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DOE OSTI2025

Modelling the brittle failure of graphite induced by the controlled impact of runaway electrons in DIII-D

The thermo-mechanical response of an ATJ graphite sample to controlled runaway electron (RE) dissipation, realized in DIII-D, is modelled with a novel work-flow that features the RE orbit code KORC, the Monte Carlo particle transport code Geant4 and the finite element multiphysics software COMSOL. KORC provides the RE striking positions and momenta, Geant4 calculates the volumetric energy deposition and COMSOL simulates the thermoelastic response. Brittle failure is predicted according to the maximum normal stress criterion, which is suitable for ATJ graphite owing to its linear elastic behavior up to fracture and its isotropic mechanical properties. Measurements of the conducted energy, damage topology, explosion timing and blown-off material volume, impose a number of empirical constraints that suffice to distinguish between different RE impact scenarios and to identify RE parameters which provide the best match to the observations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2014

Final report on NA22 Project: A Comprehensive Capability to Use Prompt Fission Signatures to Detect SNM

The objective of this project is to provide a general capability that will facilitate detection of high-sensitivity nuclear material for both active interrogation and passive systems. Currently these systems are able to use only a small fraction of the information carried by particles emitted during ssion. Our improvements in physics modeling and simulation tools would enable exhaustive use of prompt ssion signatures. In particular, this project will provide a computationally e cient event-by-event description of ssion that explicitly describes all of the correlations associated with the production and subsequent decay of fragments formed in ssion. We will use existing data to validate our model as far as possible, including trying to make use of data from current NA22 funded projects. Once complete, we will make our code, FREYA, publicly available either as a standalone code or as a module in Monte Carlo transport codes.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

NASA NTRS2014

Computation of Cosmic Ray Ionization and Dose at Mars: a Comparison of HZETRN and Planetocosmics for Proton and Alpha Particles

The ability to evaluate the cosmic ray environment at Mars is of interest for future manned exploration. To support exploration, tools must be developed to accurately access the radiation environment in both free space and on planetary surfaces. The primary tool NASA uses to quantify radiation exposure behind shielding materials is the space radiation transport code, HZETRN. In order to build confidence in HZETRN, code benchmarking against Monte Carlo radiation transport codes is often used. This work compares the dose calculations at Mars by HZETRN and the Geant4 application Planetocosmics. The dose at ground and the energy deposited in the atmosphere by galactic cosmic ray protons and alpha particles has been calculated for the Curiosity landing conditions. In addition, this work has considered Solar Energetic Particle events, allowing for the comparison of varying input radiation environments. The results for protons and alpha particles show very good agreement between HZETRN and Planetocosmics.

Gronoff, Guillaume↗

NASA NTRSDate not supplied

Coupled Reactor and Engine Nuclear Thermal Propulsion Modeling Methodology

Jacob Stonehill↗

NASA NTRSDate not supplied

Coupled Reactor and Engine Nuclear Thermal Propulsion Modeling Methodology

The design and development process of a Nuclear Thermal Propulsion (NTP) system requires extensive multiphysics modeling to couple the neutron physics and thermal feedback effects to determine the reactor’s power shape. Propulsion system performance codes utilize this power shape to determine NTP key performance parameters. While the power shape is heavily dependent on the temperature profile and geometry of the reactor, many analyses either assume a constant power shape, or use neutronics analysis to determine a power shape for a specific reactor configuration. The development of a coupling interface for a propulsion system performance code and a Monte Carlo neutron transport code (OpenMC) allows for the reactor power shape to be calculated in an iteration loop. The interface utilizes a file share system to transfer geometry dimensions, temperatures, and material identifiers to OpenMC, which is used to perform a neutron transport simulation of a design like the government Testing Reference Design reactor. The interface is then able to post-process the results from OpenMC and use the same file share system to share a power shape and other important neutron transport parameters to the system performance code. Initial results show that neglecting the changes to power shape when comparing reactor configurations can yield inaccurate results. Furthermore, utilizing propellants other than hydrogen gas can cause significant changes to the power shape, and thus, the thermal performance of a specific reactor design. This methodology is being expanded to allow for multiple families of NTP reactors to be analyzed, including block moderator, particle bed, and NERVA-derived reactors.

multiphysics coupling↗

NASA NTRSDate not supplied

Coupled Reactor and Engine Nuclear Thermal Propulsion Modeling Methodology

multiphysics coupling↗

NASA NTRS1984

Vectorized Monte Carlo methods for reactor lattice analysis

Some of the new computational methods and equivalent mathematical representations of physics models used in the MCV code, a vectorized continuous-enery Monte Carlo code for use on the CYBER-205 computer are discussed. While the principal application of MCV is the neutronics analysis of repeating reactor lattices, the new methods used in MCV should be generally useful for vectorizing Monte Carlo for other applications. For background, a brief overview of the vector processing features of the CYBER-205 is included, followed by a discussion of the fundamentals of Monte Carlo vectorization. The physics models used in the MCV vectorized Monte Carlo code are then summarized. The new methods used in scattering analysis are presented along with details of several key, highly specialized computational routines. Finally, speedups relative to CDC-7600 scalar Monte Carlo are discussed.

