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At least 397 records · Page 22
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Radiation Transport Simulation of the Sulfur Counting System in the Radiation Metrology Laboratory (RML)

At the request of staff members from the Radiation Metrology Laboratory (RML), a series of Monte Carlo radiation transport calculations were performed using two different models of the detector geometry of the RMLs sulfur counting system. The fraction of electrons from each β-decay of 32 P in the sulfur pellet that enter the window of the sulfur counting system was calculated with both MCNP and ITS. In addition, the differential energy distributions of the electrons entering the counting system window were computed. There was significant agreement between the integral and differential quantities calculated by the two transport codes. Summary tables are for the integral efficiency values are found in the body of the report.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Proof that Combining the Forced-collision and DXTRAN Monte Carlo Variance-reduction Techniques is Fair

This report provides mathematical proof that the MCNP DXTRAN (also known in other codes as forced-flight) and forced-collision variance-reduction techniques do not bias the expected value of Monte Carlo simulation estimates when combined. Proof is also provided that the techniques are unbiased when used independently. To prove that the techniques are unbiased, this report derives the first History Score Moment Equation (HSME) for non-multiplying media employing only forced collisions, only DXTRAN, and then both the variance-reduction techniques combined, as defined next. The HSMEs are found by forming the History Score Probability Density Functions (HSPDFs) and then taking the first score moment. Following its derivation, each HSME with a variance-reduction technique employed is reduced to the HSME for an analog simulation. Because the HSME represents the expected contribution to estimators in the simulation, reducing the HSME with variance reduction to the analog HSME shows that the simulation is unbiased despite the variance-reduction technique considered in the HSME. Analysis regarding higher-score moments of each technique is the subject of prior work and is not addressed herein. Throughout this work, the phase space p is defined to be the particle position x, direction-of-flight unit vector $\hat{Ω}$, energy E, and statistical weight w, $$p ≡ (x; \hat{Ω}; E; w).$$ A reduced phase-space excluding the statistical weight of the particle, $$r ≡ (x; \hat{Ω}; E),$$ is also used.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2021

Verification of Combined VR Techniques, Derivation of Future Time Equation, and Integration of LLNL Pulsed Sphere V&V Suite [Slides]

To determine whether the combination of forced-collision and DXTRAN variance-reduction (VR) techniques is unbiased, and to gain insight into the operation of these techniques, proof that first-moment estimates from Monte Carlo simulations employing both techniques are unbiased is developed. A general background on the forced-collision and DXTRAN VR techniques and their combination is given. Proof of an unbiased simulation is outlined by showing the equivalence of the history score moment equations of simulations with these techniques in use. A report with detailed proof of this equivalence is available upon request. The derivation of the future time equation using a similar approach, as well as a summary of the addition of the LLNL Pulsed Sphere experiments to the MCNP verification and validation suite, is also briefly discussed.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2021

Neutron Transmission Imaging with a Portable D-T Neutron Generator

Fast-neutron transmission imaging provides complementary information to x-ray transmission imaging. While fast neutron imaging resolution is generally below x-ray imaging, 14-MeV neutrons have an advantage over portable x-ray systems. Neutrons have higher transmission through high-Z materials due to a more uniform attenuation as a function of material atomic number Z compared to X-rays, and can therefore image low-Z materials inside high-Z materials. As a result, portable neutron transmission imaging has many applications, including inspection of concrete and welds for corrosion in vehicles, bridges, and other infrastructure, measurement of material levels in containers, and inspection of suspicious packages. Fast-neutron imaging is also more practical for field use than thermal-neutron imaging due to the size and shielding requirements typical of thermal-imaging systems compared to the availability of small 14.1 MeV D-T neutron generators. However, there are limitations in portable fast-neutron imaging systems, including limited neutron output, limited light produced by neutron scintillators, and lower resolution due to neutron source spot size and 2-3 mm scintillator thickness. In addition, digital-panel dark-noise is roughly 100x higher than neutron scintillator light, and variations in noise across the panel and in time is comparable to the imaging signal. Here we discuss recent efforts in developing a portable fast-neutron radiography system, including an improved neutron scintillator, mitigation of panel noise, and new commercial portable D-T neutron generators. We also present MCNP efforts to model neutron imaging, including scintillator resolution and the effects of neutron scattering from the object and surrounding materials.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

