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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

(U) Using the MCNP6 Perturbation Capability for Source Nuclide Density Sensitivities

The Taylor series (differential operator) perturbation method, as implemented in the PERT capability in MCNP6 (Ref. 1), can be used for first- and second-order sensitivity analyses in fixed-source problems. Unfortunately, the PERT capability does not yet work with point-detector (F5) or pulse-height (F8) tallies. Also, the second-order Taylor series term lacks the cross terms that would allow the calculation of mixed sensitivities. Finally, the PERT capability does not compute any sensitivities involving the source in a fixed-source problem. When neutrons or gamma rays are emitted due to radioactive decay, the source rate density depends intrinsically on the nuclide densities in the source material. When the PERT capability in MCNP6 is used to compute sensitivities of a response to nuclide and material densities, those sensitivities do not include the effect of the source.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Adaptive-in-temperature method for fast on-the-fly sampling of thermal neutron scattering data in MCNP6 [Slides]

The project objective and motivation is to develop thermal data libraries for selected isotopes in MCNP6 to support on-the-fly S(alpha, beta) sampling for temperature ranges applicable to nuclear criticality safety, thereby enhancing the physics treatment in MCNP6 so that it can perform fast on-the-fly sampling of S(alpha, beta) data at arbitrary temperature.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

A Prototype Thick-Target Bremsstrahlung Model with Angularly-Dependent Emission in the MCNP6 ® Code

This document summarizes the current thick-target bremsstrahlung (TTB) model in MCNP and provides test results for an alternative implementation to improve the accuracy with reduced cost compared to full electron transport. It has been observed that the current TTB model produces inaccurate results in problems where the medium is thick with respect to electrons, but the photon distribution in the problem has a strong directionality. An example of such a simulation is detectors surrounding a metal target irradiated with a radiographic beam of high energy photons. The primary cause of this discrepancy is the current TTB method emits all bremsstrahlung photons in the same direction as the primary electron produced from each (γ, e ± ) interaction, leading to artificially forward peaked photon distributions for intermediate to high-energy incident photons. To improve the TTB model, we have implemented an angularly-dependent TTB model in a developer version of the MCNP6 ® code; for developers, this was done on the branch prototype/angular_ttb in the mcnp6 repo on bitbucket. The angularly-dependent TTB model accounts for the energy and scattering of electrons as they slow down in the current material, but does not sample the computationally expensive energy straggling, secondary electron events, and tracking electrons; this approach is significantly less computationally expensive than full electron transport and can be comparable to the original TTB method for problems with sufficiently complex materials and geometry. To evaluate the method, we have modeled a simple problem of a beam of 5 MeV photons incident on a sphere of plutonium surrounded by detectors at different deflection angles. For this problem, the angularly-dependent TTB produces a photon flux within 8.0% for a 90 degree deflection angle and 0.8% along the beam axis, as compared to the electron transport solution. This is an improvement compared to a 43% and 81% discrepancy with the original TTB method, respectively. The rest of this work includes the following: the first section details the current TTB treatment in MCNP, which has not been well documented elsewhere. Then, the modified TTB algorithm is detailed and the approximations compared to the condensed history algorithm are compared. Results are given comparing the two TTB methods to the condensed history transport algorithm. The appendix includes details for code developers on relevant electron transport implementation details and potential code improvements for future work.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

An update to δ-ray production in MCNP6

As originally introduced in MCNP6.2, the δ-ray production subroutine had a number of flaws related to treatment of particle or nuclear spin contributions to the production cross section. A recent update by the author to the MCNP6 code corrects these flaws. This report documents those changes for the benefit of MCNP6 users who use the delta ray production capability. Additionally, this report aims to provide a more comprehensive documentation of the delta ray production, building on the prior documentation by C. Anderson and coworkers. On this latter point, the goal is to save future MCNP developers from a visit to the library to dig out ancient texts from 1952, or at least to provide useful points of reference should such a visit be necessary.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Parallel Programming in MCNP6

Monte Carlo N-Particle (MCNP)1 is a general-purpose Monte Carlo particle transport code developed by Los Alamos National Laboratory (LANL). To efficiently handle long simulations, MCNP version 6 (MCNP6) supports parallel execution using two primary programming models: • Shared-memory task-based threading using OpenMP (Open Multi-Processing), and • Distributed-memory calculations using MPI (Message Passing Interface). The OpenMP and MPI programming models enable MCNP6 to scale from desktop systems to high-performance computing (HPC) clusters, allowing users to run MCNP in one of three parallel modes: • OpenMP-only, • MPI-only, and • Hybrid (MPI + OpenMP). The choice of parallelization mode depends on the underlying computer architecture and the characteristics of the simulation problem.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2012

