bluuebunny/crossref_metadata_2025 huggingface dataset:
CODE VERIFICATION OF MPACT USING GANAPOL CRITICAL ROD BENCHMARK
Authors: Wang Jipu , Martin William R , Collins Benjamin S
Published: February 2021

This work is dedicated to the code verification of MPACT, which is developed under the Consortium for Advanced Simulation of Light Water Reactors by the University of Michigan and Oak Ridge National Laboratory, where the numerical solution is compared to the reference solution of a benchmark problem with a known analytical solution. In this work, Benchmark Problem 3.4 in Barry Ganapol’s benchmark book was chosen as an MOC code verification test problem. Problem 3.4 is a bare cylinder of infinite height, which is an excellent benchmark problem for 2D MOC. To ensure that this benchmark problem exercised the same code as typically used by MPACT, the bare rod configuration was surrounded by a bounding box filled with a non-scattering material. To avoid implementing a critical rod search in the MPACT code, the MPACT analysis was performed using cross sections that yielded the given c, the average number of secondary neutrons per collision, and a rod radius that was the corresponding critical rod radius. MPACT agreed with all cases to within a few pcm. The convergence behavior was studied. The results show a 2nd order radial convergence, consistent with flat-source approximation. The convergence curves with respect to ray spacing and polar angle quadrature set order were also obtained. The other quantity of interest tabulated for Problem 3.4 was the radial distribution of the scalar flux. Two configurations were analyzed, and the resultant radial flux profiles agreed very well with the tabulated results. The verification of the production neutronics code MPACT has been augmented by the addition of the analytical solutions for an infinite cylinder from the Ganapol benchmark book. These test cases can be included in the regression suite for MPACT.
IDENTIFICATION OF REACTOR PHYSICS BENCHMARKS FOR NUCLEAR DATA TESTING: TOOLS AND EXAMPLES
Authors: Hill I.
Published: February 2021

Measurements of reactor physics quantities aimed at identifying the reactivity worth of materials, spectral ratios of cross-sections, and reactivity coefficients have ensured reactor physics codes can accurately predict nuclear reactor systems. These measurements were critical in the absence of sufficiently accurate differential data, and underpinned the need for experiments through the 50s, 60s, 70s and 80s. Data from experimental campaigns were routinely incorporated into nuclear data libraries either through changes to general nuclear data libraries, or more commonly in the local libraries generated by a particular institution or consortium interested in accurately predicting a specific nuclear system (e.g. fast reactors) or parameters (e.g. fission gas release, yields). Over the last three decades, the model has changed. In tandem access to computing power and monte carlo codes rose dramatically. The monte carlo codes were well suited to computing k-eff, and owing to the availability of high quality criticality benchmarks and these benchmarks were increasing used to test the nuclear data. Meanwhile, there was a decline in the production of local libraries as new nuclear systems were not being built, and the existing systems were considered adequately predicted. The cost-to-benefit ratio of validating new libraries relative to their improved prediction capability was less attractive. These trends have continued. It is widely acknowledged that the checking of new nuclear data libraries is highly skewed towards testing against criticality benchmarks, ignoring many of the high quality reactor physics benchmarks during the testing and production of general-purpose nuclear data libraries. However, continued increases in computing power, methodology (GPT), and additional availability reactor physics experiments from sources such as the International Handbook of Evaluated Reactor Physics Experiments should result in better testing of new libraries and ensured applicability to a wide variety of nuclear systems. It often has not. Leveraging the wealth of historical reactor physics measurements represents perhaps the simplest way to improve the quality of nuclear data libraries in the coming decade. Resources at the Nuclear Energy Agency can be utilized to assist in interrogating available identify benchmarks in the reactor physics experiments handbook, and expediting their use in verification and validation. Additionally, high quality experimental campaigns that should be examined in validation will be highlighted to illustrate potential improvements in the verification and validation process.
VERIFICATION OF THE DIFFUSION AND TRANSPORT SOLVERS WITHIN DIF3D FOR 3D HEXAGONAL GEOMETRIES
Authors: Nelson A. G. , Smith M. A. , Heidet F.
Published: February 2021

The DIF3D code (DIFfusion 3D) has been a workhorse of fast reactor analysis work at Argonne National Laboratory for over 40 years. DIF3D was primarily built in the late 1970s as a three-dimensional multigroup diffusion equation solver operating on semi-structured grid geometries. In the mid-1990s, transport capabilities needed for high-leakage reactor configurations were added to DIF3D with the variational anisotropic nodal transport approach. Recent reactor design activities at Argonne are requiring that a thorough verification of the Argonne Reactor Computation (ARC) codes be performed. With DIF3D being central to the entire ARC system, the verification efforts are focused on the 3D Cartesian, 3D triangular, and 3D hexagonal core geometry options of DIF3D. Validation activities, while needed for the ongoing design activities at Argonne, are handled at a project-specific level.

