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Brown, Forrest B.

Publications and source records attributed to Brown, Forrest B..

At least 19 records

MCNP ® Code Version 6.3.1 Theory & User Manual

This document acts as a repository of knowledge for the Monte Carlo N-Particle (MCNP) transport computer code. It is maintained alongside the source code and attempts to introduce new users and re-familiarize experienced users with the theory and practices of using the MCNP code for the wide range of particle transport analyses that it is appropriate for. The latest version of the MCNP code, version 6.3.1, provides the Monte Carlo particle transport community with the latest feature developments and bug fixes in the MCNP code. The MCNP code version 6.0 and later is also known as the MCNP6 code.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Enhancing Monte Carlo Workflows for Nuclear Reactor Analysis with Metamodel-Driven Modeling

Monte Carlo codes are essential components of many reactor physics simulation workflows as high-fidelity continuous-energy neutron transport solvers. Among Monte Carlo radiation transport codes, MCNP is particularly notable due to its diverse simulation capabilities, large user base, and long validation history. Despite being a powerful simulation tool, MCNP provides limited capabilities to allow automated execution, model transformation, or support for user-defined logic and abstractions that limit its compatibility with modern workflows. Here, to better integrate MCNP into a modern scientific workflow, we have developed an intuitive yet full-featured MCNP Application Program Interface (API) in Python, named MCNPy, which provides a specialized set of classes for MCNP input development. Moreover, to guarantee that our reading, writing, and modeling capabilities remain self-consistent (and to render the huge scope of the MCNP API manageable), we have adopted a strategy of model-driven software development in which a generalized model of the MCNP input format has been created. From this generalized model, or “metamodel,” problem-specific implementations such as an engine for input validation or a codebase for programmatic operations may be automatically generated. Since MCNPy primarily acts as a Python front-end to the underlying Java API that directly interfaces with the metamodel, it is intrinsically linked to the metamodel and thus remains maintainable. With MCNPy, users can programmatically read, write, and modify any syntactically valid MCNP input file regardless of its origin. These capabilities allow users to automate complicated tasks like design optimization and model translation for nuclear systems. As examples, this work demonstrates the use of MCNPy to find the critical radius of a plutonium sphere and to translate a 9000+ line MCNP input file into a corresponding OpenMC model.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

MCNP ® Code V.6.3.0 Release Notes

The Monte Carlo N-Particle ® (MCNP ® ) code is a general-purpose, continuous-energy, generalized geometry, time-dependent, radiation transport code developed by the MCNP development team. The MCNP calculations provide predictive capabilities that can replace expensive or impossible-to perform experiments. Specific application problems include simulations of experimental diagnostics, intrinsic radiation, radiation detection and measurement, criticality safety, nuclear threat reduction and response, radiation health protection, nuclear weapons effects, and nuclear forensics. This MCNP code, version 6.3.0, follows the MCNP6.2.0 version. Since the release of MCNP6.2.0, many changes have been made to the MCNP code. These changes include new or improved features, a new build system, code enhancement and modernization, and bug fixes. The MCNP code, version 6.3.0, theory and user input information is documented in MCNP ® Code Version 6.3.0 Theory & User Manual, the build guidance for various platforms is documented in MCNP ® Code Version 6.3.0 Build Guide, and the verification and validation testing for various application benchmark test suites is documented in MCNP ® Code Version 6.3.0 Verification & Validation Testing.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

MCNP ® Code Version 6.3.0 Theory & User Manual

This document acts as a repository of knowledge for the Monte Carlo N-Particle (MCNP) transport computer code. It is maintained alongside the source code and attempts to introduce new users and re-familiarize experienced users with the theory and practices of using the MCNP code for the wide range of particle transport analyses that it is appropriate for. The latest version of the MCNP code, version 6.3.0, provides the Monte Carlo particle transport community with the latest feature developments and bug fixes in the MCNP code. The MCNP code version 6.0 and later is also known as the MCNP6 code.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗

Comparison of the Baseline USL Calculation Methods for Loosely-Coupled and Novel Neutronic Systems [Slides]

Current work includes reconstructing 187-group ENDF B/VII.1 covariance matrix for comparison study to 44-group ENDF B/VII.1 matrix to determine how covariance matrix structures affect the USL calculations and investigating the exact domination mechanism of region-wise sensitivities in a loosely-coupled system. Future work will involve seeking to understand how bias distributions change with reactivity and how to calculate propagate distribution error into USL calculations, as well as how cross section perturbation studies can be extended to other types of calculations, such as shielding calculations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Upcoming MCNP6.3 ® Release: New Features and Improvements [Slides]

The latest version of the MCNP6.3 code will be done and released this year. Many new features, substantial improvements, and bugfixes have made their way into the code. Improved documentation, testing, and peripheral tools will also be delivered.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

