Monte Carlo Transport Calculations Research Articles

Correlated Production and Analog Transport of Fission Neutrons and Photons using Fission Models FREYA, FIFRELIN and the Monte Carlo Code TRIPOLI-4® .

Fission modeling in general-purpose Monte Carlo transport codes often relies on average nuclear data provided by international evaluation libraries. As such, only average fission multiplicities are available and correlations between fission neutrons and photons are missing. Whereas uncorrelated fission physics is usually sufficient for standard reactor core and radiation shielding calculations, correlated fission secondaries are required for specialized nuclear instrumentation and detector modeling. For coincidence counting detector optimization for instance, precise simulation of fission neutrons and photons that remain correlated in time from birth to detection is essential. New developments were recently integrated into the Monte Carlo transport code TRIPOLI-4 to model fission physics more precisely, the purpose being to access event-by-event fission events from two different fission models: FREYA and FIFRELIN. TRIPOLI-4 simulations can now be performed, either by connecting via an API to the LLNL fission library including FREYA, or by reading external fission event data files produced by FIFRELIN beforehand. These new capabilities enable us to easily compare results from Monte Carlo transport calculations using the two fission models in a nuclear instrumentation application. In the first part of this paper, broad underlying principles of the two fission models are recalled. We then present experimental measurements of neutron angular correlations for 252Cf(sf) and 240Pu(sf). The correlations were measured for several neutron kinetic energy thresholds. In the latter part of the paper, simulation results are compared to experimental data. Spontaneous fissions in 252Cf and 240Pu are modeled by FREYA or FIFRELIN. Emitted neutrons and photons are subsequently transported to an array of scintillators by TRIPOLI-4 in analog mode to preserve their correlations. Angular correlations between fission neutrons obtained independently from these TRIPOLI-4 simulations, using either FREYA or FIFRELIN, are compared to experimental results. For 240Pu(sf), the measured correlations were used to tune the model parameters.

GEOMETRICAL EFFICIENCY OF PLANE-PARALLEL IONIZATION CHAMBERS IN PROTON SCANNING BEAM

For commissioning of a proton therapy unit depth dose distributions must be determined and introduced into the Treatment Planning System. In pencil beam scanning (PBS) technique, integral depth dose (IDD) acquisition should be performed with detector large enough to ensure entire beam laterally broadened by scattered and secondary contributions. The purpose of this article is to quantify, using measurements and Monte Carlo transport calculations, the ionization chamber's (IC) geometrical efficiency versus the chamber radius and proton beam energy. The geometrical efficiency of 0.99 was determined for energies up to 160 and 190 MeV for 4.08 and 6 cm radius IC. Much lower geometrical efficiency was obtained for the energy of 226.08 MeV and results in charge loss of 5.8 and 3.6%, respectively. Relative IDD differences between IC 4.08 and 6 cm in radius increase with proton energy and reach 2.4% at the mid-range depth for 226.08 MeV.

Abstract ID: 124 Application expansion of the Monte-Carlo based treatment planning system for BNCT to particle radiotherapy and X-ray therapy

University of Tsukuba is developing an accelerator-based treatment device for boron neutron capture therapy (BNCT). As part of the development, a Monte-Carlo based treatment planning system (developing code: Tsukuba-Plan) is being developed [1] . The Tsukuba-plan has employed PHITS as the Monte-Carlo transport calculation code [2] . PHITS can calculate neutron, photon as well as proton and heavy-ions including carbon-ions. Therefore, Tsukuba-Plan with PHITS enables to perform dose estimations for not only BNCT but also particle radiotherapy and X-ray therapy. In BNCT protocols for malignant brain tumor, X-ray therapy is added after BNCT to enhance the treatment effect. And in the future, proton therapy will be also applied in stead of the X-ray therapy in order to improve therapeutic dose distribution. Therefore, the treatment planning system is required dose estimation for the combined multi-modality therapy. Based on this background, the aim of this study is to expand application field of the Tsukuba-plan to conventional external radiotherapies. University of Tsukuba Hospital has X-ray therapy, proton therapy and BNCT. Thus, first, we have expanded Tsukuba-plan to proton therapy and X-ray therapy. Regarding proton therapy, we have constructed 155 MeV and 200 MeV proton beam source and a geometry struction data for the proton therapy device installed in our Hospital. Verification for calculation accuracy for proton irradiation were performed. Phantom experiments performed using the proton therapy device in advance were represented by the Tsukuba-Plan. The proton dose distributions in the phantom were determined, and the calculations were compared with the experimental values. Calculations for both of proton therapy and X-ray therapy were in good agreement with experimental values, respectively. The results demonstrated that the Tsukuba-Plan enables to estimate accurately doses for proton irradiation and X-ray irradiations. To put Tsukuba-Plan to practical use for proton therapy and X-ray therapy, Tsukuba-Plan is improved further and several verifications are performed. This work was supported by JSPS KAKENHI Grant Nos. JP16K15343.

