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At least 505 records · Page 28
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Uncertainty Analysis for a Criticality Benchmark [Slides]

This presentation discusses how KRUSTY was analyzed with LANL’s MCNP neutron transport code, Monte Carlo N-Particle. In the Monte Carlo method, neutrons are born in fission events and fly around the geometry having events (e.g., scatter, capture, fission, escape) according to the natural probabilities. “Natural probabilities” are given as neutron cross sections , which are measured and evaluated nuclear data. One of the goals of the benchmark project is to improve the nuclear cross sections (and the codes that use them). In this work, the response of interest is called k eff , which describes the neutron multiplication in a near critical system. Like most quantities, k eff is not measured directly; it is only inferred. In summation, MCNP results for k eff matched the measurements extremely well. The evaluated uncertainties are asymmetric and smaller than 0.00100 (e.g., + 0.00080 / – 0.00062). The largest uncertainties are due to alignment.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Implementation of Windowed Multipole into MCNP6.3

This work details the implementation of the Windowed MultiPole(WMP) method of generating cross sections into the monte carlo code MCNP6.3, the verification and validation performed, and potential future improvements to be implemented. Further, this work contextualizes WMP within the broader changes to how nuclear data is stored, iterated upon, and used that have been developing over the past decade.

97 MATHEMATICS AND COMPUTING↗
NASA NTRS2017

3DHZETRN: Inhomogeneous Geometry Issues

Historical methods for assessing radiation exposure inside complicated geometries for space applications were limited by computational constraints and lack of knowledge associated with nuclear processes occurring over a broad range of particles and energies. Various methods were developed and utilized to simplify geometric representations and enable coupling with simplified but efficient particle transport codes. Recent transport code development efforts, leading to 3DHZETRN, now enable such approximate methods to be carefully assessed to determine if past exposure analyses and validation efforts based on those approximate methods need to be revisited. In this work, historical methods of representing inhomogeneous spacecraft geometry for radiation protection analysis are first reviewed. Two inhomogeneous geometry cases, previously studied with 3DHZETRN and Monte Carlo codes, are considered with various levels of geometric approximation. Fluence, dose, and dose equivalent values are computed in all cases and compared. It is found that although these historical geometry approximations can induce large errors in neutron fluences up to 100 MeV, errors on dose and dose equivalent are modest (<10%) for the cases studied here.

Wilson, John W.↗
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↗
DOE OSTI2020

Preliminary design of control rods in the single-fluid double-zone thorium molten salt reactor (SD-TMSR)

Recent studies on Molten Salt Reactors (MSRs) showed that the excess reactivity at the beginning of the operation is large for many fueling strategies and must be compensated by a reactivity control system. The current work introduces a reliable safety system based on control rods in addition to the online feed system reactivity control in the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR). Three different initial fissile loadings are considered: 233 U, reactor-grade Pu, and transuranic (TRU) elements as a startup fuel. We applied six different absorbing materials to investigate the main operational and safety parameters in the SD-TMSR: natural B 4 C, enriched B 4 C with 90% 10 B, HfB 2 , HfH 1.62 , Eu 2 O 3 , and Gd 2 O 3 . The present work focuses on control rod design, integral and differential control rod worth, shutdown margin, and shadowing effects at steady-state. We employed the SERPENT-2 Monte-Carlo code to calculate the reactivity worth and analyze the performance of the reactivity control system. We showed that 233 U and reactor-grade Pu startup cores maintain adequate shutdown margin with all considered absorbers. Finally, this report proposes a design of control rod clusters that compensate the excess reactivity of the SD-TMSR loaded with different initial fissile material.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Preliminary Exploratory Study of Different Phase II Collimators

The Large Hadron Collider (LHC) collimation system is installed and commissioned in different phases, following the natural evolution of the LHC performance. To improve cleaning efficiency towards the end of the low beta squeeze at 7TeV, and in stable physics conditions, it is foreseen to complement the 30 highly robust Phase I secondary collimators with low impedance Phase II collimators. At this stage, their design is not yet finalized. Possible options include metallic collimators, graphite jaws with a movable metallic foil, or collimators with metallic rotating jaws. As part of the evaluation of the different designs, the FLUKA Monte Carlo code is extensively used for calculating energy deposition and studying material damage and activation. This report outlines the simulation approach and defines the critical quantities involved.

