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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 451 records · Page 25
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

An Investigation on the Sustainability of a Sodium-cooled Breed-and-Burn Fast Reactor (B and BR)

Fast reactor can effectively handle the used nuclear fuel (UNF) by transmuting long-living transuranic (TRUs) into relatively short-living fission products. It can also be used to produce fissile from fertile nuclides as fuels for another reactor. The breed-and-burn fast reactor (B and BR) is a unique concept of fast reactor, which can breed the fissile fuels and use the bred fuels in situ. Thanks to this characteristic, the fuel utilization in a B and BR can be extremely high and even the recycled fuel of a B and BR can be used as a fuel to spawn the next generation B and BR. The UNF from the first B and BR should be reprocessed or reconditioned so that it can be reused in the second B and BR. A fuel reprocessing technology named melt refining was used in EBR-II from 1964 to 1969 and it was also applied in our previous study which showed that second B and BR can be started using the recycled UNF of first B and BR. However, about 95% of rare earths are removed during the melt refining and the associated processes are not very simple. Moreover, the actinide loss can also be significant. To enhance the proliferation-resistance and economy of the reprocessing, a super-simplified melt and treatment (SSMT) process without removing rare earth nuclides is suggested and the feasibility study to start the second B and BR using SSMT is carried out in this paper. The neutronics simulations were done using the Monte Carlo code McCARD in conjunction with ENDF/B-VII cross-section library. (authors)

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

On determining dose rate constants spectroscopically

Purpose: To investigate several aspects of the Chen and Nath spectroscopic method of determining the dose rate constants of {sup 125}I and {sup 103}Pd seeds [Z. Chen and R. Nath, Phys. Med. Biol. 55, 6089-6104 (2010)] including the accuracy of using a line or dual-point source approximation as done in their method, and the accuracy of ignoring the effects of the scattered photons in the spectra. Additionally, the authors investigate the accuracy of the literature's many different spectra for bare, i.e., unencapsulated {sup 125}I and {sup 103}Pd sources. Methods: Spectra generated by 14 {sup 125}I and 6 {sup 103}Pd seeds were calculated in vacuo at 10 cm from the source in a 2.7 Multiplication-Sign 2.7 Multiplication-Sign 0.05 cm{sup 3} voxel using the EGSnrc BrachyDose Monte Carlo code. Calculated spectra used the initial photon spectra recommended by AAPM's TG-43U1 and NCRP (National Council of Radiation Protection and Measurements) Report 58 for the {sup 125}I seeds, or TG-43U1 and NNDC(2000) (National Nuclear Data Center, 2000) for {sup 103}Pd seeds. The emitted spectra were treated as coming from a line or dual-point source in a Monte Carlo simulation to calculate the dose rate constant. The TG-43U1 definition of the dose rate constant was used. These calculations were performed using the full spectrum including scattered photons or using only the main peaks in the spectrum as done experimentally. Statistical uncertainties on the air kerma/history and the dose rate/history were Less-Than-Or-Slanted-Equal-To 0.2%. The dose rate constants were also calculated using Monte Carlo simulations of the full seed model. Results: The ratio of the intensity of the 31 keV line relative to that of the main peak in {sup 125}I spectra is, on average, 6.8% higher when calculated with the NCRP Report 58 initial spectrum vs that calculated with TG-43U1 initial spectrum. The {sup 103}Pd spectra exhibit an average 6.2% decrease in the 22.9 keV line relative to the main peak when calculated with the TG-43U1 rather than the NNDC(2000) initial spectrum. The measured values from three different investigations are in much better agreement with the calculations using the NCRP Report 58 and NNDC(2000) initial spectra with average discrepancies of 0.9% and 1.7% for the {sup 125}I and {sup 103}Pd seeds, respectively. However, there are no differences in the calculated TG-43U1 brachytherapy parameters using either initial spectrum in both cases. Similarly, there were no differences outside the statistical uncertainties of 0.1% or 0.2%, in the average energy, air kerma/history, dose rate/history, and dose rate constant when calculated using either the full photon spectrum or the main-peaks-only spectrum. Conclusions: Our calculated dose rate constants based on using the calculated on-axis spectrum and a line or dual-point source model are in excellent agreement (0.5% on average) with the values of Chen and Nath, verifying the accuracy of their more approximate method of going from the spectrum to the dose rate constant. However, the dose rate constants based on full seed models differ by between +4.6% and -1.5% from those based on the line or dual-point source approximations. These results suggest that the main value of spectroscopic measurements is to verify full Monte Carlo models of the seeds by comparison to the calculated spectra.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2018

The state of Monte Carlo neutron transport on GPU accelerators - A review - 25609