Brown, F. B.↗

NASA NTRS2004

The FLUKA code for space applications: recent developments

The FLUKA Monte Carlo transport code is widely used for fundamental research, radioprotection and dosimetry, hybrid nuclear energy system and cosmic ray calculations. The validity of its physical models has been benchmarked against a variety of experimental data over a wide range of energies, ranging from accelerator data to cosmic ray showers in the earth atmosphere. The code is presently undergoing several developments in order to better fit the needs of space applications. The generation of particle spectra according to up-to-date cosmic ray data as well as the effect of the solar and geomagnetic modulation have been implemented and already successfully applied to a variety of problems. The implementation of suitable models for heavy ion nuclear interactions has reached an operational stage. At medium/high energy FLUKA is using the DPMJET model. The major task of incorporating heavy ion interactions from a few GeV/n down to the threshold for inelastic collisions is also progressing and promising results have been obtained using a modified version of the RQMD-2.4 code. This interim solution is now fully operational, while waiting for the development of new models based on the FLUKA hadron-nucleus interaction code, a newly developed QMD code, and the implementation of the Boltzmann master equation theory for low energy ion interactions. c2004 COSPAR. Published by Elsevier Ltd. All rights reserved.

Non-NASA Center↗

DOE OSTI2016

Motivation for Using Generalized Geometry in the Time Dependent Transport Code TDKENO

We are proposing to use the code, TDKENO, to model TREAT. TDKENO solves the time dependent, three dimensional Boltzmann transport equation with explicit representation of delayed neutrons. Instead of directly integrating this equation, the neutron flux is factored into two components – a rapidly varying amplitude equation and a slowly varying shape equation and each is solved separately on different time scales. The shape equation is solved using the 3D Monte Carlo transport code KENO, from Oak Ridge National Laboratory’s SCALE code package. Using the Monte Carlo method to solve the shape equation is still computationally intensive, but the operation is only performed when needed. The amplitude equation is solved deterministically and frequently, so the solution gives an accurate time-dependent solution without having to repeatedly We have modified TDKENO to incorporate KENO-VI so that we may accurately represent the geometries within TREAT. This paper explains the motivation behind using generalized geometry, and provides the results of our modifications. TDKENO uses the Improved Quasi-Static method to accomplish this. In this method, the neutron flux is factored into two components. One component is a purely time-dependent and rapidly varying amplitude function, which is solved deterministically and very frequently (small time steps). The other is a slowly varying flux shape function that weakly depends on time and is only solved when needed (significantly larger time steps).

IQS↗

DOE OSTI2012

Monte Carlo simulations of channeling spectra recorded for samples containing complex defects

The main aim of the present paper is to describe the current status of the development of McChasy, a Monte Carlo simulation code, to make it suitable for the analysis of dislocations and dislocation loops in crystals. Such factors like the shape of the bent channel and geometrical distortions of the crystalline structure in the vicinity of dislocation has been discussed. Several examples of the analysis performed at different energies of analyzing ions are presented. The results obtained demonstrate that the new procedure applied to the spectra recorded on crystals containing dislocation yields damage profiles which are independent of the energy of the analyzing beam.

Carlo↗

DOE OSTI2016

A Parametric Study of the Adjoint Flux Calculation for TDKENO

TDKENO is a time-dependent transport code designed to simulate reactor transients. One application is to study experiments performed at the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory (INL). TDKENO solves the time-dependent, three-dimensional Boltzmann transport equation with explicit representation of delayed neutrons. Instead of directly integrating this equation, the Improved Quasi Static (IQS) method is employed. This involves factoring the neutron flux into two components - a rapidly varying (in time) amplitude equation and a slowly varying shape equation. These are solved separately on different time scales. The shape equation is solved using the Monte Carlo transport code KENO, from Oak Ridge National Laboratory's SCALE code package. Using the Monte Carlo method to solve the shape equation is computationally intensive, but the operation is only performed when needed - namely when the reactor system is undergoing a significant change. The amplitude equation is solved deterministically. This operation contains the time dependence and can be calculated relatively quickly, resulting in computational savings. TDKENO begins by solving the adjoint form of the transport equation for use as a weighting function. It uses this adjoint flux for multiple aspects of the simulation, including the calculation of all of the point kinetics parameters . The flux shape is found by solving a modified form of the transport equation in 'forward' mode. This paper examines the effect the adjoint weighting function has on the solution. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