43 PARTICLE ACCELERATORS↗
DOE OSTI2013

Calculating infinite-medium {alpha}-eigenvalue spectra with a transition rate matrix method

The time-dependent behavior of the energy spectrum in neutron transport was investigated with a formulation, based on continuous-time Markov processes, for computing {alpha}-eigenvalues and eigenvectors in an infinite medium. For this, a research Monte Carlo code called TORTE was created and used to estimate elements of a transition rate matrix. TORTE is capable of using both multigroup and continuous-energy nuclear data, and verification was performed. Eigenvalue spectra for infinite homogeneous mixtures were obtained and an eigenfunction expansion was used to investigate transient behavior of the neutron energy spectrum. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2015

Rod-by-rod detailed Monte Carlo method of Cofrentes NPP as a commercial core critical - 14482

The detailed rod-by-rod three dimensional modeling of large cores such as the commercial reactors is now possible. The simultaneous improvement in fuel depletion systems based on Monte Carlo codes allows using the criticality of first and reload cores as critical experiments, usually defined as commercial core criticals (CCC). The interest of such critical configurations for reload cores is evident for the validation of codes and cross-sections used for burn-up credit. Nevertheless other areas of the fuel management can benefit from this capabilities since the realistic rod-by-rod detailed modelling of fuel assemblies can provide valuable help in the optimization of reload patterns and in the licensing of other production-oriented calculation tools In this presentation the results obtained in reactivity and power distribution in the analysis of Cofrentes NPP first criticality and of a low-burnup xenon-equilibrium state-point and its comparison with in-core instrumentation measurements are presented. The plan for development of a local rod-by-rod Monte Carlo fuel depletion capability is presented as well. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

The small-angle neutron scattering extension in MCNPX and the SANS cross section for nanodiamonds

Detonation nanodiamonds provide an intriguing prospect for the development of a very cold neutron (VCN) source or as a high albedo material for VCN reflection helping to control leakage of cold and very cold neutrons in a moderator. The investigation of source and reflector designs necessitates the development of new simulation tools for existing Monte Carlo codes that include small angle neutron scattering (SANS). The SANS extension in MCNPX 2.7 provides an analytical model of hard-sphere scattering with variable scattering particle sizes, size distributions, and packing fractions in order to supplement currently existing scattering kernels as well as the capability to read an experimentally determined small angle neutron scattering profile from a data file. The SANS extension is discussed and compared to nanodiamond SANS measurements.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2012

Preliminary design of ultra-long cycle fast reactor employing breed-and-burn strategy

A new design of ultra-long cycle fast reactor with power rate of 1000 MWe (UCFR) has been developed based on the strategy of breed-and burn. The bottom region of the core with low enriched uranium (LEU) plays a role of igniter of the core burning and the upper natural uranium (NU) region acts as blanket for breeding. Fissile materials are bred in the blanket and the active core moves upward at a speed of 5.4 cm/year. Through the core depletion calculation using Monte Carlo code, McCARD, it is confirmed that a full power operation of 60 years without refueling is feasible. Core performance characteristics have been evaluated in terms of axial/radial power shapes, reactivity feedback coefficients, etc. This design will serve as a base model for further design study of UCFRs using LWR spent fuels in the blanket region. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Benchmarking ENDF/B-VII.1, JENDL-4.0 and JEFF-3.1

Three nuclear data libraries have been tested extensively using criticality safety benchmark calculations. The three libraries are the new release of the US library ENDF/B-VII.1 (2011), the new release of the Japanese library JENDL-4.0 (2011), and the OECD/NEA library JEFF-3.1 (2006). All calculations were performed with the continuous-energy Monte Carlo code MCNP (version 4C3, as well as version 6-beta1). Around 2000 benchmark cases from the International Handbook of Criticality Safety Benchmark Experiments (ICSBEP) were used. The results were analyzed per ICSBEP category, and per element. Overall, the three libraries show similar performance on most criticality safety benchmarks. The largest differences are probably caused by elements such as Be, C, Fe, Zr, W. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