A review of the current state of Monte Carlo neutron transport of Graphical Processing Units (GPUs) has been performed. Up-to-date data has been solicited from researchers in the field and combined with data available in literature to provide the current state of Monte Carlo neutron transport on GPU hardware. Nine implementations of neutron transport on GPUs have been summarized and the performance results have been presented to allow the reader to compare performance. A comparison of the results has shown that researchers obtained similar performance when scaled to the performance of NVIDIA Tesla K40 GPUs. A strong scaling equation will be presented that can be used to describe the GPU performance against different numbers of CPU cores. One implementation of a production Monte Carlo code on GPU hardware, the code Shift, showed that newer NVIDIA Pascal and Volta based GPUs provide a significant performance improvement. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

Zero power experiment for molten salt reactor - code verification and validation request

Molten Salt Reactors (MSRs) are one of the six perspective Generation IV nuclear technologies that should fulfil sustainability, economics, safety, and proliferation resistance goals. In recent years these concepts have attracted increased interest in research as well as in industry. In the context of developing new, innovative technology, a zero-power experiment would be an ideal option for improving the understanding of MSR systems under low-risk and low-cost conditions. The given research was focused on cross-verification of the nuclear codes for MSR modelling and simulations as required for the development of the first design of a potential facility and a draft of an experimental programme. The analysis included the evaluation of lattice and Monte Carlo codes. The study showed that all considered nuclear tools were in good agreement with each other. Although this does not mean the results will match the real experimental data, which is essential for the regulatory approval of a novel technology. Currently, experimental data is almost absent for MSR technology. Hence, a zero power experiment is needed to assure we can rely on modelling and simulation studies for further development of MSR technology and deliver real-world experiments to support the licensing of future MSRs. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

A New Standard DNA Damage (SDD) Data Format

Our understanding of radiation-induced cellular damage has greatly improved over the past few decades. Despite this progress, there are still many obstacles to fully understand how radiation interacts with biologically relevant cellular components, such as DNA, to cause observable end points such as cell killing. Damage in DNA is identified as a major route of cell killing. One hurdle when modeling biological effects is the difficulty in directly comparing results generated by members of different research groups. Multiple Monte Carlo codes have been developed to simulate damage induction at the DNA scale, while at the same time various groups have developed models that describe DNA repair processes with varying levels of detail. These repair models are intrinsically linked to the damage model employed in their development, making it difficult to disentangle systematic effects in either part of the modeling chain. These modeling chains typically consist of track-structure Monte Carlo simulations of the physical interactions creating direct damages to DNA, followed by simulations of the production and initial reactions of chemical species causing so-called “indirect” damages. After the induction of DNA damage, DNA repair models combine the simulated damage patterns with biological models to determine the biological consequences of the damage. To date, the effect of the environment, such as molecular oxygen (normoxic vs. hypoxic), has been poorly considered. We propose a new standard DNA damage (SDD) data format to unify the interface between the simulation of damage induction in DNA and the biological modeling of DNA repair processes, and introduce the effect of the environment (molecular oxygen or other compounds) as a flexible parameter. Such a standard greatly facilitates intermodel comparisons, providing an ideal environment to tease out model assumptions and identify persistent, underlying mechanisms. Through inter-model comparisons, this unified standard has the potential to significantly advance our understanding of the underlying mechanisms of radiation-induced DNA damage and the resulting observable biological effects when radiation parameters and/or environmental conditions change.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗
DOE OSTI2016

Neutron flux and power in RTP core-15

PUSPATI TRIGA Reactor achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution of TRIGA core. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core with literally no physical approximation. The consistency and accuracy of the developed RTP MCNP model was established by comparing calculations to the available experimental results and TRIGLAV code calculation.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2024

Validation of the ERO2.0 code using W7-X and JET experiments and predictions for ITER operation

Abstract The paper provides an overview of recent modelling of global material erosion and deposition in the fusion devices Wendelstein 7-X (W7-X), JET and ITER using the Monte-Carlo code ERO2.0. For validating the modelling tool in a three-dimensional environment, W7-X simulations are performed to describe carbon erosion from the graphite test divertor units, which were equipped in operational phase OP 1.2 and analysed post-mortem. Synthetic spectroscopy of carbon line emission is compared with experimental results from the divertor spectrometer measurement system, showing a good agreement in the e-folding lengths in the radial intensity profiles of carbon. In the case of metallic wall materials, earlier modelling of the Be/W environment in JET and ITER is revisited and extended with an updated set of sputtering and reflection data, as well as including the mixing model for describing the Be/W dynamics in the divertor. Motivated by recent H/D/T isotope experiments in JET, limited and diverted configuration pulses are modelled, showing the expected trend of both Be and W erosion increasing with isotope mass. For the JET diverted configuration pulses, it is shown that Be migrates predominantly to the upper part of the inner divertor where it initially leads to strong W erosion. With longer exposure time, the growth of a Be deposited layer leads to a reduction of W erosion in that region. A similar trend is observed in simulations of the ITER baseline Q = 10 scenario, however with a more symmetric Be migration pattern leading to deposition also on the outer divertor.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

