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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 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

SU-E-T-561: Monte Carlo-Based Organ Dose Reconstruction Using Pre-Contoured Human Model for Hodgkins Lymphoma Patients Treated by Cobalt-60 External Beam Therapy

Purpose: Organ doses for the Hodgkin’s lymphoma patients treated with cobalt-60 radiation were estimated using an anthropomorphic model and Monte Carlo modeling. Methods: A cobalt-60 treatment unit modeled in the BEAMnrc Monte Carlo code was used to produce phase space data. The Monte Carlo simulation was verified with percent depth dose measurement in water at various field sizes. Radiation transport through the lung blocks were modeled by adjusting the weights of phase space data. We imported a precontoured adult female hybrid model and generated a treatment plan. The adjusted phase space data and the human model were imported to the XVMC Monte Carlo code for dose calculation. The organ mean doses were estimated and dose volume histograms were plotted. Results: The percent depth dose agreement between measurement and calculation in water phantom was within 2% for all field sizes. The mean organ doses of heart, left breast, right breast, and spleen for the selected case were 44.3, 24.1, 14.6 and 3.4 Gy, respectively with the midline prescription dose of 40.0 Gy. Conclusion: Organ doses were estimated for the patient group whose threedimensional images are not available. This development may open the door to more accurate dose reconstruction and estimates of uncertainties in secondary cancer risk for Hodgkin’s lymphoma patients. This work was partially supported by the intramural research program of the National Institutes of Health, National Cancer Institute, Division of Cancer Epidemiology and Genetics.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2013

Development and validation of ALEPH2 Monte Carlo burn-up code

The ALEPH2 Monte Carlo depletion code has two principal features that make it a flexible and powerful tool for reactor analysis. First of all, its comprehensive nuclear data library ensures the consistency between steady-state Monte Carlo and deterministic depletion modules. It covers neutron and proton induced reactions, neutron and proton fission product yields, spontaneous fission product yields, radioactive decay data and total recoverable energies per fission. Secondly, ALEPH2 uses an advanced numerical solver for the first order ordinary differential equations describing the isotope balances, namely a Radau IIA implicit Runge-Kutta method. The versatility of the code allows using it for time behavior simulation of various systems ranging from single pin model to full-scale reactor model. The code is extensively used for the neutronics design of the MYRRHA research fast spectrum facility which will operate in both critical and sub-critical modes. The code has been validated on the decay heat data from JOYO experimental fast reactor. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

TH-A-19A-05: Modeling Physics Properties and Biologic Effects Induced by Proton and Helium Ions

Purpose: Currently, proton and carbon ions are used for cancer treatment. More recently, other light ions including helium ions have shown interesting physical and biological properties. The purpose of this work is to study the biological and physical properties of helium ions (He-3) in comparison to protons. Methods: Monte Carlo simulations with FLUKA, GEANT4 and MCNPX were used to calculate proton and He-3 dose distributions in water phantoms. The energy spectra of proton and He-3 beams were calculated with high resolution for use in biological models. The repair-misrepairfixation (RMF) model was subsequently used to calculate the RBE. Results: The proton Bragg curve calculations show good agreement between the three general purpose Monte Carlo codes. In contrast, the He-3 Bragg curve calculations show disagreement (for the magnitude of the Bragg peak) between FLUKA and the other two Monte Carlo codes. The differences in the magnitude of the Bragg peak are mainly due to the discrepancy in the secondary fragmentation cross sections used by the codes. The RBE for V79 cell lines is about 0.96 and 0.98 at the entrance of proton and He-3 ions depth dose respectively. The RBE increases to 1.06 and 1.59 at the Bragg peak of proton and He-3 ions. The results demonstrated that LET, microdosimetric parameters (such as dose-mean lineal energy) and RBE are nearly constant along the plateau region of Bragg curve, while all parameters increase within the Bragg peak and at the distal edge for both proton and He-3 ions. Conclusion: The Monte Carlo codes should revise the fragmentation cross sections to more accurately simulate the physical properties of He-3 ions. The increase in RBE for He-3 ions is higher than for proton beams at the Bragg peak.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2012

Development of a coupling code for PWR reactor cavity radiation streaming calculation

PWR reactor cavity radiation streaming is important for the safe of the personnel and equipment, thus calculation has to be performed to evaluate the neutron flux distribution around the reactor. For this calculation, the deterministic codes have difficulties in fine geometrical modeling and need huge computer resource; and the Monte Carlo codes require very long sampling time to obtain results with acceptable precision. Therefore, a coupling method has been developed to eliminate the two problems mentioned above in each code. In this study, we develop a coupling code named DORT2MCNP to link the Sn code DORT and Monte Carlo code MCNP. DORT2MCNP is used to produce a combined surface source containing top, bottom and side surface simultaneously. Because SDEF card is unsuitable for the combined surface source, we modify the SOURCE subroutine of MCNP and compile MCNP for this application. Numerical results demonstrate the correctness of the coupling code DORT2MCNP and show reasonable agreement between the coupling method and the other two codes (DORT and MCNP). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Nuclear data uncertainty and sensitivity analysis of the VHTRC benchmark using SCALE

