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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 379 records · Page 21
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Personalized Assessment of kV Cone Beam Computed Tomography Doses in Image-guided Radiotherapy of Pediatric Cancer Patients

Purpose: To develop a quantitative method for the estimation of kV cone beam computed tomography (kVCBCT) doses in pediatric patients undergoing image-guided radiotherapy. Methods and Materials: Forty-two children were retrospectively analyzed in subgroups of different scanned regions: one group in the head-and-neck and the other group in the pelvis. Critical structures in planning CT images were delineated on an Eclipse treatment planning system before being converted into CT phantoms for Monte Carlo simulations. A benchmarked EGS4 Monte Carlo code was used to calculate three-dimensional dose distributions of kVCBCT scans with full-fan high-quality head or half-fan pelvis protocols predefined by the manufacturer. Based on planning CT images and structures exported in DICOM RT format, occipital-frontal circumferences (OFC) were calculated for head-and-neck patients using DICOMan software. Similarly, hip circumferences (HIP) were acquired for the pelvic group. Correlations between mean organ doses and age, weight, OFC, and HIP values were analyzed with SigmaPlot software suite, where regression performances were analyzed with relative dose differences (RDD) and coefficients of determination (R{sup 2}). Results: kVCBCT-contributed mean doses to all critical structures decreased monotonically with studied parameters, with a steeper decrease in the pelvis than in the head. Empirical functions have been developed for a dose estimation of the major organs at risk in the head and pelvis, respectively. If evaluated with physical parameters other than age, a mean RDD of up to 7.9% was observed for all the structures in our population of 42 patients. Conclusions: kVCBCT doses are highly correlated with patient size. According to this study, weight can be used as a primary index for dose assessment in both head and pelvis scans, while OFC and HIP may serve as secondary indices for dose estimation in corresponding regions. With the proposed empirical functions, it is possible to perform an individualized quantitative dose assessment of kVCBCT scans.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2014

SU-D-19A-03: Monte Carlo Investigation of the Mobetron to Perform Modulated Electron Beam Therapy

Purpose: Modulated electron radiotherapy (MERT) has been proposed as a mean of delivering conformal dose to shallow tumors while sparing distal structures and surrounding tissues. In intraoperative radiotherapy (IORT) utilizing Mobetron, an applicator is placed as closely as possible to the suspected cancerous tissues to be treated. In this study we investigate the characteristics of Mobetron electron beams collimated by an in-house prospective electron multileaf collimator (eMLC) and its feasibility for MERT. Methods: IntraOp Mobetron™ dedicated to perform radiotherapy during surgery was used in the study. It provides several energies (6, 9 and 12 MeV). Dosimetry measurements were performed to obtain percentage depth dose curves (PDD) and profiles for a 10-cm diameter applicator using the PTW MP3/XS 3D-scanning system and the semiflex ion chamber. MCBEAM/MCSIM Monte Carlo codes were used for the treatment head simulation and phantom dose calculation. The design of an electron beam collimation by an eMLC attached to the Mobetron head was also investigated using Monte Carlo simulations. Isodose distributions resulting from eMLC collimated beams were compared to that collimated using cutouts. The design for our Mobetron eMLC is based on our previous experiences with eMLCs designed for clinical linear accelerators. For Mobetron the eMLC is attached to the end of a spacer-mounted rectangular applicator at 50 cm SSD. Steel will be used as the leaf material because other materials would be toxic and will not be suitable for intraoperative applications. Results: Good agreement (within 2%) was achieved between measured and calculated PDD curves and profiles for all available energies. Dose distributiosn provided by the eMLC showed reasonable agreement (∼3%/1mm) with those obtained by conventional cutouts. Conclusion: Monte Carlo simulations are capable of modeling Mobetron electron beams with a reliable accuracy. An eMLC attached to the Mobteron treatment head will allow better treatment options with those machines.