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↗

Guide for Using ENDF/B-VIII.0 Nuclear Data with MCNP

This report provides guidance for using the ENDF/B-VIII.0 nuclear data that were released in 2018 and 2020 with MCNP. Appendix A: Obtaining and installing the new ENDF/B-VIII.0 ACE nuclear data. Appendix B. Obtaining & using XSDIR files that include ENDF/B-VIII.0 data. Appendix C: ENDF/B-VIII.0-based ACE files for neutron cross-sections. A listing of the available temperatures and ZA numbers for neutron cross-sections (ACE files with suffix .nnC). Appendix D: ENDF/B-VIII.0-based ACE files for thermal scattering law (TSL) data, $S$(α,β). A listing of the available thermal scattering $S$(α,β) data (ACE files with suffix .nnT) The next section below provides guidance for modifying existing $\textit{mcnp}$ input files to use the ENDF/B-VIII.0 data. Detailed results from running several benchmark suites for nuclear criticality safety with both ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data are reported in. It is important to note that this report deals only with TSL data released in 2020. With the original release of ENDF/B-VIII.0 in 2018, the ACE files for neutron cross-sections were contained in a directory called Lib80X and the ACE files for TSL data were contained in a directory called ENDF80SaB. A few years later, it was determined that many of the TSL data files had been generated with faulty data or with incorrect processing, and a new set of ACE files for TSL data was created, with the corrected files contained in directory ENDF80SaB2.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

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↗

Procedure for Creating, Reviewing, and Submitting Input Files into the Benchmark Library

Los Alamos National Laboratory (LANL) has a number of criticality benchmark input files, modeled for use with MCNP, developed and maintained by various organizations including the Nuclear Criticality Safety (NCS) Division, XCP-3, and XCP-5. The purpose of this procedure is to standardize input file creation, review, and maintenance. This procedure incorporates existing practices used by IRSN as documented in the original report and the translation from French in addition to practices used by the LANL NCS Division as documented in. The report relates to input files used for the validation of the continuous energy Monte Carlo code MCNP6 but is general enough to be extended to other codes. This report describes the procedure on how to prepare and use benchmarks (taken mainly from the ICSBEP Handbook) for the validation of calculation routes. In particular, this report describes the procedure to follow in the creation of the input files and their integration in the experimental validation databases.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Comparison Study of Upper Subcritical Limits Derived Using Sensitivity/Uncertainty Tools Case Studies of Benchmarks and Applications

Neutron transport methods used to establish subcriticality require validation by comparison to critical experiments considered to be benchmarks. Whisper is a sensitivity/uncertainty analysis tool developed to assist with the task of validation in nuclear criticality safety. Details on the Whisper methodology can be found in References 1-3 on the MCNP ® reference collection website at https://mcnp.lanl.gov. Whisper-1.0 was originally developed in 2014 and used to assist with nuclear criticality safety validation at Los Alamos National Laboratory. Whisper was upgraded in 2016 to Whisper-1.1 and prepared for release with MCNP6.2. Whisper contains a library of over 1100 critical experiment benchmarks and quantifies neutronic similarity of an application to benchmarks in the library. Using highest similarity benchmarks, Whisper computes a calculational margin (CM) encompassing of the worst-case bias and bias uncertainty at a 99% confidence level for each application. In addition, portions of the margin of subcriticality (MOS) for nuclear data uncertainty and potential code errors are computed. The baseline upper subcritical limit (USL) computed by Whisper is comprised of the CM, MOS nuclear data , and MOS code errors . The Whisper baseline USL is absent a portion of the MOS due to the area of application, which is applied based upon judgment by the criticality safety analyst. An objective of this paper is to present the baseline USL, CM and portions of the MOS as computed by Whisper for comparison with similar sensitivity/uncertainty tools, such as those used by IRSN and ORNL. The initial comparison involves four critical experiment benchmarks: HEU-MET-FAST-013-001, HEU-SOLTHERM-001-008, PU-MET-FAST-022-001, AND PU-SOL-THERM-001-001, which have been: 1. modeled independently by LANL, IRSN, and ORNL based upon information provided in the ICSBEP Handbook, 2. are common in S/U libraries for LANL, IRSN, and ORNL, 3. span a range of energy spectrum and fissile material, and 4. taken as applications for the purposes of this study and therefore excluded from use as a benchmark for calculating the upper subcritical limit. Results presented in this paper have been computed using covariance data for all isotopes in ENDF/BVII.0 using a 44-group energy structure. Benchmarks in the Whisper library were run in MCNP6.2 using 100,000 neutrons per cycle, skipping 100 cycles for 500 active cycles. Reference 10 also compares the results for baseline USL with an order of magnitude greater neutrons, using the same total number of cycles with 1,000,000 neutrons per cycle. Subsequent to the results presented in Reference 10 changes were made to the benchmark library, as discussed in Reference 11. Newer results using the revised benchmark library are presented herein.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Overview of the MCNP6® SQA Plan and Requirements [Memorandum]