On-the-fly treatment of temperature dependent cross sections in the unresolved resonance region in RMC code

With the rapid development of computational power and parallel algorithms, Monte Carlo method has been widely investigated and used in neutron transport simulations due to its advantages in geometry modelling and usage of continuous energy point-wise cross sections. Monte Carlo codes can also be coupled with thermal-hydraulics codes to consider feedbacks. One of the most important aspects of thermal-hydraulics feedback is the detailed temperature distribution, which results in the demands of updates of temperature dependent cross sections in neutron transport simulations. For thermal reactors such as PWR and HTGR, two neutron energy regions should be carefully considered: the resolved resonance region (RRR) and the thermal energy region, especially the resolved resonance region. Besides, the temperature dependence of cross sections in the unresolved resonance energy region (URR) is also important for fast reactors and some experimental critical assemblies. In this paper, the on-the-fly temperature treatment of cross sections in the RRR was implemented in RMC code. The on-the-fly method for the URR was also proposed and implemented in RMC code, which was also combined with the on-the-fly temperature treatments for resolved resonance energy. The proposed method was applied to Monte Carlo transport and depletion calculations. The accuracy and efficiency are compared for different combinations of methods of RRR and URR. The results show that the on-the-fly treatment has high efficiency and satisfactory fidelity. The probability tables with equiprobable bands were also developed to reduce the memory consumption, with keeping the same accuracy of original probability tables.

Measurement of neutron spectra in a silicon filtered neutron beam using stilbene detectors at the LVR-15 research reactor

A well-defined neutron spectrum is an essential tool for calibration and tests of spectrometry and dosimetry detectors, and evaluation methods for spectra processing. Many of the nowadays used neutron standards are calibrated against a fission spectrum which has a rather smooth energy dependence. In recent time, at the LVR-15 research reactor in Rez, an alternative approach was tested for the needs of fast neutron spectrometry detector calibration. This process comprises detector tests in a neutron beam, filtered by one meter of single-crystalline silicon, which contains several significant peaks in the fast neutron energy range. Tests in such neutron field can possibly reveal specific problems in the deconvolution matrix of the detection system, which may stay hidden in fields with a smooth structure and can provide a tool for a proper energy calibration. Test with several stilbene scintillator crystals in two different beam configurations supplemented by Monte-Carlo transport calculations have been carried out. The results have shown a high level of agreement between the experimental data and simulation, proving thus the accuracy of used deconvolution matrix. The chosen approach can, thus, provide a well-defined neutron reference field with a peaked structure for further tests of spectra evaluation methods and scintillation detector energy calibration.

Monte Carlo Sensitivity and Uncertainty Analysis with Continuous-Energy Covariance Data

The continuous-energy Monte Carlo (MC) sensitivity and uncertainty (S/U) analysis conducted using the multigroup covariance matrices has a theoretical pitfall in that it is inconsistent with the principle of continuous-energy MC neutronics calculations because the use of the multigroup covariance matrices means treating covariance data as multigroup variables rather than continuous-energy variables. As a way to get around this deficiency and perform the MC S/U analysis on the theoretically consistent principle, this paper presents a new continuous-energy MC S/U formulation which directly utilizes the continuous-energy covariance data in the evaluated nuclear data libraries instead of the multigroup covariance matrices produced by nuclear data processing codes. The validity of the new MC S/U formulation is examined in terms of the input-nuclear-data-induced k uncertainty of the Godiva critical assembly and the TMI-1 pin cell problem by inputting the continuous-energy covariance data of nuclides involved directly into the continuous-energy MC transport calculations by a Seoul National University MC code, McCARD.

AREVA Developments for an Efficient and Reliable use of Monte Carlo codes for Radiation Transport Applications

In the context of the rising of Monte Carlo transport calculations for any kind of application, AREVA recently improved its suite of engineering tools in order to produce efficient Monte Carlo workflow. Monte Carlo codes, such as MCNP or TRIPOLI, are recognized as reference codes to deal with a large range of radiation transport problems. However the inherent drawbacks of theses codes - laboring input file creation and long computation time - contrast with the maturity of the treatment of the physical phenomena. The goals of the recent AREVA developments were to reach similar efficiency as other mature engineering sciences such as finite elements analyses (e.g. structural or fluid dynamics). Among the main objectives, the creation of a graphical user interface offering CAD tools for geometry creation and other graphical features dedicated to the radiation field (source definition, tally definition) has been reached. The computations times are drastically reduced compared to few years ago thanks to the use of massive parallel runs, and above all, the implementation of hybrid variance reduction technics. From now engineering teams are capable to deliver much more prompt support to any nuclear projects dealing with reactors or fuel cycle facilities from conceptual phase to decommissioning.

