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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 613 records · Page 34
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Coupling the System Analysis Module with SAS4A/SASSYS-1

SAS4A/SASSYS-1 is a simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced reactors, with an emphasis on sodium fast reactors. SAS4A/SASSYS-1 has been under development and in active use for nearly forty-five years, and is currently maintained by the U.S. Department of Energy under the Office of Advanced Reactor Technology. Although SAS4A/SASSYS-1 contains a very capable primary and intermediate system modeling component, PRIMAR-4, it also has some shortcomings: outdated data management and code structure makes extension of the PRIMAR-4 module somewhat difficult. The user input format for PRIMAR-4 also limits the number of volumes and segments that can be used to describe a given system. The System Analysis Module (SAM) is a fairly new code development effort being carried out under the U.S. DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. SAM is being developed with advanced physical models, numerical methods, and software engineering practices; however, it is currently somewhat limited in the system components and phenomena that can be represented. For example, component models for electromagnetic pumps and multi-layer stratified volumes have not yet been developed. Nor is there support for a balance of plant model. Similarly, system-level phenomena such as control-rod driveline expansion and vessel elongation are not represented. This report documents fiscal year 2016 work that was carried out to couple the transient safety analysis capabilities of SAS4A/SASSYS-1 with the system modeling capabilities of SAM under the joint support of the ART and NEAMS programs. The coupling effort was successful and is demonstrated by evaluating an unprotected loss of flow transient for the Advanced Burner Test Reactor (ABTR) design. There are differences between the stand-alone SAS4A/SASSYS-1 simulations and the coupled SAS/SAM simulations, but these are mainly attributed to the limited maturity of the SAM development effort. The severe accident modeling capabilities in SAS4A/SASSYS-1 (sodium boiling, fuel melting and relocation) will continue to play a vital role for a long time. Therefore, the SAS4A/SASSYS-1 modernization effort should remain a high priority task under the ART program to ensure continued participation in domestic and international SFR safety collaborations and design optimizations. On the other hand, SAM provides an advanced system analysis tool, with improved numerical solution schemes, data management, code flexibility, and accuracy. SAM is still in early stages of development and will require continued support from NEAMS to fulfill its potential and to mature into a production tool for advanced reactor safety analysis. The effort to couple SAS4A/SASSYS-1 and SAM is the first step on the integration of these modeling capabilities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Enabling event-by-event precision in γ-ray cascades for neutron-induced reactions

Neutron-induced γ-ray spectra provide key inputs for modern active interrogation applications. A precise modeling of the nuclear reaction and subsequent emission of γ rays is challenging and often impossible due to limitations on evaluated data file formats and nuclear transport simulation codes. We present a framework that addresses these challenges by combining experimental data and reaction-model calculation outputs into an extended candidate version of the Generalized Nuclear Data Structure (GNDS) file, the successor format for the legacy Evaluated Nuclear Data File (ENDF-6). This proposed GNDS hierarchical format contains all the necessary ingredients for inline γ-ray cascade reproduction with event-by-event precision, including continuum–continuum and continuum–discrete transitions following neutron-capture and inelastic neutron scattering reactions. Cascade-event generation based on our approach demonstrates improved energy conservation on an event-by-event basis and permits the use of γ-γ coincidences in applications. This work offers, for the first time, a method to generate neutron-capture and inelastic neutron-scattering γ-ray cascades where energy conservation, correlations, and experimental primaries are fully accounted for.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Propagating neutronic uncertainties for FFTF LOFWOS Test #13