For all X Computational Physics Division (XCP) software under the Associate Laboratory Directorate for Weapons Physics (ALDX), the Weapons Research Services Secure Networks and Assurance Group (WRS-SNA) manages the software quality assurance (SQA) plan, requirements and guidance with respect to development processes and tools to meet the broader LANL SQA requirements. Each XCP software product is categorized into one of three software types: Safety Software, Non-Safety Risk Significant Software, and Non-Safety Commercially Controlled Software. In 2018, using LANL Form 2033, the MCNP6 code was categorized by the XCP division as Non-Safety Commercially Controlled Software, provided in Appendix A. Using WRSFORM- 0001U, the MCNP6 code was graded as a Medium Impact software product, provided in Appendix B. Given these determinations, the WRS-AD-0010U SQA plan is followed for all MCNP6 developments, documentation and code releases.

97 MATHEMATICS AND COMPUTING↗

Statistical Tests for Diagnosing Fission Source Convergence and Undersampling in Monte Carlo Criticality Calculations [Slides]

There is a very strong need for statistical testing to determine fission source convergence in Monte Carlo criticality calculations. Automation of such tests will greatly streamline and support the work carried out by NCS practitioners. Recent R&D work has shown that no single statistical test for convergence is sufficiently reliable, robust, and “guaranteed.” However, a combination of several standard statistical tests for the similarity of distributions, coupled with a high-fidelity estimate of the fission-matrix source is sufficiently robust, reliable, and repeatable that convergence can be “guaranteed.” During the course of the EG-AMCT studies, a number of statistical metrics and tests were proposed for diagnosing clustering and undersampling. None of these was robust and reliable enough for practical use in production codes. However, the expert group efforts came close. Some recent R&D work stemming from those past efforts has been very successful and promising. This Sub-Group will provide international input and collaboration on the development and implementation of statistical tests for convergence, with the primary goal of having the MC codes automatically detect convergence (or lack thereof). Newly proposed statistical tests to detect undersampling (after convergence) will also be reviewed.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

LANL Critical Benchmark Comparison Study and Subsequent Revision

As part of an international collaboration within the DOE Nuclear Criticality Safety Program (NCSP), LANL is involved in a comparison study to quantify differences in k-effective results from neutron transport simulations of critical benchmark experiments. The DOE NCSP Mission and Vision details the activity in which the French Institut De Radioprotection et De Sûreté Nucléaire (IRSN) leads the study with LANL and in conjunction with ORNL and LLNL to compare results of various neutron transport codes and nuclear data libraries to compute k-effective for ICSBEP benchmarks held in common by the entities. This report documents results obtained through partial completion of the overall effort with a focus on the changes made to LANL benchmarks modeled with MCNP6 using ENDF/B-VII.1 nuclear data that appeared to have discrepant results when compared with results of other codes. The feedback received through participation in the comparison collaboration has prompted an effort to review particular input files for benchmarks and revise when necessary. This report documents the results of review and revision of specific benchmarks highlighted as possibly discrepant in the comparison study. In addition, this effort prompted a new collaboration between LANL XCP and NCS Divisions in the development of a shared review/revision procedure and use of a new benchmark repository.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Doppler Broadening Resonance Correction for Free-gas Scattering in MCNP6.2

The traditional treatment for epithermal neutron scattering in continuous-energy Monte Carlo codes is the free-gas scattering model with the scattering cross-section assumed constant in energy. Ouislomen and Sanchez demonstrated in 1991 that scattering resonances can cause significant departures from the free-gas model, and Becker proposed a correction to the Monte Carlo free-gas scattering model in 2009. Becker’s Doppler Broadening Resonance Correction was tested in mcnp5 in 2011 and recently implemented in mcnp6.2 .

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Verification of MCNP6.2 with ENDF/B-VIII.0 Nuclear Data for Nuclear Criticality Safety Applications

MCNP6.2 was released in 2018 and is the latest production release of MCNP ® available from RSICC. MCNP6.2 includes all of the standard features for nuclear criticality safety (NCS) calculations that have been available for the past 15 years, along with new features for sensitivity-uncertainty based methods using Whisper for NCS validation. MCNP6.2 verification using ENDF/B-VII.1 nuclear data was previously reported in. This report focuses on the initial testing and verification of MCNP6.2 using the ENDF/B-VIII.0 nuclear data that were released in 2018. The sections that follow provide guidance for NCS users on modifying existing MCNP input files to use the new ENDF/B-VIII.0 ACE nuclear data, discussion of results for several ICSBEP-based benchmark suites using MCNP6.2 with ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