MCNP6 Fission Multiplicity with FMULT Card

With the merger of MCNPX and MCNP5 into MCNP6, MCNP6 now provides all the capabilities of both codes allowing the user to access all the fission multiplicity data sets. Detailed in this paper is: (1) the new FMULT card capabilities for accessing these different data sets; (2) benchmark calculations, as compared to experiment, detailing the results of selecting these separate data sets for thermal neutron induced fission on U-235.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2012

MCNP6 Status

We, the development teams for MCNP, NJOY, and parts of ENDF, would like to invite you to a proposed 3 day workshop October 30, 31 and November 1 2012, to be held at Los Alamos National Laboratory. At this workshop, we will review new and developing missions that MCNP6 and the underlying nuclear data are being asked to address. LANL will also present its internal plans to address these missions and recent advances in these three capabilities and we will be interested to hear your input on these topics. Additionally we are interested in hearing from you additional technical advances, missions, concerns, and other issues that we should be considering for both short term (1-3 years) and long term (4-6 years)? What are the additional existing capabilities and methods that we should be investigating? The goal of the workshop is to refine priorities for mcnp6 transport methods, algorithms, physics, data and processing as they relate to the intersection of MCNP, NJOY and ENDF.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2011

Advances in the Development and Verification of MCNP5 and MCNP6

The MCNP Monte Carlo code has been used for high-fidelity analyses of criticality safety problems since the 1970s. This paper reviews recent advances in the development and verification of MCNP, including the current production release of MCNP5 and the beta release of MCNP6. End-users in all application areas need to be aware of the forthcoming MCNP6 release and begin planning for the transition to the new code in 2012 and beyond.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

MCNP6 Cosmic & Terrestrial Background Particle Fluxes -- Release 4

Essentially a set of slides, the presentation begins with the MCNP6 cosmic-source option, then continues with the MCNP6 transport model (atmospheric, terrestrial) and elevation scaling. It concludes with a few slides on results, conclusions, and suggestions for future work.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2013

Comparison of MCNP6 and experimental results for neutron counts, Rossi-{alpha}, and Feynman-{alpha} distributions

MCNP6, the general-purpose Monte Carlo N-Particle code, has the capability to perform time-dependent calculations by tracking the time interval between successive events of the neutron random walk. In fixed-source calculations for a subcritical assembly, the zero time value is assigned at the moment the neutron is emitted by the external neutron source. The PTRAC and F8 cards of MCNP allow to tally the time when a neutron is captured by {sup 3}He(n, p) reactions in the neutron detector. From this information, it is possible to build three different time distributions: neutron counts, Rossi-{alpha}, and Feynman-{alpha}. The neutron counts time distribution represents the number of neutrons captured as a function of time. The Rossi-a distribution represents the number of neutron pairs captured as a function of the time interval between two capture events. The Feynman-a distribution represents the variance-to-mean ratio, minus one, of the neutron counts array as a function of a fixed time interval. The MCNP6 results for these three time distributions have been compared with the experimental data of the YALINA Thermal facility and have been found to be in quite good agreement. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Reactor physics verification of the MCNP6 unstructured mesh capability

The Monte Carlo software package MCNP6 has the ability to transport particles on unstructured meshes generated from the Computed-Aided Engineering software Abaqus. Verification is performed using benchmarks with features relevant to reactor physics - Big Ten and the C5G7 computational benchmark. Various meshing strategies are tested and results are compared to reference solutions. Computational performance results are also given. The conclusions show MCNP6 is capable of producing accurate calculations for reactor physics geometries and the computational requirements for small lattice benchmarks are reasonable on modern computing platforms. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Methodology, verification, and performance of the continuous-energy nuclear data sensitivity capability in MCNP6

A continuous-energy sensitivity coefficient capability has been introduced into MCNP6. The methods for generating energy-resolved and energy-integrated sensitivity profiles are discussed. Results from the verification exercises that were performed are given, and these show that MCNP6 compares favorably with analytic solutions, direct density perturbations, and comparisons to TSUNAMI-3D and MONK. Run-time and memory requirements are assessed for typical applications, and these are shown to be reasonable with modern computing resources. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