The Very High Temperature Critical Assembly (VHTRC) experiment represents one of the few data sets available for the validation of HTGR lattice physics. Within the framework of the IAEA Coordinated Research Project on HTGR Uncertainty Analysis in Modeling, uncertainty and sensitivity analyses of this graphite-moderated facility are performed as the validation reference to the prismatic MHTGR-350 fuel block calculations. Nominal multi-group and continuous-energy criticality calculations with the KENO-VI Monte Carlo code of the SCALE code package are compared with the continuous-energy Monte Carlo Code Serpent. Good agreement with a maximum difference of 250 pcm is obtained. The experimental data set, however, differs by several hundred pcm with the results of both codes when using the ENDF-VII.0 library. When using ENDF/B-VII.1 data, this difference is reduced to a few hundred pcm and the calculations lie within the experimental error bars. Uncertainties of the VHTRC multiplication factors due to uncertainties in nuclear data are determined with SAMPLER/KENO-VI of SCALE 6.2 and with continuous-energy TSUNAMI of SCALE 6.2. For all experimental configurations, the obtained uncertainty is found to be 0.58% when using ENDF/B-VII.0 data. The top contributor to this uncertainty is the average number of neutrons per fission event of U-235. With ENDF/B-VII.1 data, uncertainties of about 0.66% are obtained in particular due to an increased uncertainty of U-235 nu-bar in the latest ENDF release. In conclusion, when considering these nuclear data uncertainties, the obtained 1s uncertainty intervals are overlapping with the experimental error bars.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Neutronics/Thermal-hydraulics coupling with RMC and CTF for BEAVRS benchmark calculation

As the computational accuracy required for design and safety analysis of current and future nuclear reactors is continuously increasing, high fidelity numerical reactor simulations including different physics raise wide concern of researchers. The interplay between neutronic and thermal-hydraulic (N-TH) effects of a nuclear reactor core plays an important role in reactor design, safety analysis and core performance. Monte Carlo method can provide high fidelity neutronics analysis of different nuclear reactors, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. Armed with the massively parallel algorithms, the computational cost of Monte Carlo codes can be greatly reduced. Compared to single channel thermal process, sub-channel codes have enormous advantages of detailed modeling of full-core reactor or single assembly and high accuracy of calculation results. Detailed sub-channel modeling of a full-core also needs running in parallel, otherwise the computational cost will be relatively high. In this paper, a new hybrid coupling with continuous-energy Monte Carlo code RMC and the sub-channel code CTF has been developed. The coupling codes are applied to steady-state simulations of BEAVRS benchmark in hot full power condition. In order to deal with the temperature dependence of cross sections for Monte Carlo code, the on-the-fly cross sections treatment has been developed in RMC. The on-the-fly method was applied to the N-TH coupling in this work. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

SU-E-T-569: Neutron Shielding Calculation Using Analytical and Multi-Monte Carlo Method for Proton Therapy Facility

Purpose: To evaluate the shielding wall design to protect patients, staff and member of the general public for secondary neutron using a simply analytic solution, multi-Monte Carlo code MCNPX, ANISN and FLUKA. Methods: An analytical and multi-Monte Carlo method were calculated for proton facility (Sumitomo Heavy Industry Ltd.) at Samsung Medical Center in Korea. The NCRP-144 analytical evaluation methods, which produced conservative estimates on the dose equivalent values for the shielding, were used for analytical evaluations. Then, the radiation transport was simulated with the multi-Monte Carlo code. The neutron dose at evaluation point is got by the value using the production of the simulation value and the neutron dose coefficient introduced in ICRP-74. Results: The evaluation points of accelerator control room and control room entrance are mainly influenced by the point of the proton beam loss. So the neutron dose equivalent of accelerator control room for evaluation point is 0.651, 1.530, 0.912, 0.943 mSv/yr and the entrance of cyclotron room is 0.465, 0.790, 0.522, 0.453 mSv/yr with calculation by the method of NCRP-144 formalism, ANISN, FLUKA and MCNP, respectively. The most of Result of MCNPX and FLUKA using the complicated geometry showed smaller values than Result of ANISN. Conclusion: The neutron shielding for a proton therapy facility has been evaluated by the analytic model and multi-Monte Carlo methods. We confirmed that the setting of shielding was located in well accessible area to people when the proton facility is operated.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2013