SU-F-T-166: On the Nature of the Background Visible Light Observed in Fiber Optic Dosimetry of Proton Beams

Purpose: The nature of the background visible light observed during fiber optic dosimetry of proton beams, whether it is due to Cherenkov radiation or not, has been debated in the literature recently. In this work, experimentally and by means of Monte Carlo simulations, we shed light on this problem and investigated the nature of the background visible light observed in fiber optics irradiated with proton beams. Methods: A bare silica fiber optics was embedded in tissue-mimicking phantoms and irradiated with clinical proton beams with energies of 100–225 MeV at Roberts Proton Therapy Center. Luminescence spectroscopy was performed by a CCD-coupled spectrograph to analyze in detail the emission spectrum of the fiber tip across the visible range of 400–700 nm. Monte Carlo simulation was performed by using FLUKA Monte Carlo code to simulate Cherenkov light and ionizing radiation dose deposition in the fiber. Results: The experimental spectra of the irradiated silica fiber shows two distinct peaks at 450 and 650 nm, whose spectral shape is different from that of Cherenkov radiation. We believe that the nature of these peaks are connected to the point defects of silica including oxygen-deficiency center (ODC) and non-bridging oxygen hole center (NBOHC). Monte Carlo simulations confirmed the experimental observations that Cherenkov radiation cannot be solely responsible for such a signal. Conclusion: We showed that Cherenkov radiation is not the dominant visible signal observed in bare fiber optics irradiated with proton beams. We observed two distinct peaks at 450 and 650 nm whose nature is connected with the point defects of silica fiber including oxygen-deficiency center and non-bridging oxygen hole center.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2016

SU-F-T-657: In-Room Neutron Dose From High Energy Photon Beams

Purpose: To estimate neutron dose inside the treatment room from photodisintegration events in high energy photon beams using Monte Carlo simulations and experimental measurements. Methods: The Monte Carlo code MCNP6 was used for the simulations. An Eberline ESP-1 Smart Portable Neutron Detector was used to measure neutron dose. A water phantom was centered at isocenter on the treatment couch, and the detector was placed near the phantom. A Varian 2100EX linear accelerator delivered an 18MV open field photon beam to the phantom at 400MU/min, and a camera captured the detector readings. The experimental setup was modeled in the Monte Carlo simulation. The source was modeled for two extreme cases: a) hemispherical photon source emitting from the target and b) cone source with an angle of the primary collimator cone. The model includes the target, primary collimator, flattening filter, secondary collimators, water phantom, detector and concrete walls. Energy deposition tallies were measured for neutrons in the detector and for photons at the center of the phantom. Results: For an 18MV beam with an open 10cm by 10cm field and the gantry at 180°, the Monte Carlo simulations predict the neutron dose in the detector to be 0.11% of the photon dose in the water phantom for case a) and 0.01% for case b). The measured neutron dose is 0.04% of the photon dose. Considering the range of neutron dose predicted by Monte Carlo simulations, the calculated results are in good agreement with measurements. Conclusion: We calculated in-room neutron dose by using Monte Carlo techniques, and the predicted neutron dose is confirmed by experimental measurements. If we remodel the source as an electron beam hitting the target for a more accurate representation of the bremsstrahlung fluence, it is feasible that the Monte Carlo simulations can be used to help in shielding designs.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2016

The impact of absorption coefficient on polarimetric determination of Berry phase based depth resolved characterization of biomedical scattering samples: a polarized Monte Carlo investigation

The modulation of the state of polarization of photons due to scatter generates associated geometric phase that is being investigated as a means for decreasing the degree of uncertainty in back-projecting the paths traversed by photons detected in backscattered geometry. In our previous work, we established that polarimetrically detected Berry phase correlates with the mean photon penetration depth of the backscattered photons collected for image formation. In this work, we report on the impact of state-of-linear-polarization (SOLP) filtering on both the magnitude and population distributions of image forming detected photons as a function of the absorption coefficient of the scattering sample. The results, based on Berry phase tracking implemented Polarized Monte Carlo Code, indicate that sample absorption plays a significant role in the mean depth attained by the image forming backscattered detected photons.