MCNP6 Modification for the Calculation of Rossi and Feynman Distributions

The Rossi distribution is given by the number of captured neutron pairs as a function of the time interval between the two capture events. This distribution contains a random part, which is given by neutron captures caused by neutrons of different fission chains, and a correlated part, which is given by neutron captures caused by neutrons from the same fission chain. The correlated part exponentially vanishes with the {Delta}t time interval following the decay of prompt neutrons. The Rossi distribution peaks when the time interval between two capture events is small since this condition is satisfied for neutrons generated from the same fission chain. In the Poisson probability distribution, the variance is equal to the mean and the next event probability is independent of the elapsed time since the previous event. Neutron captures events do not follow the Poisson probability distribution because they are correlated events via fission chains. Consequently, the Feynman distribution Y is defined as the variance to mean ratio minus one, of the neutron counts array as a function of a fixed time interval. The MCNP6 source code has been modified to allow the calculation of the Rossi and Feynman distributions. The source code modification takes into account the time when a neutron is emitted by the external neutron source and gives a precision in the timestamp lower than a fraction of nanosecond. The modified MCNP6 version has been used to model KUKA facility driven by a californium neutron source. The prompt neutron decay constants obtained by the Rossi and Feynman methods are in agreement.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Semi-Analytical Benchmarks for MCNP6

Code verification is an extremely important process that involves proving or disproving the validity of code algorithms by comparing them against analytical results of the underlying physics or mathematical theory on which the code is based upon. Monte Carlo transport codes such as MCNP6 undergo verification and testing upon every release to ensure that the codes are properly simulating nature. Specifically, MCNP6 has multiple sets of problems with known analytic solutions that are used for code verification. Most of the verification problems are fairly simple due to the fact that they are usually composed of few materials, with simplified cross-sections, or physics approximations, and simple geometry. These approximations make the transport equation possible to solve analytically or semi-analytically, and thus they are quite valuable in verifying that the code algorithms and methods (Monte Carlo, discrete ordinates, etc.) are operating as intended. The remainder of this paper is organized as follows: first we briefly discuss the theory behind the benchmarks problems, then the numerical results are presented, and finally we conclude with general remarks on this and possible future work.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Speed and Memory Improvements to MCNP6 Delayed-Gamma Line Treatment

The delayed-gamma (DG) line-emission (line) sampling was first introduced into MCNPX in 2005. At the time, storage demands for problems with large numbers of DG lines was a major concern. In order to accommodate large amounts of line data while keeping the memory footprint of such a problem from exceeding memory limits, complicated logic was included to sift through various storage options. In mid-2015 concerns arose over the adequacy of the MCNP6 DG source option (inherited from MCNPX) with line sampling for modeling high-fidelity problems. One indication that there was a problem came when studying the MCNP6 produced DG spectra of U-238 and its progeny using high-fidelity spectroscopic tallies when compared to DG line amplitudes directly from ENDF/B-VII.1 emission data. Despite producing a correct integral number of gammas, the comparison to ENDF data showed a broadening in the energy lines and a reduction in the amplitude of more prominent lines. It was later found that the cause of this broadening was one of the memory-saving techniques used to store DG line data when problems required many DG lines. Specifically, the DG lines were integrated into fine energy bins for sampling. When a DG line was to be sampled, a random number would be used to choose which energy bin to sample based on emission probability. Once chosen, a second random number was used to perform nearest-line sampling in that bin. The integration of lines led to an oversampling of low-probability lines, especially when in close proximity to high-probability lines. This treatment also caused a similar issue with the time-dependent behavior of DG line emission which occurred when low-probability lines were sampled near high probability lines with the former decay constant being used. The effect on time-dependent gamma production was demonstrated in analytic benchmarks performed by Weldon et al.. To address these issues a new gamma line storage and sampling treatment was introduced in 2015. This new treatment, stores gamma emission data in a line-by-line CDF which is sampled directly, thereby resolving both the line-broadening issue, and the time-dependence issue as was demonstrated by Tutt el al.. In addition, the new DG sampling treatment was written to use dynamically dimensioned storage arrays as opposed to large fixed-dimensioned arrays, as in the former treatment, and also makes more efficient use of line data that is stored by reducing or eliminating repeated data. As a result, storage arrays are generated only for DG data relevant to the problem being executed. The result should be a reduction in both memory requirements and runtime. However, these reductions could not be realized until the logic for the old treatment could be removed. This paper provides an update on recent work which has focused on the removal of the old DG sampling logic and attempts to quantify the memory reduction and speed up of selected problems.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

Progress of full-scale MCNP6{sup TM} model development of NuScale small modular reactor - 25394

This paper presents the recent progress of development of full-scale MCNP6{sup TM} small modular reactor within NuScale Power LLC (NuScale) using MCNP6{sup TM} version 1.0. This model currently is being used to determine of vessel neutron fluence, nuclear damage, and radiological consequence analysis. The results of the best estimate fluence analysis to support Design Certification Application (DCA) and technical details has been included within License Technical Report (LTR) which had been submitted to US NRC for review in 2017. Currently the detailed analysis of using this full-scale NuScale SMR model is focusing on operational radiological external dose exposure, nuclear damage, nuclear heating and neutron activation of SMR components to support optimized design and RAI (request additional information) response activities. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