60 APPLIED LIFE SCIENCES↗
NASA NTRS2004

High Altitude Radiations Relevant to the High Speed Civil Transport (HSCT)

The Langley Research Center (LaRC) performed atmospheric radiation studies under the SST development program in which important ionizing radiation components were measured and extended by calculations to develop the existing atmospheric ionizing radiation (AIR) model. In that program the measured neutron spectrum was limited to less than 10 MeV by the available 1960-1970 instrumentation. Extension of the neutron spectrum to high energies was made using the LaRC PROPER-3C monte carlo code. It was found that the atmospheric neutrons contributed about half of the dose equivalent and approximately half of the neutron contribution was from high energy neutrons above 10 MeV. Furthermore, monte carlo calculations of solar particle events showed that potential exposures as large as 10-100 mSv/hr may occur on important high latitude routes but acceptable levels of exposure could be obtained if timely descent to subsonic altitudes could be made. The principal concern was for pregnant occupants onboard the aircraft. As a result of these studies the FAA Advisory Committee on the Radiobiological Aspects of the SST recommended: 1. Crew members will have to be informed of their exposure levels 2. Maximum exposures on any flight to be limited to 5 mSv 3. Airborne radiation detection devices for total exposure and exposure rates 4. Satellite monitoring system to provide SST aircraft real-time information on atmospheric radiation levels for exposure mitigation 5. A solar forecasting system to warn flight operations of an impending solar event for flight scheduling and alert status. These recommendations are a reasonable starting point to requirements for the HSCT with some modification reflecting new standards of protection as a result of changing risk coefficients.

Wilson, J. W.↗
DOE OSTI2012

Development of 3D full-core ERANOS-2.2/MCNPX-2.7.0 models and neutronic analysis of the BFS-2 zero-power facility

The present paper is addressing the development and validation against experimental data of 3D full-core models of the BFS-2 zero-power fast-reactor using both the deterministic system code ERANOS-2.2 and the stochastic code MCNPX-2.7.0. The model configuration of BFS considered for analysis is the BFS-62-3A benchmark. To extend the - deterministic/stochastic - code-to-code comparison, neutronic parameters, i.e. reactivity, neutron spectrum and reaction rates, were also simulated at the cell level with the Monte Carlo code SERPENT-1.1.7 with two modern data libraries, ENDF-B/VII and JEFF-3.1.1. The BFS-2 critical zero-power facility at the Inst. of Physics and Power Engineering (IPPE) was designed for simulations of the core and shielding of sodium-cooled, fast reactors, for neutron data validation and comparison with experimental results. At the BFS-2 facility, the BFS-62-3A critical benchmark experiment was set-up as a mock-up of the BN-600 reactor, with hybrid MOX fuel and stainless steel reflectors. A UO{sub 2} blanket and a large non-homogeneous stainless-steel reflector surround the core. The lattice is hexagonal of pitch 5.1 cm and metallic dowels are used to keep in central position cylindrical rods made of different types of material (fissile, fertile, blanket, plenum, shielding and absorber). A typical subassembly is formed in piling up various pellets of about 1 cm in height and 4.6 cm in diameter, conferring large heterogeneity in the axial direction. The full-core model development was a complex task due to the large number of subassemblies and the axial subassembly heterogeneity. In ERANOS-2.2, it was necessary to homogenize axially per region the pellets used to form the subassembly. The self-shielded macroscopic cross-sections were calculated using the cell code ECCO in association with JEFF-3.1 and ENDF/B-VI.8 data libraries. The core calculations were performed with broad cross-sections data in 33 neutron energy groups with the solver AVNM in the diffusion approximation, mostly. In MCNPX-2.7.0, a step-by-step approach was used, starting with a model in which the fissile rods were simulated on a homogeneous level, to finally integrate the actual heterogeneous description of the subassemblies. The code-to-code cell analysis performed between ECCO, SERPENT and MCNPX with different modern nuclear data library revealed that the results for the infinite multiplication factor between Monte Carlo and deterministic analysis are in good agreement ({Delta}p < 100 pcm). The differences between the results were observed to be larger for the neutron data libraries, with reactivity differences up to 350 pcm. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