This paper summarizes the verification work so far on the forward and adjoint forms of the fixed source, inhomogeneous fixed source, and k-eigenvalue steady state transport and diffusion equations as implemented specifically for 3D triangular and hexagonal geometries in DIF3D. Since analytic solutions of the neutron diffusion and transport equations are either limited in scope or not possible, this verification required multiple tiers of problems unique to each solver and geometry type, each testing features independent and complementary arguments for why this separate testing of functionalities is acceptable. This separate testing was also supplemented with a high-level integral check of each the diffusion and transport capabilities and applicable geometries.
ASSESSMENT OF HEAT TRANSFER IN FUEL CHANNEL OF INDIAN PHWR UNDER POSTULATED LARGE BREAK LOSS OF COOLANT ACCIDENT: EXPERIMENTAL AND NUMERICAL STUDY
Authors: Ajay Ketan , Kumar Ravi , Gupta Akhilesh
Published: February 2021

The behaviour of the channel under postulated large break LOCA scenario had been a prime safety concern. The radiative heat transfer is predominant in a channel when the convective cooling environment is marred. The estimation of temperature distribution in the fuel pins at elevated temperature is essential from the point of view of hydrogen gas generation and release of fission products. In this paper, the thermal characteristics of a channel for Indian PHWR under critical break failure is studied using experimental and numerical techniques. The experiment is carried out on an Indian PHWR having a fuel bundle of 37-fuel elements. The temperature profiles for different parts of the simulated channel comprising of fuel pins, PT and CT are obtained under steady condition. The numerical analysis is also performed using ANSYS Fluent 19.0. From the study, it is found that there is a significant radial temperature gradient in the fuel bundle from the center ring to the outer ring. Also, no significant circumferential temperature gradient is observed in the fuel bundle, PT and CT.
EXTENSION OF XENON OSCILLATIONS SAFETY MARGINS USING WEAKLY NONLINEAR STABILITY ANALYSIS
Authors: Kastin Nir , Meron Ehud , Kolin Assaf , Kinast Shai
Published: February 2021

Weakly nonlinear stability analysis is applied to study xenon oscillations in nuclear reactors using the approach of multiple time-scales method. This approach allows to characterize the dynamics of the system beyond the Hopf instability point. It provides important insight on the characteristics of the oscillations, namely if they diverge with time, or converge into a bounded limit cycle. Detailed derivation of the amplitude equation is presented. This equation is used to identify parameter ranges of bounded periodic oscillations, which may be allowed for safe operation. The influence of neutron generation time and the power feedback coefficients on the amplitude of limit cycles, as well as on convergence times, is discussed. The described method may be used to extend the safety margins required to prevent xenon unstable oscillations in reactor cores.
MAXIMISING DISCHARGE BURNUP IN AN OPEN CYCLE MOLTEN SALT REACTOR
Authors: Constable Charlie , Lindley Ben , Parks Geoff
Published: February 2021

This paper discusses work done to find an estimate of the maximum achievable discharge burnup in an open cycle molten salt reactor (MSR). An in-development deterministic code (WIMS11) is used to create a model of a simple generic MSR, and the methodology employed is discussed. Some experimentation is done with regards to the internal set-up of the ‘unit cells’ within the core, which shows there is a strong link between this geometry and the achievable burnup. Work is done to quantify the effects of removing volatile fission products and implementing a two-batch refuelling scheme. Finally, an optimization process is carried out whereby the optimal proportion of graphite moderator within the core is found which balances power across the regions while maximising discharge burnup. Two fuels are tested, one which carries only 235U and 238U, and another which also carries 232Th. It is found that the maximum achievable discharge burnup is approximately 155 MWd/kg, which is considerably higher than modern PWRs, despite a lower enrichment and only two batches of fuel being used.
MINIMIZING CRUD DEPOSITION THROUGH OPTIMIZATION OF ASSOCIATED PARAMETERS
Authors: Andersen Brian , Hou Jason , Kropaczek Dave
Published: February 2021