MT71x: Multi-Temperature Library Based on ENDF/B-VII.1

The Nuclear Data Team has released a multitemperature transport library, MT71x, based upon ENDF/B-VII.1 with a few modifications as well as additional evaluations for a total of 427 isotope tables. The library was processed using NJOY2012.39 into 23 temperatures. MT71x consists of two sub-libraries; MT71xMG for multigroup energy representation data and MT71xCE for continuous energy representation data. These sub-libraries are suitable for deterministic transport and Monte Carlo transport applications, respectively. The SZAs used are the same for the two sub-libraries; that is, the same SZA can be used for both libraries. This makes comparisons between the two libraries and between deterministic and Monte Carlo codes straightforward. Both the multigroup energy and continuous energy libraries were verified and validated with our checking codes checkmg and checkace (multigroup and continuous energy, respectively) Then an expanded suite of tests was used for additional verification and, finally, verified using an extensive suite of critical benchmark models. We feel that this library is suitable for all calculations and is particularly useful for calculations sensitive to temperature effects.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Radiation Shielding Analysis of Linac Extension Area in APS Using FLUKA and PHITS Monte Carlo Radiation Transport Codes

The Linac extension area tunnel can accept 300- to 500-MeV electrons for experiments. Radiations from various normal and accident beam loss scenarios were analyzed with the FLUKA and PHITS Monte Carlo radiation transport codes. In normal beam loss scenarios, the beam was dumped on the horizontal and vertical beam dumps, while the accident scenarios were horizontal and vertical mis-steered beams that can occur due to incorrect beam energy and magnet settings. The extent of beam deviation, which was established using particle tracking studies, was used to define the beam directions in the Monte Carlo transport calculations. Further, a mis-steered condition, where the beam was not intercepted by the dumps, was also analyzed. For the horizontal dump, the PHITS results above the berm were higher for neutrons and somewhat lower for photons, compared to the FLUKA results. Comparison of the secondary particle yields from FLUKA and PHITS from an Al block with dimensions comparable to what was used in the beam dumps showed a 40% higher neutron yield and a 10% lower photon yield. Finally, to ensure compliance with Argonne National Laboratory's shielding policy, the design of the horizontal beam dump was modified, and additional shielding was recommended.

FLUKA↗
DOE OSTI2016

SU-F-T-153: Experimental Validation and Calculation Benchmark for a Commercial Monte Carlo Pencil Beam Scanning Proton Therapy Treatment Planning System in Water

Purpose: Eclipse proton Monte Carlo AcurosPT 13.7 was commissioned and experimentally validated for an IBA dedicated PBS nozzle in water. Topas 1.3 was used to isolate the cause of differences in output and penumbra between simulation and experiment. Methods: The spot profiles were measured in air at five locations using Lynx. PTW-34070 Bragg peak chamber (Freiburg, Germany) was used to collect the relative integral Bragg peak for 15 proton energies from 100 MeV to 225 MeV. The phase space parameters (σx, σθ, ρxθ) number of protons per MU, energy spread and calculated mean energy provided by AcurosPT were identically implemented into Topas. The absolute dose, profiles and field size factors measured using ionization chamber arrays were compared with both AcurosPT and Topas. Results: The beam spot size, σx, and the angular spread, σθ, in air were both energy-dependent: in particular, the spot size in air at isocentre ranged from 2.8 to 5.3 mm, and the angular spread ranged from 2.7 mrad to 6 mrad. The number of protons per MU increased from ∼9E7 at 100 MeV to ∼1.5E8 at 225 MeV. Both AcurosPT and TOPAS agree with experiment within 2 mm penumbra difference or 3% dose difference for scenarios including central axis depth dose and profiles at two depths in multi-spot square fields, from 40 to 200 mm, for all the investigated single-energy and multi-energy beams, indicating clinically acceptable source model and radiation transport algorithm in water. Conclusion: By comparing measured data and TOPAS simulation using the same source model, the AcurosPT 13.7 was validated in water within 2 mm penumbra difference or 3% dose difference. Benchmarks versus an independent Monte Carlo code are recommended to study the agreement in output, filed size factors and penumbra differences. This project is partially supported by the Varian grant under the master agreement between University of Pennsylvania and Varian.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2014