Relativistic kinematics for photoneutron production in Monte Carlo transport calculations

MC21 is a continuous-energy Monte Carlo radiation transport code for the calculation of the steady-state spatial distributions of reaction rates in three-dimensional models. The code supports neutron and photon transport in fixed source problems, as well as iterated-fission source (eigenvalue) neutron problems. The capability to simulate the production of photon-induced neutrons has been added to the code. In this paper, relativistic two-body kinematics are used to derive exact expressions for the secondary energy and angle distribution of photoneutrons. This treatment, implemented in MC21, is important in cases where the evaluated photonuclear data does not give an explicit energy distribution. Comparisons of the relativistic relations were made to approximations in MCNP5 and TRIPOLI-4, highlighting the magnitude of the error of those approximations.

Use of Effective Diffusion Homogenization method with the Monte Carlo code for light water reactor

The use of Monte Carlo transport method with the Serpent code for generating unit cell cross sections of a light-water reactor is investigated. The geometry is a 3×3 array of cells, where homogenization is performed over the central cell, while the neighboring cells represent a kind of color-set scheme to model the radial leakage from the central cell. Instead of the default homogenization method of Serpent, the Effective Diffusion Homogenization method is applied externally, which conserves reaction rates, as well as the boundary partial currents of the central cell. The exercise serves to explore the potential of the Monte Carlo method for core design calculations and to validate and improve the existing computational scheme in which unit-cell calculations are based on the 1-D deterministic transport model in the WIMSD code. The cross sections by both methods are compared and applied to predict the hot-zero-power critical boron concentration and radial power distribution of the Krško NPP in comparison with measured values. The results confirm applicability of Monte Carlo transport calculations with EDH homogenization at the unit-cell level and warrant further extension to burnup and whole-assembly Monte Carlo modeling, at least for validation purposes due to present computational time constraints.

Comparative assessment of different approaches for the use of CAD geometry in Monte Carlo transport calculations

Computer aided design (CAD) is an important industrial way to produce high quality designs. Therefore, CAD geometries are in general used for engineering and the design of complex facilities like the ITER tokamak. Although Monte Carlo codes like MCNP are well suited to handle the complex 3D geometry of ITER for transport calculations, they rely on their own geometry description and are in general not able to directly use the CAD geometry. In this paper, three different approaches for the use of CAD geometries with MCNP calculations are investigated and assessed with regard to calculation performance and user-friendliness. The first method is the conversion of the CAD geometry into MCNP geometry employing the conversion software McCad developed by KIT. The second approach utilizes the MCNP6 mesh geometry feature for the particle tracking and relies on the conversion of the CAD geometry into a mesh model. The third method employs DAGMC, developed by the University of Wisconsin-Madison, for the direct particle tracking on the CAD geometry using a patched version of MCNP. The obtained results show that each method has its advantages depending on the complexity and size of the model, the calculation problem considered, and the expertise of the user.

One-Group Cross-Section Generation for Monte Carlo Burnup Codes: Multigroup Method Extension and Verification

Allowing Monte Carlo (MC) codes to perform fuel cycle calculations requires coupling to a point depletion solver. In order to perform depletion calculations, one-group cross sections must be provided in advance. This paper focuses on generating accurate one-group cross-section values using Monte Carlo transport codes. The proposed method is an alternative to the conventional direct reaction rate tally approach, which requires substantial computational effort. The method presented here is based on the multigroup approach, in which pregenerated multigroup sets are collapsed with MC calculated flux. In our previous studies, we showed that generating accurate one-group cross sections requires their tabulation against the background cross section (σ0) to account for the self-shielding effect in the unresolved resonance energy range. However, in previous studies, the model used for the calculation of σ0 was simplified by relying on user-specified Bell and Dancoff factors.This work demonstrates that the one-group cross-section values calculated under the previous simplified model assumptions may not always agree with the directly tallied values. More specifically, the assumption is not universally applicable to the analysis of reactor systems with different neutron spectra and may be inaccurate when the number of energy groups is reduced (i.e., from tens of thousands to hundreds of groups). Therefore, the original background cross-section model was extended by implicitly accounting for the Dancoff and Bell factors.The method developed here reconstructs the correct value of σ0 by utilizing statistical data generated within the MC transport calculation by default. The proposed method was implemented in the BGCore code system. The one-group cross-section values generated by BGCore were compared with those tallied directly from the MCNP code. Very good agreement in the one-group cross-section values was observed. The method does not carry any additional computational burden, and it is universally applicable to the analysis of thermal as well as fast reactor systems. Adopting this multigroup methodology, which accounts for self-shielding, allows generation of highly accurate cross sections even if the number of energy groups is significantly reduced (to hundreds versus tens of thousands of groups). This reduction considerably improves the computational efficiency, which makes the analysis of large-scale reactor problems feasible.