A strong correlation exists between subcooled boiling in assembly subchannels and CRUD deposition. In this work a genetic algorithm is used to optimize a 17 x 17 PWR fuel assembly to have minimized subcooled boiling, minimized peak kinf, and maximized end of cycle kinf. Optimization of these parameters act as a surrogate for the optimization of CRUD deposition in fuel assemblies due to their strong correlation. Subcooled boiling, measured by vapor void in a sub channel, and values of kinf, are calculated using VERA-CS. Due to the high computational cost of VERA-CS, artificial neural networks are used as surrogate models to VERA-CS in order for the optimization to be performed in a timely manner making design work possible. Two neural networks were trained using a training library of 1200 randomly generated assembly designs and a validation library of 100 assembly designs evaluated using VERA-CS. The combination of neural networks and genetic algorithms formed an extremely fast optimization algorithm capable of evaluating designed a set of optimized pin lattices in a matter of minutes. The optimization showed a clear reduction in vapor void in the optimization solutions. This provides a proof of principle that complex phenomena requiring coupled, Multiphysics calculations, such as CRUD deposition, may be optimized.
INTERNET REACTOR LABORATORY: A NUCLEAR REACTOR PHYSICS AND ENGINEERING EDUCATION MODALITY FOR THE 21st CENTURY
Authors: Hawari A. I. , Lassell S. A. , Gibson G. I.
Published: February 2021

The “second generation” Internet Reactor Laboratory (IRL) has been installed and is currently functional at the PULSTAR reactor of the Nuclear Reactor Program (NRP) at North Carolina State University (NCSU). The NRP has offered IRL based reactor physics experiments and training to external academic institutions and organizations since 2004. In 2016, the United States and the Republic of Vietnam entered into an Administrative Arrangement under Section 123 of the Atomic Energy Act, with a goal of enhancing nuclear training and education for Vietnamese engineering students. Funding was provided by the DOE to upgrade the PULSTAR IRL infrastructure to state-of-the-art remote data acquisition and video teleconferencing capabilities supporting content sharing across multiple remote user platforms. Subsequently, a Cisco Telepresence and LabView reactor data acquisition system were installed and commissioned at the PULSTAR facility. Using this “second generation” system, and during IRL sessions, up to 100 remote students may interact with the control room staff via the Telepresence system using their mobile device or PC, and view 130 channels of live reactor telemetry via the remote client LabView application while recording the data for subsequent review and analysis. LabView graphical user interfaces (GUI) are provided for each IRL module, and incorporate embedded supporting documents such as schematics and diagrams that aid in illustrating associated technical details and concepts. Consequently, the educational experience offered to remote students and trainees is enhanced and assured to be of the same quality as offered to onsite students/trainees at NCSU.
UNCERTAINTY CHARACTERIZATION FRAMEWORK FOR STEADY-STATE AND TRANSIENT NEUTRONICS SIMULATIONS OF A CANDU REACTOR
Authors: Abdel-Khalik Hany S. , Trottier Alexandre , Serghiuta Dumitru , Huang Dongli
Published: February 2021

This paper reports on the development and testing of a comprehensive few-group cross section input uncertainty library for the NESTLE-C nodal diffusion-based nuclear reactor core simulator. This library represents the first milestone of a first-of-a-kind framework for the integrated characterization of uncertainties in steady-state and transient CANDU reactor simulations. The objective of this framework is to propagate, prioritize and devise a mapping capability for uncertainties in support of model validation of best-estimate calculations. A complete framework would factor both input and modeling uncertainty contributions. The scope of the present work is limited to the propagation of multi-group cross-section uncertainties through lattice physics calculations down to the few-group format, representing the input to the NESTLE-C core simulator, and finally to core responses of interest.
NUCLEAR DATA SENSITIVITY/UNCERTAINTY PRE-ANALYSIS OF FNG WCLL FUSION BENCHMARK
Authors: Kodeli Ivan A. , Angelone Maurizio , Flamini Davide
Published: February 2021