Collisionless electron heating in periodic arrays of inductively coupled plasmas

A novel mechanism of collisionless heating in large planar arrays of small inductive coils operated at radio frequencies is presented. In contrast to the well-known case of non-local heating related to the transversal conductivity, when the electrons move perpendicular to the planar coil, we investigate the problem of electrons moving in a plane parallel to the coils. Two types of periodic structures are studied. Resonance velocities where heating is efficient are calculated analytically by solving the Vlasov equation. Certain scaling parameters are identified. The concept is further investigated by a single particle simulation based on the ergodic principle and combined with a Monte Carlo code allowing for collisions with Argon atoms. Resonances, energy exchange, and distribution functions are obtained. The analytical results are confirmed by the numerical simulation. Pressure and electric field dependences are studied. Stochastic heating is found to be most efficient when the electron mean free path exceeds the size of a single coil cell. Then the mean energy increases approximately exponentially with the electric field amplitude.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Beam-machine Interaction at the CERN LHC

The radiation field generated by a high energy and intensity accelerator is of concern in terms of element functionality threat, component damage, electronics reliability, and material activation, but also provides signatures that allow actual operating conditions to be monitored. The shower initiated by an energetic hadron involves many different physical processes, down to slow neutron interactions and fragment de-excitation, which need to be accurately described for design purposes and to interpret operation events. The experience with the transport and interaction Monte Carlo code FLUKA at the Large Hadron Collider (LHC), operating at CERN with 4 TeV proton beams (and equivalent magnetic rigidity Pb beams) and approaching nominal luminosity and energy, is presented. Design, operation and upgrade challenges are reviewed in the context of beam-machine interaction account and relevant benchmarking examples based on radiation monitor measurements are shown.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

MONK10: Burnup credit capability - 14596

MONK{sup R} is a Monte Carlo code for nuclear criticality and reactor physics analyses. It has a proven track record of application to the whole of the nuclear fuel cycle and is well established in the UK criticality community. Furthermore it is increasingly being used for reactor physics analysis (as described at ICNC 2011), which makes it an ideal tool for burn-up credit (BUC) calculations. Throughout the paper, example calculations based on a PWR are presented to illustrate the capabilities of the MONK10 code. In order to account for the spatial dependence of material burn-up it has in the past been necessary to design models with multiple regions and materials specifically to allow material burn-up to vary spatially. This is very labour intensive and difficult to change at a later stage. A new code version, MONK10, was released last year which includes the facility to allow a burn-up (BU) mesh to be superimposed on an existing model in order to account for the spatial dependence of the burn-up. This facility is used to consider the effect of radial position of a fuel element in a PWR core on BUC. Additionally, a thermal hydraulics (TH) mesh can be used to specify region dependent temperature. This, coupled with the fact that MONK10 also incorporates an on-the-fly Doppler broadening methodology facilitates the modelling of spatially dependent temperatures for the different components. A TH mesh is used to superimpose a temperature profile on a PWR based model and the effect of this on BUC is considered. The burn-up modelling in MONK has been benchmarked against the ANSWERS WIMS deterministic reactor physics code. Once the burn-up calculation has been completed and the depleted fuel compositions determined the spent fuel compositions can be transferred into a model of a storage facility or transport flask in order to perform burn-up credit analysis. The initial model is usually described as the donor model and the latter model as the receiver model. This transfer is carried out using the COWL option which allows the specification of a material in the receiver model based on the material's composition in a given BU mesh cell from the donor model. This allows compositions and densities to be transferred and also allows user specified adjustments to be made. For example, this could include omitting the fission products in order to estimate their contribution to burn-up credit and provide an actinide-only analysis. The effect of excluding appropriate nuclides is presented. An example of how the ANSWERS SPRUCE code can be used to quantify uncertainty in a BUC calculation is also presented. (authors)

42 ENGINEERING↗