of portions of MOS to the USL for nuclear data uncertainty of and potential code margin is referred to as USL herein. In either case, additional MOS is considered for actual application cases.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2023

Improvements in MCNP6.3 (Rev. 1) [Slides]

This presentation outlines improvements in the Monte Carlo N-Particle Code (MCNP) version 6.3. MCNP is an internationally renowned Monte Carlo particle transport code developed by Los Alamos National Laboratory.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2022

Sensitivity/Uncertainty Comparison Study Involving IRSN, LANL, and ORNL Tools to Support Validation [Abstract]

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2022

Sensitivity/Uncertainty Comparison Study Involving IRSN, LANL, and ORNL Tools to Support Validation

Under a DOE Nuclear Criticality Safety Program (NCSP) task involving Analytical Methods, three Laboratories collaborated in a comparison of results obtained from Sensitivity/Uncertainty (S/U) packages relevant to validation of transport codes. The task involves Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Los Alamos National Laboratory (LANL), and Oak Ridge National Laboratory (ORNL) comparing results of MORET 5/MACSENS V3.0, MCNP6.2/Whisper-1.1, and SCALE 6.2.3/TSUNAMI/USLSTATS respectively. All Monte Carlo transport code results utilize nuclear data from ENDF/B-VII.1 evaluation. This study examines five cases from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) selected as application models: IEU-MET- FAST-002-001, LEU-COMP-THERM-001-001, LEU-SOL-THERM-004-001, MIX-COMP- THERM-001-001, and U233-SOL-THERM-001-001. This is a continuation of a previous study to examine Pu and HEU cases: HEU-MET-FAST-013-001, HEU-SOL-THERM-001-008, PU-MET- FAST-022-001, and PU-SOL-THERM-001-001. Ultimately, comparison is made between Upper Subcritical Limits (USLs) obtained using each code package for each application case. Since differences exist in whether packages take into account margin of subcriticality (MOS), the USL is computed using only bias and bias uncertainty, also known as the calculational margin (CM) in ANSI/ANS-8.24. Results comparison appears to show that benchmark selection has a greater influence on the USL than the method used for calculation of bias and bias uncertainty.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2015

GENIE final state interactions

Final state interactions are an important component of any neutrino-nucleus Monte Carlo program. GENIE has 2 FSI programs which serve different purposes. Each has fair-good agreement with a wide range of hadron-nucleus data. Recent improvements and planned advancements are described.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