To assure tritium self-sufficiency in future fusion reactors such as DEMO the accuracy of TRP calculations has to be demonstrated within the design uncertainties. A new neutronics experiment representing a mock-up of the Water Cooled Lithium Lead (WCLL) Test Blanket Module (TBM) is under preparation at the Frascati neutron generator (FNG) with the objective to provide an experimental validation of accuracy of nuclear data and neutron transport codes for the tritium production rate (TPR) calculations. The mock-up will consist of LiPb bricks, EUROFER plates and Perspex substituting water. The mock-up will be irradiated by 14 MeV neutrons at the FNG facility, and the TPR and detector reaction rates will be measured using Li2CO3 pellets and activation foils placed at different positions up to about 55 cm inside the mock-up. Computational pre-analyses for the design of the WCLL neutronics experiment using the SUSD3D sensitivity/uncertainty (S/U) code system is described and compared with the results of some similar FNG experiments performed in the past, in particular the FNG HCPB Tritium Breeder Module Mock-up (2005) and FNG-HCLL Tritium Breeder Module Mock-up (2009). The objective of the pre-analysis is to provide the calculated nuclear responses including the uncertainties due to the uncertainties in nuclear data and thus contributes to the optimisation of the design of the experimental set-up.
UNCERTAINTY QUANTIFICATION IN STEADY STATE SIMULATIONS OF A MOLTEN SALT SYSTEM USING POLYNOMIAL CHAOS EXPANSION ANALYSIS
Authors: Santanoceto Mario , Tiberga Marco , Perkó Zoltán , Dulla Sandra , Lathouwers Danny
Published: February 2021

Uncertainty Quantification (UQ) of numerical simulations is highly relevant in the study and design of complex systems. Among the various approaches available, Polynomial Chaos Expansion (PCE) analysis has recently attracted great interest. It belongs to nonintrusive spectral projection methods and consists of constructing system responses as polynomial functions of the stochastic inputs. The limited number of required model evaluations and the possibility to apply it to codes without any modification make this technique extremely attractive. In this work, we propose the use of PCE to perform UQ of complex, multi-physics models for liquid fueled reactors, addressing key design aspects of neutronics and thermal fluid dynamics. Our PCE approach uses Smolyak sparse grids designed to estimate the PCE coefficients. To test its potential, the PCE method was applied to a 2D problem representative of the Molten Salt Fast Reactor physics. An in-house multi-physics tool constitutes the reference model. The studied responses are the maximum temperature and the effective multiplication factor. Results, validated by comparison with the reference model on 103 Monte-Carlo sampled points, prove the effectiveness of our PCE approach in assessing uncertainties of complex coupled models.
A MESOSCOPIC OXIDE FUEL CLUSTERING AND ITS GLOBAL PERFORMANCE
Authors: Shaginyan Ruben , Kolesov Valery , Ivanov Evgeny
Published: February 2021

Transient fuel behavior in a Light Water-cooled Reactor core depends on nuclear properties (Doppler broadening, moderation ratio, and, sometimes, neutron gas temperature etc.) and on variations of thermal-physics parameters (temperature distributions, fuel elongation and moderator density).

Usually, in a rough reactor analysis one ignores the very details of temperature distributions largely staying in a frame of so-called adiabatic assumptions (when temperature and density distribution are changing in sync keeping given spatial shapes).

In majority of practical applications the radially distributed temperature fields are represented as monotonically smeared ones as if fissile and other materials are homogeneously mixed.

Moreover, no one measurement technique allows counting precise correlation between reactivity feedback and in-pellet temperature and materials space-time distributions.

However, if fuel is made of Mixed Oxide Plutonium-Uranium compound the behavior of Light Water Reactor would be impacted by an appearance of Pu-rich agglomerates that could be large enough to change physical processes.

In such case the fuel reacts on power and temperature variations no more as a homogeneous but a heterogeneous media (on a mesoscopic scale, of course).

It leads to changes in a fission product distributions, a fission gas release and, even, to an appearance of multiple components in a Fuel Temperature Coefficient and in a Power Reactivity feedback.

These components would depend non-linearly on power, power rate and on some details of a heat transfer.