The safety evaluation conducted for licensing a Sodium-cooled Fast Reactor (SFR) may require modeling transients with best-estimate calculation tools that must first be validated against real-world measurements. To provide the community with a valuable benchmarking opportunity for validating SFR analysis tools and methods, the International Atomic Energy Agency (IAEA) initiated a coordinated research project (CRP) in 2018 for the analysis of the Fast Flux Test Facility (FFTF) Loss of Flow Without Scram (LOFWOS) Test #13. The impact of nuclear data uncertainties on neutronics parameters was previously investigated based on the COMMARA-2.0 covariance matrix. Since the transient simulation results are very sensitive to certain reactivity coefficients, it was decided to employ rigorous uncertainty propagation methods to quantify the impact of nuclear data uncertainties on the best-estimate predication of FFTF LOFWOS Test #13. The DAKOTA code is used to propagate neutronic uncertainties through SAS4A/SASSYS-1 transient simulations, while taking into account spatial and reaction-wise correlations within these uncertainties. This study shows that the remaining discrepancies observed between the Argonne National Laboratory (ANL) best-estimate results and the experimental measurements can be partly explained by the uncertainty associated with Gas Expansion Module (GEM) worth, which contributes the majority of the overall nuclear data uncertainty on the output from the FFTF LOFWOS Test #13 transient simulation. This study also confirmed the importance of including spatial and reaction-wise correlations of nuclear data uncertainties on feedback coefficients in the uncertainty propagation to avoid under-estimating their impact during the transient simulations.

42 ENGINEERING↗
DOE OSTI2015

Neutron activation of natural materials in a PWR spectrum: feedback on {sup 116m}In relative γ emission intensities and half-life

During the MAESTRO program, carried out between 2011 and 2014 in MINERVE zero power reactor, common Gen-II and Gen-III light water reactor materials were irradiated. For some of these materials, the decay of their activation products was also measured by γ spectrometry. Initially devoted to the measurement of the integral capture cross section by activation and reactivity-oscillation method, these results can also provide useful information on decay data of various radionuclides. This approach of this experiment led to a common roadmap shared by the Experimental Physics Section and the Henri Becquerel National Laboratory to improve decay data in nuclear data libraries. Results discussed in this paper concern the relative emission intensities of the main γ rays of {sup 116m}In. Six irradiations of samples with various physical forms of {sup nat}In were carried out. Measurements were analyzed using decay data from several evaluations and it is shown that γ ray activities are not consistent. Analyses were carried out to provide new relative γ emission intensities from these measurements. The {sup 116m}In half-life has also been measured and shows a good agreement with existing values. Finally, an overview of the foreseen results on additional decay data from the MAESTRO program is given. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Solar Modulation of {sup 2}H/{sup 1}H and {sup 3}He/{sup 4}He according to Data Obtained in the PAMELA Satellite-Borne Experiment between 2006 and 2014

Preliminary results of measurements of the solar modulation of the isotopic hydrogen and helium content in galactic cosmic rays are presented for the first time. These data were obtained between 2006 and 2014 in the PAMELA international orbital experiment via measuring the time of flight of nuclei in the Pamela scintillation telescope and via analyzing, in the multilayered calorimeter of the PAMELA magnetic spectrometer, ionization losses of nuclei whose rigidity was known from trajectory measurements and which traversed the instrument without undergoing nuclear interactions. The results of these measurements are compared with existing experimental and calculated data, which are quite scanty.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2024

Jemima Plate Characterization

The HEU Jemima plates have been utilized in numerous International Criticality Safety Benchmark Evaluation Project (ICSBEP) evaluations. ICSBEP evaluations are a vital com ponent of nuclear data validation for applications across the nuclear community. In the past, different combinations of mass, caliper measurements, and drawing dimensions have been used to define the HEU Jemima plates in ICSBEP evaluations. Some of these combinations result in large density ranges (17 g/cm 3 to 19 g/cm 3 ) or unrealistic densities (greater than 19 g/cm 3 ). Additionally, oxidation of the plates has led to questioning the flatness of each plate, which can add additional gaps in experiment configurations. This work proposes a standardized method of defining the commonly used HEU Jemima plates across different experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Semi-automatic image annotation using 3D LiDAR projections and depth camera data