Performance of the Los Alamos National Laboratory spallation-driven solid-deuterium ultra-cold neutron source

In this paper, we describe the performance of the Los Alamos spallation-driven solid-deuterium ultra-cold neutron (UCN) source. Measurements of the cold neutron flux, the very low energy neutron production rate, and the UCN rates and density at the exit from the biological shield are presented and compared to Monte Carlo predictions. The cold neutron rates compare well with predictions from the Monte Carlo code MCNPX and the UCN rates agree with our custom UCN Monte Carlo code. The source is shown to perform as modeled. The maximum delivered UCN density at the exit from the biological shield is 52(9) UCN/cc with a solid deuterium volume of {approx}1500 cm{sup 3}.

43 PARTICLE ACCELERATORS↗
DOE OSTI2023

ITS Version 6.7: The Integrated TIGER Series of Coupled Electron/Photon Monte Carlo Transport Codes User's Manual

ITS is a powerful software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the make system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 6, the latest version of ITS, contains (1) improvements to the ITS 5.0 codes, and (2) conversion to Fortran 95. The general user friendliness of the software has been enhanced through memory allocation to reduce the need for users to modify and recompile the code.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

MCNP6 and SERPENT2 Performances with Unstructured Mesh Geometry

Traditionally, Monte Carlo codes model a nuclear reactor by combinatorial geometry. In this model, volume primitives (e,g, spheres, cylinders, parallelepipeds...) are combined together by Boolean operations (e.g. union, intersection, and complement) and replicated by lattice definitions to build the core geometry. Recently, high performances computing platforms allowed using unstructured mesh geometry in Monte Carlo codes. When this option is chosen, any type of core geometry can be modeled since the restriction of using volume primitives is removed. In an unstructured mesh geometry, Boolean operations and lattice definitions are not used since each volume is defined by triangular surfaces (facets). This study focuses on two Monte Carlo codes: MCNP (version 6.1.1 beta) developed at Los Alamos National Laboratory since 1957 and written in FORTRAN90, C, and C++ (267,000 total source code lines), and SERPENT (version 2.1.27) developed at VTT Technical Research Centre since 2004 and written in C (195,000 total source code lines). These two codes follow different approaches for the unstructured mesh geometry modeling: MCNP uses ABAQUS format, whereas SERPENT uses the STL format. In the STL (STereoLithographic) geometry model only the surfaces of the volume are meshed; the interior part of the volumes is not meshed. In the STL geometry model, the whole geometry is defined by the sequence of SOLID and ENDSOLID sections. Each volume of the geometry must have a SOLID and ENDSOLID section. The STL file contains no information specifying the materials of the volumes. MCNP and SERPENT can model any kind of complicated geometry using unstructured meshes. MCNP and SERPENT use the ABAQUS and STL file formats, respectively. The STL file format does not have tetrahedrons and has only triangular facets on the volumes boundary. This allows SERPENT simulations to run 10 times faster than MCNP. However, if Monte Carlo results have to be coupled to thermal-hydraulics codes, the interior of the geometry volumes must be meshed, which may reduce SERPENT performances.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2026

On the Statistical Uncertainty of Monte Carlo-Calculated Scattering Sensitivities

Sensitivity coefficients calculated with Monte Carlo codes are widely used for nuclear data uncertainty quantification in the modeling and simulation of complex 3D reactor systems. This study systematically compares sensitivity coefficients and associated statistical uncertainties for the multiplication factor and fuel temperature reactivity across multiple Monte Carlo codes (SCALE/KENO, SCALE/Shift, MCNP, and Serpent) using simple models representing light-water reactors and advanced reactor concepts. For multiplication factor sensitivities, statistical uncertainties are generally acceptable, although scattering sensitivities show significantly larger statistical uncertainties than, for example, fission and capture reactions. Fuel temperature reactivity sensitivities show significantly larger statistical uncertainties across all reactions. Elastic scattering sensitivities are the most problematic: all Monte Carlo codes fail to resolve energy-dependent coefficients, and they produce dramatically different energy-collapsed values. Critically, the use of these sensitivity coefficients in nuclear data uncertainty propagation leads to reduced statistical uncertainties in individual uncertainty contributions. This can lead to the masking of unusable sensitivity coefficients and producing misleading uncertainty results. The findings of this study show that new or enhanced methods are needed to improve Monte Carlo elastic scattering sensitivity calculations. Additionally, this study shows the relevance of verifying sensitivity coefficients through direct perturbation calculations for individual nuclide reactions, instead of only for total cross sections as commonly done.