This paper is the only first step of a broad research program where we are estimating the relevant phenomena just by an order of magnitude.
EXPLORING TRANSIENT, NEUTRONIC, REDUCED-ORDER MODELS USING DMD/POD-GALERKIN AND DATA-DRIVEN DMD
Authors: Elzohery Rabab , Roberts Jeremy
Published: February 2021

There is growing interest in the development of transient, multiphysics models for nuclear reactors and analysis of uncertainties in those models. Reduced-order models (ROMs) provide a computationally cheaper alternative to compute uncertainties. However, the application of ROMs to transient systems remains a challenging task. Here, a 1-D, twogroup, time-dependent, diffusion model was used to explore the potential of three different ROMs: the intrusive POD-Galerkin and DMD-Galerkin methods and the purely datadriven DMD. For the problem studied, POD-Galerkin exhibited by far the best accuracy and was selected for further application to uncertainty propagation. Perturbations were introduced to the initial condition and to the cross-section data. A greedy-POD sampling procedure was used to construct a reduced space that captured much of the variation in the uncertain these parameters. Results indicate that relatively few samples of the uncertain parameters are needed to produce a basis for POD-Galerkin that leads to distributions of the quantities of interest that match well with those obtained from the full-order model using brute-force, forward sampling.
ON THE IMPORTANCE OF SECOND-ORDER RESPONSE SENSITIVITIES TO NUCLEAR DATA IN REACTOR PHYSICS UNCERTAINTY ANALYSIS
Authors: Cacuci Dan G.
Published: February 2021

This invited keynote presentation compares the relative importance of 1st-order versus 2nd-order sensitivities of the leakage response of an OECD/NEA benchmark (polyethylene-reflected plutonium sphere) to the nuclear data characterizing this benchmark. The imprecisely known parameters underlying the neutron transport computational model for this benchmark include 180 group-averaged total microscopic cross sections, 21600 group-averaged scattering microscopic cross sections, 60 parameters describing the fission process, 30 parameters describing the fission spectrum, 10 parameters describing the system’s sources, and 6 isotopic number densities. Thus, this benchmark comprises 21886 1st-order sensitivities of the leakage response with respect to the model parameters, and 478,996,996 2nd-order sensitivities, of which 239,509,441 are distinct. The exact deterministic computation of all of these 1st- and 2nd-order sensitivities was made possible by the application of the Second-Order Adjoint Sensitivity Analysis Methodology (2nd-ASAM) developed by Cacuci. Thousands (out of the 32 400 elements) of the 2nd-order sensitivities of the leakage response with respect to the total cross sections turned out to be significantly larger than the largest corresponding 1st-order sensitivities, contrary to some previously held beliefs in the reactor physics community. Hence, it will be shown that neglecting the 2nd-order sensitivities to total cross sections would cause very large non-conservative errors by under-reporting the response’s variance and expected value. The 2nd-order sensitivities also cause the response distribution to be skewed towards positive values relative to the expected value, which, in turn, is significantly larger than the computed value of the leakage response. The result presented in this paper also underscore the need for obtaining reliable cross section covariance data, which are not available at this time.
ESTIMATED CRITICALITY LOWER-LIMIT MULTIPLICATION FACTOR OF LOW-ENRICHED URANIUM DIOXIDE-CONCRETE SYSTEM USING THE BOOTSTRAP METHOD
Authors: Hayashi Takuto , Nishioka Fuga , Endo Tomohiro , Yamamoto Akio
Published: February 2021

The present paper aims to evaluate the estimated criticality lower-limit multiplication factor of fuel debris in a form of uranium dioxide-concrete mixture for a study of criticality control on the fuel debris generated through the molten core concrete interaction in a severe accident of a light water reactor. The estimated criticality lower-limit multiplication factor is evaluated using the bootstrap method where the assumption of the normal distribution is not necessary. In addition, it is calculated taking into account correlation coefficients that represent the degrees of neutronic similarity between the target system and benchmark critical experiment systems, experimental uncertainties of benchmark data, and statistical uncertainties of calculated effective multiplication factor by a continuous energy Monte Carlo code. This paper shows that the estimated criticality lower-limit multiplication factor using the bootstrap method can be comparable with a baseline upper-subcritical-limit which is evaluated by Whisper-1.1 without margins of subcriticality for uncertainties from nuclear covariance data and undetected errors in software.
CRITICALITY SAFETY ANALYSIS OF GBC-32 SPENT FUEL CASK WITH IMPROVED NEUTRON ABSORBER CONCEPT
Authors: Lovecký Martin , Závorka Jiří , Jiřičková Jana , Škoda Radek
Published: February 2021