Efficient image annotation is necessary to utilize deep learning object recognition neural networks in nuclear safeguards, such as for the detection and localization of target objects like nuclear material containers (NMCs). This capability can help automate the inventory accounting of different types of NMCs within nuclear storage facilities. The conventional manual annotation process is labor-intensive and time-consuming, hindering the rapid deployment of deep learning models for NMC identifications. This paper introduces a novel semi-automatic method for annotating 2D images of nuclear material containers (NMCs) by combining 3D light detection and ranging (LiDAR) data with color and depth camera images collected from a handheld scan system. The annotation pipeline involves an operator manually marking new target objects on a LiDAR-generated map, and projecting these 3D locations to images, thereby automatically creating annotations from the projections. The semi-automatic approach significantly reduces manual efforts and the expertise in image annotation that is required to perform the task, allowing deep learning models to be trained on-site within a few hours. The paper compares the performance of models trained on datasets annotated through various methods, including semi-automatic, manual, and commercial annotation services. The evaluation demonstrates that the semi-automatic annotation method achieves comparable or superior results, with a mean average precision (mAP) above 0.9, showcasing its efficiency in training object recognition models. Additionally, the paper explores the application of the proposed method to instance segmentation, achieving promising results in detecting multiple types of NMCs in various formations.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2014

A Longitudinal Evaluation of Partial Lung Irradiation in Mice by Using a Dedicated Image-Guided Small Animal Irradiator

Purpose: In lung cancer radiation therapy, the dose constraints are determined mostly by healthy lung toxicity. Preclinical microirradiators are a new tool to evaluate treatment strategies closer to clinical irradiation devices. In this study, we quantified local changes in lung density symptomatic of radiation-induced lung fibrosis (RILF) after partial lung irradiation in mice by using a precision image-guided small animal irradiator integrated with micro-computed tomography (CT) imaging. Methods and Materials: C57BL/6 adult male mice (n=76) were divided into 6 groups: a control group (0 Gy) and groups irradiated with a single fraction of 4, 8, 12, 16, or 20 Gy using 5-mm circular parallel-opposed fields targeting the upper right lung. A Monte Carlo model of the small animal irradiator was used for dose calculations. Following irradiation, all mice were imaged at regular intervals over 39 weeks (10 time points total). Nonrigid deformation was used to register the initial micro-CT scan to all subsequent scans. Results: Significant differences could be observed between the 3 highest (>10 Gy) and 3 lowest irradiation (<10 Gy) dose levels. A mean difference of 120 ± 10 HU between the 0- and 20-Gy groups was observed at week 39. RILF was found to be spatially limited to the irradiated portion of the lung. Conclusions: The data suggest that the severity of RILF in partial lung irradiation compared to large field irradiation in mice for the same dose is reduced, and therefore higher doses can be tolerated.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2021

Pulsed Thermal Tomography Nondestructive Examination of Additively Manufactured Reactor Materials and Components. Third Annual Progress Report

Additive manufacturing (AM) of high-strength corrosion resistance alloys for nuclear energy applications, such as stainless steel and Inconel, is currently based on laser powder bed fusion (LPBF) process. Some of the challenges with using LPBF method for nuclear manufacturing include the possibility of introducing pores into metallic structures. Probability of crack initiation at the pore depends on size, shape, and orientation of the defect. Pulsed Infrared Thermography Imaging (PIT) provides a capability for non-destructive evaluation (NDE) of sub-surface defects in arbitrary size structures. The PIT method is based on recording material surface temperature transients with infrared (IR) camera following thermal pulse delivered on material surface with flash light. The PIT method has advantages for NDE of actual AM structures because the method involves one-sided non-contact measurements and fast processing of large sample areas captured in one image. Following initial qualification of an AM component for deployment in a nuclear reactor, a PIT system can also be used for in-service nondestructive evaluation (NDE) applications. In this report, we describe recent progress in enhancing PIT capabilities in detecting microscopic subsurface defects in metals, and classifying shapes and orientation of pores in thermal images. For detection of microscopic defects in PIT imaging data, we have developed Spatial Temporal Denoised Thermal Source Separation (STDTSS) unsupervised machine learning (ML) image processing algorithm. We show that flat bottom hole (FBH) defects as small as 200µm in SS316 and IN718 specimens, can be detected with STDTSS algorithm. To the best of our knowledge, these are the smallest detected defects which are reported in literature. For classification of defects shapes, we have previously developed thermal tomography (TT) algorithm to obtain depth reconstructions of material defects from data cube of sequentially recorded surface temperatures. However, interpretation of TT images is non-trivial because of blurring with increasing depth. To address this challenge, we have developed a deep learning convolutional neural network (CNN) to classify size and orientation subsurface defects in simulated TT images.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