光子・電子のモンテカルロ輸送計算コードの内容と利用

光子・電子のモンテカルロ輸送計算コードの内容と利用

Improved cache performance in Monte Carlo transport calculations using energy banding

We present an energy banding algorithm for Monte Carlo (MC) neutral particle transport simulations which depend on large cross section lookup tables. In MC codes, read-only cross section data tables are accessed frequently, exhibit poor locality, and are typically too much large to fit in fast memory. Thus, performance is often limited by long latencies to RAM, or by off-node communication latencies when the data footprint is very large and must be decomposed on a distributed memory machine. The proposed energy banding algorithm allows maximal temporal reuse of data in band sizes that can flexibly accommodate different architectural features. The energy banding algorithm is general and has a number of benefits compared to the traditional approach. In the present analysis we explore its potential to achieve improvements in time-to-solution on modern cache-based architectures.

Quasi-Differential Neutron Scattering in Zirconium from 0.5 to 20 MeV

High-energy-neutron-scattering experiments for elemental zirconium were performed at the electron linear accelerator facility at Rensselaer Polytechnic Institute. The scattering experiments were performed in the energy region from 0.5 to 20 MeV using the time-of-flight technique. The scattering system is composed of an array of eight EJ301 liquid scintillator detectors coupled to photomultiplier tubes. The detector array collects data simultaneously at various angles. The raw signals from each detector were digitized and transferred to a personal computer hard drive for storage. The digitized data were postprocessed, and pulse-shape analysis was performed to determine whether the pulse was the result of a gamma ray or a neutron being detected. The experimental results were compared with Monte Carlo transport calculations that simulated the experiment. This comparison was a way to benchmark several nuclear data libraries used in the Monte Carlo code. Ratios of the calculated data to the experimental data (C/E values) are presented and used to compare the nuclear data libraries. Results show that the experimentally observed scattering cross section is smaller than the one used in the evaluated libraries at energies between 10 and 20 MeV. For all energies and angles, the investigated nuclear data libraries agree with the experimental data to within 9%. Overall, the JEFF-3.1 and JENDL-4.0 libraries provide the best match to the experimental data.

A source model for modulated electron radiation therapy using dynamic jaw movements

The development of fast and accurate source models (SMs) might be of crucial importance for the future clinical implementation of modulated electron radiation therapy (MERT). In this study, a SM is presented for reconstructing phase-space information of modulated electron beams using a few-leaf electron collimator (FLEC) and the photon jaws. During a FLEC-based delivery, two collimation devices (jaws and FLEC) modulate the electron beam characteristics dynamically. The SM separates the beam into a primary and a scattered component. The primary component is derived by a fast Monte Carlo (MC) transport calculation in air using the EGSnrc/BEAMnrc code. The scattered beam is modeled analytically. The accelerator was decomposed into its individual leaf components and the scattered beam was characterized at various levels of the accelerator. Scattered particles are assigned an energy and position by sampling pre-calculated probability distributions. The direction is estimated by geometrical arguments. Particles were assumed to emerge from tunable virtual sources on the side of each collimator leaf. A leaf-hit algorithm was developed to dynamically reject particles that are incident on any collimating leaf. Electron transport in air between the two collimation levels was calculated based on a MC-modified version of the Fermi-Eyges scattering theory. Correlations between direction and position were observed and taken into account at the final collimation level. To validate the model, reconstructed phase-space data were compared with the full accelerator MC phase-space data. The model accurately reproduced the beam characteristics and preserved important correlations. Depth and profile dose distributions in water were derived for square, rectangular, and off-axis field sizes and for a range of clinical energies. Discrepancies in the dose distributions and dose output were within 3% in all cases. Fast and accurate SMs open the possibility for fast treatment planning in MERT, based on an inverse optimization MC treatment planning scheme.