Higher enrichment of nuclear fuel along the manufacturing limit of boron content in steel and aluminum alloys represents a significant challenge in designing spent fuel transport and storage facilities. One possible solution for spent fuel pools and casks is the burnup credit method that allows for decreasing very high safety margins associated with fresh fuel assumption in spent fuel facilities. An alternative solution based on placing neutron absorber material directly into the fuel assembly is proposed here. A neutron absorber permanently fixed in guide tubes decreases system reactivity more efficiently than absorber sheets between the assemblies. The efficiency of the newly proposed concept is demonstrated on the criticality safety analysis of the GBC-32 spent fuel cask. Absorber rods from 8 different elements are placed within Westinghouse OFA 17x17 guide tubes. Currently used boron is a good option because of high absorption cross section, low atomic mass and chemical compatibility with various alloys. Alternative options (e.g., Sm, Eu, Gd, Dy, Hf, Re, Ir) are based on very good absorbers that do not require alloy compatibility since the absorbers can be placed inside zirconium or steel cladding. Because of high efficiency of the newly proposed absorber concept, boron content in BORAL sheets can be decreased to more competitive economics. Moreover, fuel assembly pitch is investigated in order to change cask wall inner diameter that will result in lower material consumption for the cask wall with the same shielding thickness.
CORE DESIGN STUDY FOR A SMALL MODULAR BOILING WATER REACTOR
Authors: Wibisono Andhika Feri , Shwageraus Eugene
Published: February 2021

The hybrid Small Modular Boiling Water Reactor (SMBWR) is a new conceptual design of BWR-type SMR. The main features of SMBWR include a natural circulation loop in its coolant recirculation system and external superheaters system integrated into the steam cycle. A full core analysis of SMBWR is performed with the nodal diffusion code PANTHER using homogenised constant libraries generated by WIMS. The study compared a number of core geometry configurations and fuel management schemes to suppress excess reactivity throughout fuel depletion. Three options for SMBWR core aspect ratio using the same power density are investigated with the aim to assess the effect on the neutronic and thermal-hydraulic performance of the SMBWR. It is found that the thin and tall core configuration (192 fuel assemblies and 3.60 m) showed the least favourable performance out of the three options as it has the largest core pressure drop and thus requires taller chimney to develop natural circulation.
OVERVIEW OF THE HINKLEY POINT C EPR PROJECT ‘J0’ AND BEYOND
Authors: Dyrda James , Morrison Richard
Published: February 2021

The twin unit EPR construction at Hinkley Point C is the first nuclear new build project to be undertaken in the UK in a generation. Moreover, it is now the country’s only new reactor project still in progress. The station’s two 1650 MWe plants [1] will complement the UK’s decarbonisedm energy mix with addition of approximately 7% of the current national requirement. In 2019, less than three years from the final investment decision, the ‘J0’ milestone was reached. This represents the end of the preparatory works, including all the common raft nuclear island concrete for Unit 1, enabling the upwards erection of the reactor containment and other support buildings. In addition to this, important milestones such as launch of the first tunnel boring machine, to tunnel the Unit 1 sea-water intake pipe have all been achieved. Over the coming years, the project will need to evolve in order to meet the increasing complexity and multi-organisation scope, which will present a challenging delivery environment. In order to support this, a UK-based Design Authority works on review, verification and acceptance of the HPC design including its UK specificities. This allows NNB, as licensee, to own its design and prepare the pre-commissioning safety case and operational strategies, defensible to the relevant UK regulators. As an example, this paper presents the roles and responsibilities in the Reactor Technology Team. This small, flexible team is involved in technical oversight of all aspects of the fuel & core design, through normal operations, anticipated and design-basis faults, severe accidents and radiological consequence estimation.
THE PRINCESS PROJECT: FROM DIFFERENTIAL TO INTEGRAL EXPERIMENTS
Authors: Duhamel Isabelle , Brovchenko Mariya , Clavel Jean-Baptiste , Duluc Matthieu , Ichou Raphaëlle , Leal Luiz , Leclaire Nicolas , Monange Wilfried
Published: February 2021

Following the shutdown of the CEA Valduc experimental facilities, where, for more than 50 years, IRSN used to perform experiments related to criticality safety, IRSN initiated a new project named PRINCESS (PRoject for IRSN Neutron physics and Criticality Experimental data Supporting Safety). The objective is to continue collecting experimental data necessary for the IRSN missions in nuclear safety. For this purpose, collaborations with various national and international laboratories have been established. The PRINCESS project covers various nuclear physics fields from nuclear data to criticality-safety and reactor physics providing information to both differential and integral data improvements.
MONTE CARLO-BASED DYNAMIC CALCULATIONS OF STATIONARY PERTURBATIONS
Authors: Demazière C , Tatidis A , Vinai P
Published: February 2021