High conversion Th-U{sup 233} fuel assembly for current generation of PWRs

This paper presents a preliminary design of a high conversion Th-U{sup 233} fuel assembly applicable for current generation of Pressurized Water Reactor (PWRs). The considered fuel assembly has a typical 17 x 17 PWR lattice. However in order to increase the conversion of Th{sup 232} to U{sup 233}, the assembly was subdivided into the two regions called seed and blanket. The central seed region has a higher than blanket U{sup 233} content and acts as a neutron source for the peripheral blanket region. The latest acts as a U{sup 233} breeder. While the seed fuel pins have a standard dimensions the blanket fuel radius was increased in order to reduce the moderation and to facilitate the resonance neutron absorption in blanket Th{sup 232}. The U{sup 233} content in the seed and blanket regions was optimized to achieve maximal initial to discharged fissile inventory ratio (FIR) taking into account the target fuel cycle length of 12 months with 3-batch reloading scheme. In this study the neutronic calculations were performed on the fuel assembly level using Helios deterministic lattice transport code. The fuel cycle length and the core k{sub eff} were estimated by applying the Non Linear Reactivity Model. The applicability of the HELIOS code for the analysis of the Th-based high conversion designs was confirmed with the help of continuous-energy Monte-Carlo code SERPENT. The results of optimization studies show that for the heterogeneous seed and blanket (SB) fuel assembly the FIR of about 0.95 can be achieved. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Full 3D visualization tool-kit for Monte Carlo and deterministic transport codes

We propose a package of tools capable of translating the geometric inputs and outputs of many Monte Carlo and deterministic radiation transport codes into open source file formats. These tools are aimed at bridging the gap between trusted, widely-used radiation analysis codes and very powerful, more recent and commonly used visualization software, thus supporting the design process and helping with shielding optimization. Three main lines of development were followed: mesh-based analysis of Monte Carlo codes, mesh-based analysis of deterministic codes and Monte Carlo surface meshing. The developed kit is considered a powerful and cost-effective tool in the computer-aided design for radiation transport code users of the nuclear world, and in particular in the fields of core design and radiation analysis. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Parallel Performance of the Time Dependent Transport Code TDKENO with Applications to TREAT Simulations

Renewed interest in high fidelity simulation of excursion events has prompted the improvement of codes that solve the time-dependent Boltzmann transport equation. Often these codes make approximations to the transport equation in order to achieve results in a reasonable amount of time. One method, the improved quasi-static (IQS), makes few approximations compared to adiabatic, diffusion, and point kinetics. This is the methodology employed by the code TDKENO. One downside to this approach is a portion of the calculation (flux shape) is done with a Monte Carlo code and must be called upon several times. To minimize run time the flux shape calculation is solved with a modified version of KENO-VI from Scale 6.2. This version of KENO runs in parallel across hundreds of nodes, thus improving performance. The parallel capabilities of KENO allows TDKENO to simulate complex transient experiments with a large number of histories. These improvements are applied to the simulation of TREAT calibration experiments. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2013

Development of the Code RITRACKS

A document discusses the code RITRACKS (Relativistic Ion Tracks), which was developed to simulate heavy ion track structure at the microscopic and nanoscopic scales. It is a Monte-Carlo code that simulates the production of radiolytic species in water, event-by-event, and which may be used to simulate tracks and also to calculate dose in targets and voxels of different sizes. The dose deposited by the radiation can be calculated in nanovolumes (voxels). RITRACKS allows simulation of radiation tracks without the need of extensive knowledge of computer programming or Monte-Carlo simulations. It is installed as a regular application on Windows systems. The main input parameters entered by the user are the type and energy of the ion, the length and size of the irradiated volume, the number of ions impacting the volume, and the number of histories. The simulation can be started after the input parameters are entered in the GUI. The number of each kind of interactions for each track is shown in the result details window. The tracks can be visualized in 3D after the simulation is complete. It is also possible to see the time evolution of the tracks and zoom on specific parts of the tracks. The software RITRACKS can be very useful for radiation scientists to investigate various problems in the fields of radiation physics, radiation chemistry, and radiation biology. For example, it can be used to simulate electron ejection experiments (radiation physics).