backscattered Mueller matrix↗
DOE OSTI2014

SciDAC Center for Simulation of Wave-Plasma Interactions - Iterated Finite-Orbit Monte Carlo Simulations with Full-Wave Fields for Modeling Tokamak ICRF Wave Heating Experiments - Final Report

This final report describes the work performed under U.S. Department of Energy Cooperative Agreement DE-FC02-08ER54954 for the period April 1, 2011 through March 31, 2013. The goal of this project was to perform iterated finite-orbit Monte Carlo simulations with full-wall fields for modeling tokamak ICRF wave heating experiments. In year 1, the finite-orbit Monte-Carlo code ORBIT-RF and its iteration algorithms with the full-wave code AORSA were improved to enable systematical study of the factors responsible for the discrepancy in the simulated and the measured fast-ion FIDA signals in the DIII-D and NSTX ICRF fast-wave (FW) experiments. In year 2, ORBIT-RF was coupled to the TORIC full-wave code for a comparative study of ORBIT-RF/TORIC and ORBIT-RF/AORSA results in FW experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

Accelerator shield design of KIPT neutron source facility

Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the design development of a neutron source facility at KIPT utilizing an electron-accelerator-driven subcritical assembly. Electron beam power is 100 kW, using 100 MeV electrons. The facility is designed to perform basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The biological shield of the accelerator building is designed to reduce the biological dose to less than 0.5-mrem/hr during operation. The main source of the biological dose is the photons and the neutrons generated by interactions of leaked electrons from the electron gun and accelerator sections with the surrounding concrete and accelerator materials. The Monte Carlo code MCNPX serves as the calculation tool for the shield design, due to its capability to transport electrons, photons, and neutrons coupled problems. The direct photon dose can be tallied by MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is less than 0.01 neutron per electron. This causes difficulties for Monte Carlo analyses and consumes tremendous computation time for tallying with acceptable statistics the neutron dose outside the shield boundary. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were developed for the study. The generated neutrons are banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron and secondary photon doses. The weight windows variance reduction technique is utilized for both neutron and photon dose calculations. Two shielding materials, i.e., heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total dose outside the shield boundary at less than 0.5-mrem/hr. The shield configuration and parameters of the accelerator building have been determined and are presented in this paper. (authors)

43 PARTICLE ACCELERATORS↗
DOE OSTI2020

Validation of MCNP Critical Benchmark Models of Moderated Highly Enriched Uranium Slabs

A new centralized repository of high-quality MCNP models of critical benchmark experiments is currently under development at Los Alamos National Laboratory (LANL). The benchmark experiments are described in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook, and the initial set of benchmark models are derived from the Whisper Suite provided with MCNP6.2. This effort is a collaboration among the Nuclear Criticality Safety, Nuclear Data, and Monte Carlo code development/application organizations at LANL. The goal is to build a single LANL benchmark collection that is up to date with the latest ICSBEP revision, has a formal review and revision process, is contained in an open-source repository, and utilizes new Python tools for improved input and output file review. This paper describes the validation of the models associated with HEU-MET-FAST-007, “Uranium Metal Slabs Moderated with Polyethylene, Plexiglas, and Teflon”. The Monte Carlo n-Particle (MCNP) models were compared to the second revision of HEU-MET-FAST-007. The experiment considered critical configurations of highly enriched uranium (HEU) metal slabs and various moderators in 43 unique cases. The slabs of uranium were separated by layers of moderating material, forming an assembly. The assembly was separated in two halves with one half placed on a stationary table and the other attached to a moveable table that could be raised and lowered via pulleys connected to the ceiling of the shielded room. To reach criticality, the two halves were brought together with a negligible gap between the assemblies.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2011

Advances in the Development and Verification of MCNP5 and MCNP6

The MCNP Monte Carlo code has been used for high-fidelity analyses of criticality safety problems since the 1970s. This paper reviews recent advances in the development and verification of MCNP, including the current production release of MCNP5 and the beta release of MCNP6. End-users in all application areas need to be aware of the forthcoming MCNP6 release and begin planning for the transition to the new code in 2012 and beyond.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Second-Order Uncertainty Analysis for Radiation Shielding Problems Using MCNP

Standard uncertainty propagation is based on a first-order Taylor expansion. This approximation is usually justified by assuming that 'uncertainties are very small numbers compared with the values of their corresponding quantities'. Furthermore, the second derivatives needed for a second-order analysis are not usually available, and a second-order analysis requires more knowledge regarding the distribution of the random variables. For radiation shielding problems, the MCNP6 Monte Carlo code computes first and second derivatives of a response with respect to nuclide density or reaction cross section. This paper explores the application of these derivatives to second-order uncertainty analysis.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