Capitalizing on some earlier work, this paper presents a novel Monte Carlo-based approach that allows estimating the neutron noise induced by stationary perturbations of macroscopic cross-sections in the frequency domain. This method relies on the prior computation using Monte Carlo of modified Green’s functions associated to the real part of the dynamic macroscopic cross-sections, mimicking equivalent subcritical problems driven by external neutron sources. Once such modified Green’s functions are estimated, the neutron noise induced by any type of perturbations can be recovered, by solving a linear algebra problem accounting for the interdependence between the real and imaginary parts of the governing balance equations. The newly derived method was demonstrated on a large homogeneous test system and on a small heterogeneous test system to provide results comparable to a diffusion-based solver specifically developed for neutron noise applications. The new method requires the specification by the user of the real part of the Fourier transform of the macroscopic cross-sections. This is accomplished using ACE-formatted cross-section files defined by the user. Beyond this input data preparation, no change to the Monte Carlo source code is necessary. This represents the main advantage of the proposed method as compared to similar efforts requiring extensive modifications to the Monte Carlo source code.
NEUTRON NOISE-BASED ANOMALY CLASSIFICATION AND LOCALIZATION USING MACHINE LEARNING
Authors: Demazière C , Mylonakis A , Vinai P , Durrant A , De Sousa Ribeiro F , Wingate J , Leontidis G , Kollias S
Published: February 2021

A methodology is proposed in this paper allowing the classification of anomalies and subsequently their possible localization in nuclear reactor cores during operation. The method relies on the monitoring of the neutron noise recorded by in-core neutron detectors located at very few discrete locations throughout the core. In order to unfold from the detectors readings the necessary information, a 3-dimensional Convolutional Neural Network is used, with the training and validation of the network based on simulated data. In the reported work, the approach was also tested on simulated data. The simulations were carried out in the frequency domain using the CORE SIM+ diffusion-based two-group core simulator. The different scenarios correspond to the following cases: a generic “absorber of variable strength”, axially travelling perturbations at the velocity of the coolant flow (due to e.g. fluctuations of the coolant temperature at the inlet of the core), fuel assembly vibrations, control rod vibrations, and core barrel vibrations. In all those cases, various frequencies were considered and, when relevant, different locations of the perturbations and different vibration modes were taken into account. The machine learning approach was able to correctly identify the different scenarios with a maximum error of 0.11%. Moreover, the error in localizing anomalies had a mean squared error of 0.3072 in mesh size, corresponding to less than 4 cm. The proposed methodology was also demonstrated to be insensitive to parasitic noise and will be tested on actual plant data in the near future.
ANALYSIS OF THE FIRST COLIBRI FUEL RODS OSCILLATION CAMPAIGN IN THE CROCUS REACTOR FOR THE EUROPEAN PROJECT CORTEX
Authors: Lamirand V. , Rais A. , Pakari O. , Hursin M. , Laureau A. , Pohlus J. , Paquee U. , Pohl C. , Hübner S. , Lange C. , Frajtag P. , Godat D. , Perret G. , Fiorina C. , Pautz A.
Published: February 2021

The Horizon2020 European project CORTEX aims at developing an innovative core monitoring technique that allows detecting anomalies in nuclear reactors, such as excessive vibrations of core internals, flow blockage, or coolant inlet perturbations. The technique will be mainly based on using the fluctuations in neutron flux recorded by in-core and ex-core instrumentation, from which the anomalies will be differentiated depending on their type, location and characteristics. The project will result in a deepened understanding of the physical processes involved, allowing utilities to detect operational problems at a very early stage. In this framework, neutron noise computational methods and models are developed. In parallel, mechanical noise experimental campaigns are carried out in two zero-power reactors: AKR-2 and CROCUS. The aim is to produce high quality neutron noise-specific experimental data for the validation of the models. In CROCUS, the COLIBRI experimental program was developed to investigate experimentally the radiation noise induced by fuel rods vibrations. In this way, the 2018 first CORTEX campaign in CROCUS consisted in experiments with a perturbation induced by a fuel rods oscillator. Eighteen fuel rods located at the periphery of the core fuel lattice were oscillated between ±0.5 mm and ±2.0 mm around their central position at a frequency ranging from 0.1 Hz to 2 Hz. Signals from 11 neutron detectors which were set at positions in-core and ex-core in the water reflector, were recorded. The present article documents the results in noise level of the experimental campaign. Neutron noise levels are compared for several oscillation frequencies and amplitudes, and at the various detector locations concluding to the observation of a spatial dependency of the noise in amplitude.
Multi-stage appointment scheduling for outpatient chemotherapy unit: a case study
Authors: Bouras Asma , Masmoudi Malek , Saadani Nour El Houda , Bahroun Zied , Abdeljaouad Mohamed Amine
Published: February 2021