Plante, Ianik↗
DOE OSTI2015

Flattening of fuel importance for heterogeneous system to estimate maximal multiplication factor - 14595

In order to quantify risks of re-criticality incidents associated with loss of coolant accidents (LOCA), the maximum reactivity insertion by relocation of damaged fuel should be quantified. As the first step for the quantification, an automated search scheme is studied for an optimal geometry of damaged fuel in the coolant channel where the neutron multiplication factor k{sub eff} takes a maximal value. The scheme is based on the flattening of the fuel importance. The importance can be recognized as the substitution reactivity from fuel to water in heterogeneous systems. The flattening is achieved by substitution of water and the fuel so as to raise the k{sub eff} referring the local substitution reactivities. To calculate the reactivities efficiently, the first order perturbation scheme is adopted. In the scheme, the adjoint flux is calculated with the PARTISN code. The multi group constants used for the calculation and the perturbed reaction rates are estimated with the continuous energy Monte Carlo code MCNP-5. The proposed methods are tested for a quarter pin cell model that is constructed using Cartesian mesh. From a random distribution of a given amount of fragmented fuel, an optimal geometry is found where the fragmented fuels are aggregated. The multiplication factor in the geometry is greater than that in the original random distribution by 1.4%Δk/k. That indicates the significance of this study. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Scrape-off layer plasma and neutral characteristics and their interactions with walls for FNSF

Simulations of the heat flux on plasma facing components from exhausting core plasma are reported for two possible Fusion Nuclear Science Facility (FNSF) divertor configurations. One configuration utilizes divertor plates strongly inclined with respect to the poloidal magnetic flux surfaces like that planned for ITER and results in a partially detached divertor-plasma. The second configuration has divertor plates orthogonal to the flux surfaces, which leads to a fully detached divertor-plasma if the width of the divertor region is sufficient. Both configurations use scrape-off layer impurity seeding to yield an acceptable peak heat flux of ~10 MW/m 2 or smaller on the divertor plates and chamber walls. The roles of recycled hydrogenic atoms and molecules are investigated and distribution of sputtering tungsten throughout the edge region modeled. Furthermore, the simulations are performed with the UEDGE 2D transport code to model both plasma and neutral components with supplementary neutral modeling performed with the DEGAS 2 Monte Carlo code.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Development of nuclear data processing code FRENDY version 2

Nuclear data processing is an important interface between an evaluated nuclear data library and neutronics calculation codes. JAEA has been developed the new nuclear data processing code FRENDY from 2013. FRENDY version 1 generates the ACE files used for the continuous-energy Monte Carlo codes including PHITS, Solomon, Serpent,and MCNP and it was released as the open-source software under the 2-clause BSD license in 2019.After we released FRENDY version 1, many functions, e.g., the multi-group neutron cross-section library generation, the statistical uncertainty quantification of the probability tables for unresolved resonance cross-section, the 'perturbation' of the ACE file for uncertainty quantification, and the modification of the ENDF-6 formatted nuclear data file, have been developed. We released FRENDY version 2 including these functions. The present paper gives an overview of FRENDY and presents features of new functions implemented in FRENDY version 2. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2011

Comparison of DSMC and CFD Solutions of Fire II Including Radiative Heating

The ability to compute rarefied, ionized hypersonic flows is becoming more important as missions such as Earth reentry, landing high mass payloads on Mars, and the exploration of the outer planets and their satellites are being considered. These flows may also contain significant radiative heating. To prepare for these missions, NASA is developing the capability to simulate rarefied, ionized flows and to then calculate the resulting radiative heating to the vehicle's surface. In this study, the DSMC codes DAC and DS2V are used to obtain charge-neutral ionization solutions. NASA s direct simulation Monte Carlo code DAC is currently being updated to include the ability to simulate charge-neutral ionized flows, take advantage of the recently introduced Quantum-Kinetic chemistry model, and to include electronic energy levels as an additional internal energy mode. The Fire II flight test is used in this study to assess these new capabilities. The 1634 second data point was chosen for comparisons to be made in order to include comparisons to computational fluid dynamics solutions. The Knudsen number at this point in time is such that the DSMC simulations are still tractable and the CFD computations are at the edge of what is considered valid. It is shown that there can be quite a bit of variability in the vibrational temperature inferred from DSMC solutions and that, from how radiative heating is computed, the electronic temperature is much better suited for radiative calculations. To include the radiative portion of heating, the flow-field solutions are post-processed by the non-equilibrium radiation code HARA. Acceptable agreement between CFD and DSMC flow field solutions is demonstrated and the progress of the updates to DAC, along with an appropriate radiative heating solution, are discussed. In addition, future plans to generate more high fidelity radiative heat transfer solutions are discussed.