SU-E-T-212: Comparison of TG-43 Dosimetric Parameters of Low and High Energy Brachytherapy Sources Obtained by MCNP Code Versions of 4C, X and 5

Purpose: Different versions of MCNP code are widely used for dosimetry purposes. The purpose of this study is to compare different versions of the MCNP codes in dosimetric evaluation of different brachytherapy sources. Methods: The TG-43 parameters such as dose rate constant, radial dose function, and anisotropy function of different brachytherapy sources, i.e. Pd-103, I-125, Ir-192, and Cs-137 were calculated in water phantom. The results obtained by three versions of Monte Carlo codes (MCNP4C, MCNPX, MCNP5) were compared for low and high energy brachytherapy sources. Then the cross section library of MCNP4C code was changed to ENDF/B-VI release 8 which is used in MCNP5 and MCNPX codes. Finally, the TG-43 parameters obtained using the MCNP4C-revised code, were compared with other codes. Results: The results of these investigations indicate that for high energy sources, the differences in TG-43 parameters between the codes are less than 1% for Ir-192 and less than 0.5% for Cs-137. However for low energy sources like I-125 and Pd-103, large discrepancies are observed in the g(r) values obtained by MCNP4C and the two other codes. The differences between g(r) values calculated using MCNP4C and MCNP5 at the distance of 6cm were found to be about 17% and 28% for I-125 and Pd-103 respectively. The results obtained with MCNP4C-revised and MCNPX were similar. However, the maximum difference between the results obtained with the MCNP5 and MCNP4C-revised codes was 2% at 6cm. Conclusion: The results indicate that using MCNP4C code for dosimetry of low energy brachytherapy sources can cause large errors in the results. Therefore it is recommended not to use this code for low energy sources, unless its cross section library is changed. Since the results obtained with MCNP4C-revised and MCNPX were similar, it is concluded that the difference between MCNP4C and MCNPX is their cross section libraries.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2023

Deep Learning for Multigroup Cross-Section Representation in Two-Step Core Calculations

Here we investigate using deep learning, a type of machine-learning algorithm employing multiple layers of artificial neurons, for the mathematical representation of multigroup cross sections for use in the Griffin reactor multiphysics code for two-step deterministic neutronics calculations. A three-dimensional fuel element typical of a high-temperature gas reactor as well as a two-dimensional sodium-cooled fast reactor lattice are modeled using the Serpent Monte Carlo code, and multigroup macroscopic cross sections are generated for various state parameters to produce a training data set and a separate validation data set. A fully connected, feedforward neural network is trained using the open-source PyTorch machine-learning framework, and its accuracy is compared against the standard piecewise linear interpolation model. Additionally, we provide in this work a generic technique for propagating the cross-section model errors up to the k eff using sensitivity coefficients with the first-order uncertainty propagation rule. Quantifying the eigenvalue error due to the cross-section regression errors is especially practical for appropriately selecting the mathematical representation of the cross sections. We demonstrate that the artificial neural network model produces lower errors and therefore enables better accuracy relative to the piecewise linear model when the cross sections exhibit nonlinear dependencies; especially when a coarse grid is employed, where the errors can be halved by the artificial neural network. However, for linearly dependent multigroup cross sections as found for the sodium-cooled fast reactor case, a simpler linear regression outperforms deeper networks.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Progress in modelling fast-ion D-alpha spectra and neutral particle analyzer fluxes using FIDASIM

FIDASIM is a code that models signals produced by charge-exchange reactions between neutrals and ions (both fast and thermal) in magnetically confined plasmas. With the ion distribution function as input, the code predicts the efflux to a neutral particle analyzer diagnostic and the photon radiance of Balmer-alpha light to a fast-ion D α diagnostic, in addition to many other related quantities. A new, parallelized version of the Monte Carlo code FIDASIM has been developed in Fortran90 that is substantially faster than the original interactive data language version. Modified algorithms include more accurate treatments of the time dependent collisional-radiative equations that describe neutral energy levels, of the cloud of ‘halo’ neutrals that surround the injected neutral beam, and of finite Larmor radius effects. Enhanced physics capabilities include modelling ‘passive’ signals from cold edge neutrals, the ability to treat general three-dimensional magnetic confinement configurations, and calculations of diagnostic-specific weight functions that enable tomographic reconstructions of the fast-ion distribution function. Neutral beam attenuation, beam emission, and fast-ion birth profiles are also modelled. Finally, the new algorithms have been successfully validated against experimental data and new features have been tested through benchmarks between two independently developed versions of the code.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