A parameterization method and application in breast tomosynthesis dosimetry

Purpose: To present a parameterization method based on singular value decomposition (SVD), and to provide analytical parameterization of the mean glandular dose (MGD) conversion factors from eight references for evaluating breast tomosynthesis dose in the Mammography Quality Standards Act (MQSA) protocol and in the UK, European, and IAEA dosimetry protocols.Methods: MGD conversion factor is usually listed in lookup tables for the factors such as beam quality, breast thickness, breast glandularity, and projection angle. The authors analyzed multiple sets of MGD conversion factors from the Hologic Selenia Dimensions quality control manual and seven previous papers. Each data set was parameterized using a one- to three-dimensional polynomial function of 2–16 terms. Variable substitution was used to improve accuracy. A least-squares fit was conducted using the SVD.Results: The differences between the originally tabulated MGD conversion factors and the results computed using the parameterization algorithms were (a) 0.08%–0.18% on average and 1.31% maximum for the Selenia Dimensions quality control manual, (b) 0.09%–0.66% on average and 2.97% maximum for the published data by Dance et al. [Phys. Med. Biol. 35, 1211–1219 (1990); ibid. 45, 3225–3240 (2000); ibid. 54, 4361–4372 (2009); ibid. 56, 453–471 (2011)], (c) 0.74%–0.99% on average and 3.94% maximum for the published data by Sechopoulos et al. [Med. Phys. 34, 221–232 (2007); J. Appl. Clin. Med. Phys. 9, 161–171 (2008)], and (d) 0.66%–1.33% on average and 2.72% maximum for the published data by Feng and Sechopoulos [Radiology 263, 35–42 (2012)], excluding one sample in (d) that does not follow the trends in the published data table.Conclusions: A flexible parameterization method is presented in this paper, and was applied to breast tomosynthesis dosimetry. The resultant data offer easy and accurate computations of MGD conversion factors for evaluating mean glandular breast dose in the MQSA protocol and in the UK, European, and IAEA dosimetry protocols. Microsoft Excel™ spreadsheets are provided for the convenience of readers.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

STRUCTURAL MODELING TO SUPPORT POST-YIELD ACCEPTANCE CRITERIA FOR SPENT NUCLEAR FUEL CLADDING

Spent nuclear fuel (SNF) is evaluated for structural failure during storage and transportation scenarios. The U.S. Department of Energy’s Spent Fuel and Waste Science and Technology (SFWST) program has sponsored significant research in quantifying mechanical loads on SNF during storage and transportation scenarios using experimental and modeling methods. The SFWST program has also performed significant research on measuring the mechanical behavior of irradiated SNF as defueled cladding segments and cladding with fuel pellets to measure composite behavior. This paper considers some of the key material data from the Sibling Pin testing and uses structural modeling and analysis methods that have been informed by testing to consider post-yield acceptance criteria for SNF cladding structural analysis. Test data published by Oak Ridge National Laboratory (ORNL) and Pacific Northwest National Laboratory (PNNL) are the foundation for informing the material behavior of the models developed in this study. In particular, four-point bend (4PB) tests of fueled and defueled cladding segments provide significant information about the bending failure mode of SNF. ORNL’s 4PB test data is on fueled cladding segments, so the composite behavior of SNF is demonstrated. This paper describes PNNL’s coincident beam model that was developed to approximate the composite behavior of SNF. This paper also presents PNNL’s structural dynamic finite element models of a cask tip-over scenario, which is predicted to cause the strongest mechanical loads on SNF of all postulated storage and transportation scenarios. SNF bending loads predicted in the cask tip-over scenario and cladding acceptance criteria beyond yield are considered, with justification based on the Sibling Pin test data. ASME Boiler and Pressure Vessel code stress intensity limits are also considered. The ultimate goal of this work is to aid in the justification of structural acceptance criteria for SNF cladding beyond the cladding’s irradiated yield strength for use in structural analysis of all storage and transportation scenarios.