Liechty, Derek S.↗
DOE OSTI2018

Application of adaptive multilevel splitting on coupled neutron-photon TRIPOLI-4{sup R} Monte Carlo simulations - 25427

In the context of radiation protection simulations, the Adaptive Multilevel Splitting (AMS) algorithm is a challenging variance reduction (VR) technique that has been recently investigated in the field of particle transport simulation. It has been implemented in the forthcoming version 11 of the Monte Carlo code TRIPOLI-4{sup R} and successfully tested in neutron-only and photon-only configurations. This paper addresses the application of the AMS algorithm to coupled simulations, and particularly to neutron-photon Monte Carlo calculations. The branching process occurring during the Monte Carlo coupled transport is taken into account in the new coupled-AMS algorithm and is explained in this paper. Two different neutron-photon configurations are then investigated, leading to a comparison of the coupled-AMS algorithm with the analog simulation on the one hand, and with the Exponential Transform (ET) on the other hand, which is the standard VR technique of TRIPOLI-4. Gains up to 30 are obtained in terms of Figure of Merit relatively to the analog simulation, which is about 4 to 6 times more efficient than the ET method for these configurations. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2017

Flux in a 1D-Reactor Cell by Response Matrix

Legacy computational reactor physics software tools and protocols currently used in support of the Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice. Thus, they are difficult, if not impossible, to verify and validate (V and V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of their application and protocols from the 1960's and 1970's is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity transport theory based computational software and models to replace the legacy methods, with appropriate (V and V) protocols according to applicable national standards. The new computational tools include: - The 2-dimensional HELIOS deterministic transport code for integrated ATR core fuel and experiment management and Core Safety Analysis Package (CSAP) physics data generation. The MCNP5 3-dimensional stochastic simulation (Monte Carlo) code for verification and validation applications and for specialized ATR core analysis along with its longstanding traditional role in experiment support - The 2-dimensional deterministic NEWT transport code and the 3-dimensional KENO6 stochastic simulation code, with their SCALE-based support infrastructure for use in V and V applications and as supporting modules for sensitivity-uncertainty analysis via the TSUNAMI system. The ATTILA 3-dimensional discrete-ordinates transport code for specialized experiment support and possibly additional applications. Regardless of application, every neutron transport code, new or legacy, requires verification. The aim of this presentation is to provide a neutron transport benchmark for a simple 1D-reactor cell in plane geometry. The benchmark finds use as a standard of comparison for the codes listed above intended for the ATR upgrade. The method developed here is in contrast to the finite Fourier transform solution found in Ref. 1, which is limited to isotropic scattering and particularly difficult to numerically implement unless the total cross section is constant over the cell.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

The MARS Code System User’s Guide Version 15(2016)

This paper is a user’s guide to the current version of the MARS15 Monte Carlo code. MARS allows inclusive and, in many cases, exclusive simulations of three-dimensional hadronic and electromagnetic cascades and modeling of heavy ion, muon and low energy neutron-photon transport in accelerator, detector, spacecraft and shielding components. Particles can have energies from a fraction of an electronvolt up to about 100 TeV. The code has undergone substantial improvements since the last documented version MARS13(95) and since previous releases of MARS15; all these as well as other specific features of the MARS code system are explained in detail. Descriptions of general input and output with commentary and recommendations are given. Examples are given for running the program with distributed sources, complex compounds, arbitrary geometries, and magnetic fields. A built-in graphical-user interface allows the visualization and detailed analysis of the geometry description. Enhanced options are described for the MARS code being coupled with: the DPMJET hadron event generator, the MCNP code for low-energy neutron transport, the ANSYS system for thermal and stress analysis, physics analysis and graphics packages, the MAD accelerator lattice description via a universal MAD-MARS beam line builder, the STRUCT program for tracking particles in accelerator lattices with beam loss recording, the DETRA code for nuclide decay and transmutation calculations, the EGS5 code for precise modeling of low-energy electromagnetic showers, and the ROOT system for description and visualization of arbitrary geometries. Use of the code in a multistage mode, coupled with event generators (DPMJET), with the STRUCT program for tracking particles in accelerator lattices with beam loss recording, and with physics analysis and graphics packages is demonstrated with typical input and output examples.