Bostelmann, Rike [ORNL] (ORCID:0000000165968088)↗
DOE OSTI2011

Development of Monte Carlo Simulation Code to Model Behavior of Hydrogen Isotopes Loaded into Tungsten Containing Vacancies

The behavior of hydrogen isotopes implanted into tungsten containing vacancies was simulated using a Monte Carlo technique. The correlations between the distribution of implanted deuterium and fluence, trap density and trap distribution were evaluated. Throughout the present study, qualitatively understandable results were obtained. In order to improve the precision of the model and obtain quantitatively reliable results, it is necessary to deal with the following subjects: (1) how to balance long-time irradiation processes with a rapid diffusion process, (2) how to prevent unrealistic accumulation of hydrogen, and (3) how to model the release of hydrogen forcibly loaded into a region where hydrogen densely exist already.

08 HYDROGEN↗
DOE OSTI2023

Monte Carlo Analyses of the FOEHN Experiment

The FOEHN critical experiment, which has been carried out at Cadarache (France) in the reactor EOLE in the early 70s, was designed to verify RHF design analyses. This study focuses on the impact of the FOEHN experiment uncertainties on the calculation of the energy deposition. The latter has been obtained by the Monte Carlo codes Serpent and MCNP. In the calculation of the energy deposition, the first code relies on KERMA factors whereas the latter code relies on Q-values. The two Monte Carlo codes share the same geometry and material specifications. In addition to the design and experiment parameters uncertainties, nuclear data also impacts the calculation of the energy deposition. In this study, the MCNP simulations use three different nuclear data library sets: ENDF/B versions VII.0, VII.1, and VIII.0. Finally, the ksens card of MCNP, using perturbation theory, has been used to identify the cross sections and isotopes with largest impact on the effective multiplication factor. By showing that the use of different software and computational methodologies and design parameters uncertainties do not significantly impact the core power distribution, this work demonstrate the appropriateness of the tools and methods employed for the conversion of involute-plate reactors.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

MCNP 2020: Preparing LANL Monte Carlo for Exascale Computer Systems (White Paper)

The MCNP Monte Carlo code for simulating particle transport is the product of over 60 years of Monte Carlo code development at LANL. MCNP is the most-used code worldwide for particle transport in nuclear systems, with over 10,000 users in dozens of application areas. At LANL, MCNP is used extensively in applications for nuclear criticality safety, stockpile stewardship, homeland security, radiological safety, experiment design and analysis, etc., and is considered a critical technology resource for national defense. MCNP is often referred to as the “gold standard” for radiation transport calculations, based on the long development history and extensive validation work. Today, however, this critical technology and its well-deserved reputation are at risk. We propose a targeted effort over the next several years to upgrade the core software portions of the MCNP code. This effort will focus on fundamental structural changes to the underlying MCNP algorithms, data storage, and code architecture. These changes will directly improve code maintainability, reduce the development time for new features, improve the code performance, and make the code more flexible and adaptable for expected future computer systems. To move forward with new computer systems and additional new features, it is vital to have a solid base.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Status of Cross-section Data Libraries for MCNP

The Data Team, part of the XCP-5 group at LANL, uses the NJOY code and other codes to prepare the cross-section data files used by the Monte Carlo codes MCNP5, MCNPX, and MCNP6. They perform extensive quality assurance checking of those data files. The cross-section data files are included with the releases of MCNP that are distributed by RSICC. MCNP5 is developed and supported by the Monte Carlo Codes group (XCP-3) at LANL. MCNPX is developed and supported by the NEN-5 group at LANL. MCNP6 is a merger of MCNP5 and MCNPX, with numerous additional features, and is jointly developed and supported by the XCP-3 and NEN-5 groups. MCNP5 and MCNPX are "frozen" codes, and all new development is being carried out only in the MCNP6 code. In the past, when new types of data or improved representations were developed, the Data Team would incorporate the changes into the cross-section data files, and the Monte Carlo groups would make corresponding changes to the MCNP codes to use the new data. At present, however, only MCNP6 is updated to handle new data representations; MCNP5/X are not.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

A review of the use and potential of the GATE Monte Carlo simulation code for radiation therapy and dosimetry applications

In this paper, the authors' review the applicability of the open-source GATE Monte Carlo simulation platform based on the GEANT4 toolkit for radiation therapy and dosimetry applications. The many applications of GATE for state-of-the-art radiotherapy simulations are described including external beam radiotherapy, brachytherapy, intraoperative radiotherapy, hadrontherapy, molecular radiotherapy, and in vivo dose monitoring. Investigations that have been performed using GEANT4 only are also mentioned to illustrate the potential of GATE. The very practical feature of GATE making it easy to model both a treatment and an imaging acquisition within the same frameworkis emphasized. The computational times associated with several applications are provided to illustrate the practical feasibility of the simulations using current computing facilities.

62 RADIOLOGY AND NUCLEAR MEDICINE↗