Klymyshyn, Nicholas A.↗
DOE OSTI2019

Verification of R-matrix calculations for charged-particle reactions in the resolved resonance region for the 7 Be system

R-matrix theory is used to present nuclear reactions in the resolved resonance region. It uses information on bound states and low energy resonances to accurately parametrize cross sections on the resonances as well as the non-resonant background. Since the seminal work of Lane and Thomas (1958), the method has been widely used to analyze experimental cross-section data in a broad range of fields spanning nuclear reaction dynamics, nuclear astrophysics, ion beam analysis and their applications. Different R-matrix codes have been developed and used in these different applications with very little communication among the developers or practitioners on the capabilities, achievements or limitations of the codes. A limited comparison among three R-matrix codes on neutron-induced reactions was performed by the International Atomic Energy Agency (IAEA) International Evaluation of Neutron Cross Section Standards project (2007). Since then, significant progress has been made in their implementation of the R-matrix algorithms, and R-matrix codes have enhanced capabilities. In this paper we introduce, for the first time, the results of a comprehensive effort to verify the most widely used R-matrix codes in the various fields of nuclear science and applications: AMUR, AZURE2, CONRAD, EDA, FRESCO, GECCCOS, and SAMMY. In addition to the description of the capabilities of the codes and their specifications, we discuss the results of a joint exercise which was coordinated by the International Atomic Energy Agency. The aim of the exercise was to compare calculations of charged-particle reaction cross sections for the light composite system 7Be. The calculations were performed by the codes using identical input R-matrix parameters and other specifications and were limited to charged-particle channels.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS1990

Risk assessment methodologies for target fragments produced in high-energy nucleon reactions

Collisions of energetic protons with target nuclei yield nuclear products of large linear energy transfer, and it is usually assumed that the energy of the heavy products is absorbed locally with a quality factor of 20. Past methods of risk assessment have relied on the Bertini (1969) model to evaluate the nuclear reaction products. If the Q(F) of 20 is correct, then a significant underestimate of biological risk from the nuclear products could result from the use of Bertini's model. On the other hand, a Q(F) of 20 is probably too large for some of the lighter fragments, leaving prior risk estimates as possibly correct or even conservative. This issue is discussed here in the context of the recently proposed nuclear data set of Wilson et al. (1989).

Shinn, J. L.↗
DOE OSTI2025

The Technical, Economic, Risk, and Adoption Assessment for Evaluating Work Reduction Opportunities in the Nuclear Industry

Automation and cost-saving initiatives, such as process automation with advances in artificial intelligence, are gaining traction in modernization efforts across the nuclear industry. As these innovations are increasingly adopted, it becomes crucial to evaluate their impacts comprehensively. Various technical and economic attributes, along with risk and human readiness factors, must be achieved to ensure that innovative projects enabling automation and modernization are successful. However, no systematic or integrated framework exists that allows plants to evaluate these innovative projects. To address this gap, the Technical, Economic, Risk, and Adoption (TERA) assessment offers a structured method to evaluate innovative technologies, ensuring solutions meet both operational and safety standards. The TERA framework integrates the disparate perspectives to assess modernization opportunities in nuclear operations. It combines qualitative and quantitative models to evaluate the relationship between performance and business impacts, while also enabling continuous re-evaluation during project development. This approach helps plant owners identify high-priority opportunities, optimize cost savings, and minimize risks, ensuring projects remain on track to achieve desired returns. By providing a comprehensive, systematic methodology, TERA enables informed, data-driven decisions that support successful modernization efforts, enhancing efficiency, safety, and cost savings across nuclear operations. This paper explains the TERA framework and its benefits for the nuclear industry.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Evaluation of a novel collimator for molecular breast tomosynthesis