ECAR 4951 BSU-8242 3 DPA As-Run Physics Analysis

The purpose of this report is to document the as-run physics analysis for the Boise State University (BSU-8242) experiment in the A6, A7 and A8 positions, correlating with the 3 DPA specimens irradiated in Cycles 164A, 164B, and 166A. In this paper the heat rate, flux, fluence, and DPA for all specimens are scaled to the appropriate as-run lobe powers for an accurate assessment and description of the irradiated materials. Source terms per capsule are provided, and a bounded source term per specimen type is given on a per gram basis. Verified and validated MCNP full-core models were used for the neutronics analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

DRiFT - Release 1.0.0 Organic Scintillators

DRiFT (a Detector Response Function Toolkit) is LANL-developed software that postprocesses output from the extensively validated radiation transport code, MCNP [1], and generates realistic nuclear instrumentation response. DRiFT is designed to be flexible, enabling users to specify detector type and many experimental settings, as well as accommodating the addition of their own desired features. Although DRiFT development has included scintillator [2], gas [3], and semiconductor features [4], the focus of this release is on organic scintillator and associated capabilities. Organic scintillators are widely used in the areas of nuclear safeguards and nuclear non-proliferation efforts [5, 6]. DRiFT has several diagnostic and detector physics features relevant to detailed scintillator simulations including: tracking source particle information, scintillation light production, the effects of PMT quantum efficiency and gain, and digitizer settings. Users can select responses from many scintillator and PMT types supported natively by DRiFT, or add their own by following the instructions in this document. We acknowledge that DRiFT is under active development, bug reports and general questions and comments should be directed to Madison Andrews, madison@lanl.gov. This manual is divided into four parts: I) An overview of DRiFT, including how to obtain and install the executable, II) A description of the detector physics related to scintillators available, III) a description of more general DRiFT features the user may find useful, and IV) a description of the test suite and examples made available with the code release.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Mini minutes for a mini CSEWG Virtual mini-CSEWG Meeting 16-19 August, 2021

This meeting is a collaboration meeting of the Cross Section Evaluation Working Group (CSEWG), a collaboration including members from various DOE NNSA and non-NNSA laboratories as well as several university groups that support these laboratory missions. The focus of this meeting will be preparations for the next major release of the ENDF/B nuclear data library. The ENDF/B library is the main neutronics library used in various US neutronics codes such as MCNP and SCALE for advance reactor work, safeguards, stockpile stewardship, etc.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

External Release of Organic Scintillator DRiFT Software - Technology Evaluation and Demonstration Project Accomplishments [Slides]

Overview and Objectives: DRiFT post processes radiation transport output from MCNP, and generates realistic nuclear instrumentation response; Organic scintillator response has been developed, tested, and validated with measurements, these instruments are widely used in nuclear safeguards. At the beginning of this project, DRiFT was only available to LANL employees with access to yellow High Performance Computing clusters. Additionally, there are many complicated external software dependencies. Our TED project request proposed using funds to: Comment and clean up code, friendly testing, generating test suites and examples; Expansion of scintillator types natively supported by DRiFT; External release of DRiFT accompanied by documentation. We accomplished all objectives (and many others), they are highlighted in these slides.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Recent MCNP6 ® Code Developments and Improvements for Nuclear Engineering Applications [Slides]

The Los Alamos MCNP Monte Carlo radiation transport code has been the international gold standard for particle transport applications for over three decades. Many developments to the code have taken place with several significant new feature additions, major improvements, and enhancements to existing features. With significant institutional and programmatic investment in the code since the time of the last public release in 2018, important code development and infrastructure modernization has taken place and remains a high priority for all ongoing efforts across the code development team.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

TNSL Overview

Thermal neutron scattering law (TNSL) data describe low-energy neutrons scattering off of bound materials, and can have a significant impact on modeling any system with slow neutrons, including nuclear reactors. Previous work to introduce TNSL data to neutron transport codes at LLNL focused on COG and TART [1], with the limitation that these codes require highly specialized data processing and formatting. We have recently increased efforts to process TNSL data with the central LLNL nuclear data processing code FUDGE, to be stored in the generalized nuclear database structure (GNDS) for use in any general transport code with the ability to read GNDS data. The first step in this effort is to verify that the TNSL processing with FUDGE yields results comparable to results obtained using the LANL nuclear data processing code NJOY. The next step is to verify the transport of thermal neutrons in Mercury (a Monte Carlo code) and Ardra (a deterministic code) against one another, as well as against the LANL Monte Carlo neutron transport code MCNP. This verification step has not been completed, due to a number of discrepancies between results obtained using differently processed data. There is ongoing effort to understand differences between FUDGE and NJOY. Finally, we map out our current capability to validate TNSL data against benchmark systems.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