NASA NTRS2004

Role of shielding in modulating the effects of solar particle events: Monte Carlo calculation of absorbed dose and DNA complex lesions in different organs

Distributions of absorbed dose and DNA clustered damage yields in various organs and tissues following the October 1989 solar particle event (SPE) were calculated by coupling the FLUKA Monte Carlo transport code with two anthropomorphic phantoms (a mathematical model and a voxel model), with the main aim of quantifying the role of the shielding features in modulating organ doses. The phantoms, which were assumed to be in deep space, were inserted into a shielding box of variable thickness and material and were irradiated with the proton spectra of the October 1989 event. Average numbers of DNA lesions per cell in different organs were calculated by adopting a technique already tested in previous works, consisting of integrating into "condensed-history" Monte Carlo transport codes--such as FLUKA--yields of radiobiological damage, either calculated with "event-by-event" track structure simulations, or taken from experimental works available in the literature. More specifically, the yields of "Complex Lesions" (or "CL", defined and calculated as a clustered DNA damage in a previous work) per unit dose and DNA mass (CL Gy-1 Da-1) due to the various beam components, including those derived from nuclear interactions with the shielding and the human body, were integrated in FLUKA. This provided spatial distributions of CL/cell yields in different organs, as well as distributions of absorbed doses. The contributions of primary protons and secondary hadrons were calculated separately, and the simulations were repeated for values of Al shielding thickness ranging between 1 and 20 g/cm2. Slight differences were found between the two phantom types. Skin and eye lenses were found to receive larger doses with respect to internal organs; however, shielding was more effective for skin and lenses. Secondary particles arising from nuclear interactions were found to have a minor role, although their relative contribution was found to be larger for the Complex Lesions than for the absorbed dose, due to their higher LET and thus higher biological effectiveness. c2004 COSPAR. Published by Elsevier Ltd. All rights reserved.

Non-NASA Center↗

DOE OSTI2016

Implementation of Material and Geometry Criticality Search in RMC Code

RMC is a continuous-energy Reactor Monte Carlo neutron and photon transport code being developed by Department of Engineering Physics at Tsinghua University, Beijing. For the use of simulation of reactor control or design of reactor with unknown material and geometry parameters, an accurate and efficient method of eigenvalue search is implemented in RMC code. The eigenvalue search focuses on two types of parameters. One is nuclide atom density, called material search, which includes the fuel density, enrichment or boron concentration. Perturbation theory of Monte Carlo Code, like differential operator method and adjoint-weighted perturbation method has been widely researched in sensitivity and uncertainty study. For material search, the perturbation algorithm needs to coupled with flexible material control block and optimized strategy. The other type of parameter is surface position, called geometry search, which includes fuel rod diameter, pitch and control rod position. MC21 iterated on surface position based on bisection method or regula-falsi method, and directly recalculated k-eff until it converged to target error interval. In this paper, the differential operator method, improved Newton-Raphson method is introduced. Based on these methods, the global material search is implemented. Then based on material search and finite method, global geometry search is implement. Finally, fuel enrichment, boron concentration and control rod position in normal PWR assembly are searched. The results show that RMC can search material and geometry parameters accurately and efficiently. In this paper, global eigenvalue search of material and geometry parameters have been implemented in RMC code. They are based on the differential operator method in Monte Carlo perturbation theory. The differential operator method samples δk/δρ{sub r}, δ{sup 2}k/δρ{sub r}{sup 2} along with k-eff estimator. Therefore it won't take up too much additional computational resource. For homogeneous material search, fission source effect has little effect of accuracy and can be ignored. The results of fuel enrichment and boron concentration searches in assembly show that improved Newton-Raphson method can iterate efficiently until k-eff is in the error interval. Based on material search and finite difference formula, the geometry parameter, like control rod axial position, can iterate until k-eff is in the error interval. However, the tiny perturbed region may not contain enough tracks to simulate. Besides, the large number density's relative change up to -1 may cause underestimation of Δk-eff, and then slope of k-x curve. This will cause non-convergence for Newton- Raphson method. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2014

Measurement and Analysis of Dose Rate Distribution in the Seismic Base-Isolation pit of the Electron Linac Oncology Facility - Paper 87