43 PARTICLE ACCELERATORS↗
DOE OSTI2016

New Version of the MCNP Analytic Criticality Benchmark Suite

Analytical benchmarks provide an invaluable tool for verifying computer codes used to simulate neutron transport. Several collections of analytical benchmark problems are used routinely in the verification of production Monte Carlo codes such as MCNP{sup R}. Verification of a computer code is a necessary prerequisite to the more complex validation process. The verification process confirms that a code performs its intended functions correctly. The validation process involves determining the absolute accuracy of code results vs. nature. In typical validations, results are computed for a set of benchmark experiments using a particular methodology (code, cross-section data with uncertainties, and modeling) and compared to the measured results from the set of benchmark experiments. The validation process determines bias, bias uncertainty, and possibly additional margins. Verification is generally performed by the code developers, while validation is generally performed by code users for a particular application space. The VERIFICATION-KEFF suite of criticality problems was originally a set of 75 criticality problems found in the literature for which exact analytical solutions are available. Even though the spatial and energy detail is necessarily limited in analytical benchmarks, typically to a few regions or energy groups, the exact solutions obtained can be used to verify that the basic algorithms, mathematics, and methods used in complex production codes perform correctly. The present work has focused on revisiting this benchmark suite. A thorough review of the problems resulted in discarding some of them as not suitable for MCNP benchmarking. For the remaining problems, many of them were reformulated to permit execution in either multigroup mode or in the normal continuous-energy mode for MCNP. Execution of the benchmarks in continuous-energy mode provides a significant advance to MCNP verification methods. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

L3:RTM.MCH.P19.06: Neutronics UQ

This report summarizes some of the work performed in the last year on the nuclear data uncertainty tasks using the multipole representation. This approach was proposed to provide a simple and integrated way of evaluating the nuclear data uncertainty in a Monte Carlo simulation and evaluating their impact on multigroup cross sections. As previously described, the windowed multipole format allows for a resonance based representation of the cross sections with direct Doppler broadening. This representation also lends itself nicely to the evaluation of sensitivities of resonance parameters and the direct perturbation of resonance parameters in the simulation environment. Both of these approaches rely on the need of conversion of the covariance data (File 32 of the ENDF library) into a pole and residues form: i.e. into a multipole covariance. Section 5 details the current pathways that have been used so far to achieve this. In order to verify the conversion process of the covariance data, the sensitivies to the resonance parameters and the sampling from these parameters, an analytical benchmark of the slowing process was developed. The benchmark provides a closed-form solution for the flux in the presence of Hydrogen and multiple resonant nuclides. Section 2 details the analytical benchmark and provides the closed-form solutions for the flux and its sensitivity to resonance parameters. A verification of the proposed methodology is performed and compared with the analytic solution to demonstrate the accuracy of the benchmark and the applicability of our embedded resonance perturbation approach in a small Monte Carlo code. Papers are currently being drafted documenting the benchmark and the generation of sensitivities. As paths to best generate multipole covariance matrices are being compared, we also explored new methods to propagate nuclear data uncertainty across the Monte Carlo neutron transport calculation. Section 4 studies the use of multipole covariance matrices to perform sensitivity analysis. Section 3.2 documents the results of our investigations into a new way of propagating nuclear data uncertainty across Monte Carlo transport simulations by sampling new epistemic cross sections for each new neutron.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Multigroup Cross-section Generation in MCNP6.3 [Slides]

This presentation states that in comparison to the NJOY-produced multigroup cross sections, the MCNP-produced multigroup cross sections are generally consistent. Statistical uncertainties, however, are challenging and the unresolved resonance region may be looked at in the future. It also discusses how the SPM and LCS options were compared to each other for internal consistency. Additionally, some reactor pin-cell-like problems were used to compare to multigroup capabilities in other Monte Carlo codes (e.g., Serpent, OpenMC).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