Validation of the Photon Dose Calculation Model in the VARSKIN 4 Skin Dose Computer Code

An updated version of the skin dose computer code VARSKIN, namely VARSKIN 4, was examined to determine the accuracy of the photon model in calculating dose rates with different combinations of source geometry and radionuclides. The reference data for this validation were obtained by means of Monte Carlo transport calculations using MCNP5. The geometries tested included the zero volume sources point and disc, as well as the volume sources sphere and cylinder. Three geometries were tested using source directly on the skin, source off the skin with an absorber material between source and skin, and source off the skin with only an air gap between source and skin. The results of these calculations showed that the non-volume sources produced dose rates that were in very good agreement with the Monte Carlo calculations, but the volume sources resulted in overestimates of the dose rates compared with the Monte Carlo results by factors that ranged up to about 2.5. The results for the air gap showed poor agreement with Monte Carlo for all source geometries, with the dose rates overestimated in all cases. The conclusion was that, for situations where the beta dose is dominant, these results are of little significance because the photon dose in such cases is generally a very small fraction of the total dose. For situations in which the photon dose is dominant, use of the point or disc geometries should be adequate in most cases except those in which the dose approaches or exceeds an applicable limit. Such situations will often require a more accurate dose assessment and may require the use of methods such as Monte Carlo transport calculations.

New integral experiments for large angle scattering cross section data benchmarking with DT neutron beam at JAEA/FNS

Abstract A new integral experiment with a deuteron–triton fusion (DT) neutron beam started in order to validate scattering cross section data. First the DT neutron beam was constructed with a collimator. The performance of the collimator system and the characteristics of the DT neutron beam were measured. Second a new integral experiment for type 316 stainless steel (SS316) was carried out with this DT neutron beam. The DT neutron beam of 3.5 cm in diameter was injected to the front surface center of an SS316 cylindrical assembly. Reaction rates of the 93Nb(n,2n)92mNb reaction in the assembly were measured with the activation foil method and were calculated with the Monte Carlo transport calculation code. The measurement points were located at three positions, on the center of the beam axis and at 15 cm and 30 cm apart from the axis. The ratio of calculation to experiment of the 93Nb(n,2n)92mNb reaction rate became smaller than 1 with the distance from the beam axis. Then, the dependency of each reaction rate on scattering angle was calculated. It was proved that at off-axis positions, where C/E is smaller than 1, 90° scattering contribute relatively larger than at on-axis positions and backward scattering made little contribution to the results in this experiment. The reasons of the discrepancy between the measured and calculated data will be investigated.

Towards Scalable Parallelism in Monte Carlo Particle Transport Codes Using Remote Memory Access

One forthcoming challenge in the area of high-performance computing is having the ability to run large-scale problems while coping with less memory per compute node. In this work, we investigate a novel data decomposition method that would allow Monte Carlo transport calculations to be performed on systems with limited memory per compute node. In this method, each compute node remotely retrieves a small set of geometry and cross-section data as needed and remotely accumulates local tallies when crossing the boundary of the local spatial domain. Initial results demonstrate that while the method does allow large problems to be run in a memory-limited environment, achieving scalability may be difficult due to inefficiencies in the current implementation of RMA operations.

A Monte Carlo Lattice Code with Probability Tables and Optimized Energy Meshes

The possibility of performing Monte Carlo transport calculations using cross-section probability tables on the entire energy spectrum is discussed in this paper. This method possesses straight advantages toward other representations: Self-shielding effects are represented during the random walk in a straightforward way, and the calculation cost remains below continuous-energy simulations. This study takes advantage of previous contributions made in subgroup-based self-shielding models, regarding the definitions of optimized energy meshes and adequate numerical methods for consistently computing cross-section probability tables. Moment-based probability-table cross sections along with an energy mesh comprising only 295 groups lead to results with a similar level of accuracy to those obtained with a continuous-energy Monte Carlo method. Another innovative aspect of this work is related to the introduction of correlated weight matrices into a Monte Carlo algorithm. These correlated weights are used to represent mutual self-shielding effects occurring where resonances of different isotopes overlap.

Comparison of global variance reduction techniques for Monte Carlo radiation transport simulations of ITER

It is known that a priori information on the distribution of neutron fluxes around a neutron transport model can improve the computational efficiency of Monte Carlo transport calculations. The uncertainty lies in finding a method which obtains this information as fast and effortlessly as possible. A number of global variance reduction techniques were tested on the ITER reference neutronics model for their applicability for performing large three-dimensional transport simulations for ITER. Among these, we propose a new method which is computationally fast and efficient in determining the a priori flux information. The new method, known as MAGIC, was found to perform well achieving an efficiency 74 times that of an analogue run; the next most efficient was FW-CADIS, performing 48 times better than analogue.