MCNP6 Elevation Scaling of Cosmic Ray Backgrounds - Paper 29

The galactic cosmic-ray (GCR) source option was implemented in the all-energy, all-particle transport code MCNP62 in 2010. This source option along with other newly added features to MCNP6 have been used to produce increasingly accurate global cosmic and terrestrial background flux data files (background.dat).3 This data file is produced on a grid and is read and sampled by MCNP6 whenever a user invokes the background source option. The most recent adaptation of this data file, Version 4, will be completed in 2014. This paper reports on current progress toward enhancements that have been made to the modeling and simulation of these background spectra, in particular the elevation scaling used to produce more accurate spectra at off-grid locations. Cosmic radiation continuously bombards Earth with various particles, such as protons, alpha particles, and heavier nuclei, some of which are deflected by the Earth's shielding magnetic field. Particles that carry sufficient momentum can overcome the deflection and penetrate into the atmosphere. The sufficient momentum is dependent on the terrestrial coordinates due to the shape of the Earth's magnetic field and the Lorentz force's proportionality to the sine of the angle between the velocity vector of the incoming particle and the magnetic field direction. As the particles propagate through the atmosphere, collisions with atmospheric molecules generate secondary particles such as neutrons, protons, photons, muons, pions, and other exotic particles. These secondary particles often have sufficient energy to undergo additional nuclear interactions, and so on, forming what is known as a cascade shower. The particles in the cascade shower are constantly undergoing energy losses due to interactions, especially charged particles whose paths are bent into helical trajectories around magnetic field lines. These energy losses make the integral flux of the cosmic background at the ground level highly dependent on the atmospheric depth traveled by the incident particle and its progeny, and therefore dependent on the elevation of the ground where the cascade deposits its energy. For relatively small depth changes of approximately 100 g/cm{sup 2} or less, the ground-level cosmic flux follows an exponential model. The tabulation of background particle fluxes on the surface of the earth is important for a variety of reasons, one of which is the design of nuclear material detection systems. This method is also useful in producing user-specified effective attenuation lengths for particles. Attenuation lengths are an essential parameter used to compare particle spectra from different locations, elevations, and energy ranges. Traditionally these attenuation lengths are inferred from measured data. Now with the ability to simulate an accurate galactic cosmic ray source and transport it and its shower through a modeled atmosphere, these attenuation lengths can be calculated for any set of conditions. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

Validation of a Monte Carlo model for the MLC-Elekta Precise with MCNP6 - Paper 76

Cancer treatment using radiotherapy techniques needs an accurate prediction of the dose delivered to the patient at each position. Treatment planning systems (TPS) allow the optimization of the dose delivered to the tumor volume while minimizing the dose absorbed by the surrounding healthy tissue. Commercial treatment planning systems use empirical deterministic algorithms that do not correctly estimate the dose in the surroundings of heterogeneities. This lack of accuracy is taken into account as bigger security margins in the treatment, reducing the potential effect of the treatment. However, Monte Carlo (MC) simulation techniques have proved to predict the dose distribution more accurately and also representing experimental data better. Nevertheless, the clinical application of these techniques has been limited by the long calculation times they require. As high performance computers decrease in cost MC simulations become more feasible to be used in TPS. The present work improves a simplified model of the MLC-Elekta Precise linear accelerator (linac or unit from now on) previously developed and correctly validated by the group. A multileaf collimator (MLC) has been modelled according to the manufacturer specifications and added to the previous model to allow higher conformity in the studies. The full geometry has been entered to the MCNP6 Monte Carlo code to simulate the transport of particles through the linac. The geometry also includes a cubic water phantom for dosimetry studies and validation with experimental data. The experimental data employed for the dose distribution comparison has been measured at the linac from the linac Hospital Universitari Clinic de Valencia (Spain). Additionally, particles are stored in a phase space surface before their entrance in the water to be analyzed and for performance optimization. Given the results presented, we conclude that this first approach to validate a model of the Elekta Precise linac from the Hospital Cl nic Universitari de Valancia represents well the depth-dose distribution, but fails at representing the dose profiles. It has been shown that they predict too much penumbra, probably caused by the filtration between the leaves. In order to solve this problems, further work is needed to improve the simulated model. This work will mainly focus on the multileaf collimator. Also, the codes developed by the group to analyze phase space files produced by MCNP6 have proven highly useful. As work progress and the codes are debugged, they will be published so that the community can take advantage of them.

61 RADIATION PROTECTION AND DOSIMETRY↗