Here, this study investigated a novel gamma camera for molecular breast tomosynthesis (MBT), which is a nuclear breast imaging method that uses limited angle tomography. The camera is equipped with a variable angle, slant-hole (VASH) collimator that allows the camera to remain close to the breast throughout the acquisition. The goal of this study was to evaluate the spatial resolution and count sensitivity of this camera and to compare contrast and contrast-to-noise ratio (CNR) with conventional planar imaging using an experimental breast phantom. Methods The VASH collimator mounts to a commercial gamma camera for breast imaging that uses a pixelated (3.2 mm), 15 × 20 cm NaI crystal. Spatial resolution was measured in planar images over a range of distances from the collimator (30-100 mm) and a range of slant angles (–25° to 25°) using 99m Tc line sources. Spatial resolution was also measured in reconstructed MBT images including in the depth dimension. The images were reconstructed from data acquired over the -25° to 25° angular range using an iterative algorithm adapted to the slant-hole geometry. Sensitivity was measured over the range of slant angles using a disk source. Measured spatial resolution and sensitivity were compared to theoretical values. Contrast and CNR were measured using a breast phantom containing spherical lesions (6.2 mm and 7.8 mm diameter) and positioned over a range of depths in the phantom. The MBT and planar methods had equal scan time, and the count density in the breast phantom data was similar to that in clinical nuclear breast imaging. The MBT method used an iterative reconstruction algorithm combined with a postreconstruction Metz filter. Results The measured spatial resolution in planar images agreed well with theoretical calculations over the range of distances and slant angles. The measured FWHM was 9.7 mm at 50 mm distance. In reconstructed MBT images, the spatial resolution in the depth dimension was approximately 2.2 mm greater than the other two dimensions due to the limited angle data. The measured count sensitivity agreed closely with theory over all slant angles when using a wide energy window. At 0° slant angle, measured sensitivity was 19.7 counts sec -1 μ Ci -1 with the open energy window and 11.2 counts sec -1 μ Ci -1 with a 20% wide photopeak window (126 to 154 keV). The measured CNR in the MBT images was significantly greater than in the planar images for all but the lowest CNR cases where the lesion detectability was extremely low for both MBT and planar. The 7.8 mm lesion at 37 mm depth was marginally detectable in the planar image but easily visible in the MBT image. The improved CNR with MBT was due to a large improvement in contrast, which out-weighed the increase in image noise. Conclusion The spatial resolution and count sensitivity measurements with the prototype MBT system matched theoretical calculations, and the measured CNR in breast phantom images was generally greater with the MBT system compared to conventional planar imaging. These results demonstrate the potential of the proposed MBT system to improve lesion detection in nuclear breast imaging.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2014

Absolute determination of photoluminescence quantum efficiency using an integrating sphere setup

An integrating sphere-based setup to obtain a quick and reliable determination of the internal quantum efficiency of strongly scattering luminescent materials is presented. In literature, two distinct but similar measurement procedures are frequently mentioned: a “two measurement” and a “three measurement” approach. Both methods are evaluated by applying the rigorous integrating sphere theory. It was found that both measurement procedures are valid. Additionally, the two methods are compared with respect to the uncertainty budget of the obtained values of the quantum efficiency. An inter-laboratory validation using the two distinct procedures was performed. The conclusions from the theoretical study were confirmed by the experimental data.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Passive microrheology of normal and cancer cells after ML7 treatment by atomic force microscopy

Mechanical properties of living cancer and normal thyroidal cells were investigated by atomic force microscopy (AFM). Cell mechanics was compared before and after treatment with ML7, which is known to reduce myosin activity and induce softening of cell structures. We recorded force curves with extended dwell time of 6 seconds in contact at maximum forces from 500 pN to 1 nN. Data were analyzed within different frameworks: Hertz fit was applied in order to evaluate differences in Young’s moduli among cell types and conditions, while the fluctuations of the cantilever in contact with cells were analyzed with both conventional algorithms (probability density function and power spectral density) and multifractal detrended fluctuation analysis (MF-DFA). We found that cancer cells were softer than normal cells and ML7 had a substantial softening effect on normal cells, but only a marginal one on cancer cells. Moreover, we observed that all recorded signals for normal and cancer cells were monofractal with small differences between their scaling parameters. Finally, the applicability of wavelet-based methods of data analysis for the discrimination of different cell types is discussed.

62 RADIOLOGY AND NUCLEAR MEDICINE↗