Recently electric linac oncology facilities incorporating seismic base-isolation structure have become very popular in Japan to prevent strong impact to the equipment by the earthquakes. When the boundary of the controlled area is set below the base-isolated foundations, additional shielding of iron plate is needed and the cost for shielding sometimes increases up to $1 M Since no one usually accesses underneath the treatment room, except short maintenance period for base-isolation structure, it is reasonable to set the boundary around the downstairs room (pit) by fence. However, it is very difficult to estimate the dose rate of new boundary by using the simple shielding calculation method, because of rotating gantry head and effect of scattering radiation in the base-isolation floor. In this study detail calculations by using Monte-Carlo calculation code MCNP5 have been performed to find out the practical shielding design. To be more specific, investigation for the complicated behavior of photons in the 'pit' have been performed by the analysis, i.e. (1) How much can we reduce the thickness of additional iron plate?, (2) Which angle of the beam direction will influence the most to the dose rate at the new boundary?, (3) Which component of the radiation is dominant at the new boundary? We also verified the calculated results by the measurement. We have performed the analysis of the behavior of photons in the pit and found out no additional iron shielding was necessary, which carried out an extreme reduction of costs. The main contribution to the dose rate at the boundary was scattered photons in the case of 0- degree beam direction. Good agreement between calculated and measured results were obtained, and passed the examination by the authority. It is concluded that the Monte-Carlo calculation code MCNP5 is very effective for the optimistic design of the electron linac oncology facilities incorporating seismic base-isolation structure. (authors)

43 PARTICLE ACCELERATORS↗

DOE OSTI2017

Transport modeling of convection dominated helicon discharges in Proto-MPEX with the B2.5-Eirene code

Data-constrained interpretative analyses of plasma transport in convection dominated helicon discharges in the Proto-MPEX linear device, and predictive calculations with additional Electron Cyclotron Heating/Electron Bernstein Wave (ECH/EBW) heating, are reported. The B2.5-Eirene code, in which the multi-fluid plasma code B2.5 is coupled to the kinetic Monte Carlo neutrals code Eirene, is used to fit double Langmuir probe measurements and fast camera data in front of a stainless-steel target. The absorbed helicon and ECH power (11 kW) and spatially constant anomalous transport coefficients that are deduced from fitting of the probe and optical data are additionally used for predictive simulations of complete axial distributions of the densities, temperatures, plasma flow velocities, particle and energy fluxes, and possible effects of alternate fueling and pumping scenarios. The somewhat hollow electron density and temperature radial profiles from the probe data suggest that Trivelpiece-Gould wave absorption is the dominant helicon electron heating source in the discharges analyzed here. There is no external ion heating, but the corresponding calculated ion temperature radial profile is not hollow. Rather it reflects ion heating by the electron-ion equilibration terms in the energy balance equations and ion radial transport resulting from the hollow density profile. With the absorbed power and the transport model deduced from fitting the sheath limited discharge data, calculated conduction limited higher recycling conditions were produced by reducing the pumping and increasing the gas fueling rate, resulting in an approximate doubling of the target ion flux and reduction of the target heat flux.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

NASA NTRS2020

Revised and extended benchmark results for Rayleigh scattering of sunlight in spherical atmospheres

While most of traditional Earth-atmosphere satellite remote sensing relies on radiative transfer in the plane parallel geometry, effects of sphericity are important at high sun and view zenith angles. Broad understanding of these effects is limited and, contrary to the plane-parallel case, finding accurate numerical results to test spherical RT codes is not easy. This paper aims to partially fill in this gap. Using the full-spherical RT code MYSTIC (Monte Carlo), and the plane-parallel RT code VLIDORT (discrete ordinates) corrected for atmospheric sphericity in the single and multiple scattering, we reproduced with better accuracy and extended the benchmark results by Adams & Kattawar [1978].

spherical atmospheres↗

DOE OSTI2016

Continuous Energy Photon Transport Implementation in MCATK

The Monte Carlo Application ToolKit (MCATK) code development team has implemented Monte Carlo photon transport into the MCATK software suite. The current particle transport capabilities in MCATK, which process the tracking and collision physics, have been extended to enable tracking of photons using the same continuous energy approximation. We describe the four photoatomic processes implemented, which are coherent scattering, incoherent scattering, pair-production, and photoelectric absorption. The accompanying background, implementation, and verification of these processes will be presented.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2013

On the Use of SRIM for Computing Radiation Damage Exposure

The SRIM (formerly TRIM) Monte Carlo simulation code is widely used to compute a number of parameters relevant to ion beam implantation and ion beam processing of materials. It also has the capability to compute a common radiation damage exposure unit known as atomic displacements per atom (dpa). Since dpa is a standard measure of primary radiation damage production, most researchers who employ ion beams as a tool for inducing radiation damage in materials use SRIM to determine the dpa associated with their irradiations. The use of SRIM for this purpose has been evaluated and comparisons have been made with an internationally-recognized standard definition of dpa, as well as more detailed atomistic simulations of atomic displacement cascades. Differences between the standard and SRIM-based dpa are discussed and recommendations for future usage of SRIM in radiation damage studies are made. In particular, it is recommended that when direct comparisons between ion and neutron data are intended, the Kinchin-Pease option of SRIM should be selected.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