As-Run Physics Analysis for the EPRI Zirconium Growth C Capsule

This engineering calculation and analysis report documents the as-run nuclear-heating term, source-term, and radiation damage (measured in displacements per atom [dpa]) accumulation for the Electric Power Research Institute zirconium growth experiment's C capsule (EPRI-ZG-C). These data have been calculated using MCNP 6.1 and SCALE 6.2.3. These data can be used for: the shipping of the EPRI-ZG-C capsule, the acceptance of the capsule at the Materials and Fuels Complex, and programmatic scientific needs.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Radiation Damage Calculation Methodology

This report outlines the consensus radiation damage calculation methodology used by the Neutronics Analysis Group at Idaho National Laboratory. This methodology includes calculation of radiation damage rates via the Monte Carlo N-Particle (MCNP) code, and calculation of radiation damage in mixed materials, such as alloys. A consistent method for calculating the average displacement threshold energy of a mixed material has been established by the Neutronics Analysis Group. No standard method currently exists to account for radiation damage caused by neutron-induced transmutations. Only one documented case addresses transmutation radiation damage in materials containing nickel, and that method was incorporated by the Neutronics Analysis Group into their radiation damage calculation methodology.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Nuclear Data: What, Why How, Who? [Slides]

Nuclear data is data that describes physics in forms and formats that computer code can read to perform simulations of nuclear processes. The data is put into ”evaluations” following the combination of experiment and theory. Various physics are applied to these evaluations by the NJOY nuclear data processing code to produce application files, which simulation codes (e.g., MCNP, Partisn, etc.) can use to model various scenarios. The Nuclear Data Team at Los Alamos National Laboratory—in partnership with a variety of national and international organizations—provides nuclear data for use at LANL and throughout the world. This data is verified and validated to ensure that it performs as expected and accurately represents Mother Nature.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Fission Matrix Processing Using the MCNP6.3 HDF5 Restart File

This paper describes an approach to interrogating the fission matrix available in the HDF5-formatted restart file (also known as the “runtape” file), which is produced by the upcoming public release of the MCNP R code, version 6.3. The fission matrix, its eigenvalues, and its eigenvectors, are important tools for characterizing fissile systems such as research and power reactors and for accelerating the convergence of the Monte Carlo fission-source site distribution.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Methods for Estimating Dose from Galactic Cosmic Rays and Jovian Radiation Belts on Reduced-Order Monte Carlo Geometries

This document outlines the methods and tools developed to estimate dose rates from galactic cosmic rays and the trapped radiation belts surrounding Jupiter using Monte Carlo particle transport software. Furthermore, the techniques described here are specific for translating high-resolution MAVRIC models to analogous but lower resolution Monte Carlo N-Particle Transport (MCNP) code models to estimate charged particle effects on models originally developed to study neutral particles only. This capability was initially developed to predict doses to various components of dynamic radioisotope power systems for spaceflight applications, but the tools and methods can be easily applied to other problem geometries and expanded to other space radiation environments.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2022

CAAS-3S Radiation Testing for Y-12 and UPF with Godiva-IV

Goals: 1) Qualify the CAAS-3S system to a mixed-field radiation dose and dose rate, and 2) Extend the Y-12 Shielding MCNP Validation to rad-si. This paper represents the efforts of Y-12 and UPF personnel, and their interpretation of the test results.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Thermal Neutron Scattering Law (TNSL) Implementation and Testing in FUDGE

Thermal neutron scattering law processing capabilities have been recently implemented in the LLNL code FUDGE (For Updating Data and Generating Evaluations). FUDGE is now capable of producing processed TNSL data for use in Monte Carlo or deterministic transport. To test this new capability, LLNL scientists performed an extensive intercomparison between the Mercury, Ardra, COG and MCNP transport codes. This intercomparison helped probe differences between how TNSL data are handled by the transport codes and by two different processing codes (LLNL’s FUDGE and LANL’s NJOY). This report summarizes recent improvements in FUDGE TNSL capabilities as well as results from the code intercomparison.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