This paper deals with a multi stage hybrid flow-shop problem (HFSP) that arises in a privately Chemotherapy clinic. It aims to optimize the makespan of the daily chemotherapy activity. Each patient must respect the cyclic nature of chemotherapy treatment plans made by his referent on- cologist while taking into account the high variability in resource requirements (treatment time, nurse time, pharmacy time). The problem requires the assignment of chemotherapy patients to oncologists, pharmacists, chemotherapy beds or chairs and nurses over a 1-day period. We provided a Mixed Integer Program (MIP) to model this issue, which can be considered as a five-stage hybrid flow-shop scheduling problem with additional resources, dedicated machines, and no-wait constraints. Since this problem is known to be NP-hard, we provided a lower bound expression and developed an approximated solving algorithm: a tabu search inspired metaheuristic based on a constructive heuristic that can quickly reach satisfying results. To assess the empirical performance of the proposed approach, we conducted experi- ments on randomly generated instances based on real-world data of a Tunisian private clinic: Clinique Ennasr. Computational experiments show the efficiency of the proposed procedures: The mathematical model provided optimal solutions in reasonable computational time only for small instances (up to 10 patients). Meta-heuristic’s results demonstrate, also, that the proposed approach offers good results in terms of solution quality and computational times with an average relative gap to the MIP solution equal to 3.13% and to the lower bound equal to 5.37% for small instances (up to 15 patients). The same gap to the lower bound increases to 25% for medium and large size instances (20–50 patients).
Psychotherapeutic Treatment Options in Glaucoma Patients
Authors: Bertelmann Thomas , Strempel Ilse
Published: February 2021

Glaucoma poses the second largest cause of severe visual impairment and loss of vision worldwide. Despite the progress in both pharmaceutical and surgical treatments, the attempts to decrease intraocular pressure and prevent progression of glaucomatous optic neuropathy fail in many cases. Hence there is a high demand for additional complementary methods, which can reduce intraocular pressure and improve ocular blood flow as well as mental well-being. On the basis of literature research, the study results published so far on the effectiveness of psychotherapeutical methods in glaucoma therapy will be presented here. The methods of autogenic training, hypnosis and music therapy have already been demonstrated to have a positive effect on intraocular pressure, ocular blood flow and psychological well-being of patients affected by glaucoma. With these methods being not only effective but also cost-efficient, free of side effects and easily administered, they could gain
importance in terms of an adjuvant treatment option for patients with glaucoma. However, regular ophthalmological examinations still remain obligatory.
Globalization, economic development, and corruption: A cross-lagged contingency perspective
Authors: Aïssaoui Rachida , Fabian Frances
Published: February 2021

The 2020 health and economic crisis has exacerbated tensions and debates over whether globalization benefits economic development, as countries face both pressures to enhance economic opportunities through globalization and populist movements seeking protection from global forces. We first review perspectives that offer competing evidence about the role of globalization in regard to economic development and corruption. Drawing on resource dependence and institutional theory, we test the two contingencies of the country’s stage of economic development (low, lower-middle, upper-middle, high) and the globalization dimension (economic, social, political) at play to reconcile competing findings. Using a cross-lagged panel design, we show that these contingencies significantly explain when and what type of globalization can benefit a country’s economy and affect corruption. In doing so, the study provides a platform for future research, and identifies important patterns that can better guide policymaking. Among other results, we find low-income countries’ GDP and corruption benefit the most from the formal dimensions of globalization. With increased wealth, countries are more responsive to the legitimacy accrued with the informal dimensions of globalization, which we find comes at the expense of economic efficiency for high-income countries.
Diffusion of charge carriers in pentacene
Authors: Rani Varsha , Sharma Akanksha , Chaudhuri Sumita Ray , Ghosh Subhasis
Published: February 2021

The diffusion coefficient (D) of charge carriers in pentacene has been determined independently using current–voltage and capacitance–frequency characteristics of asymmetric metal/pentacene/metal structures. The values of D measured using these two methods are found to be in excellent agreement. D has been estimated using first principles calculations and compared with experimental values. The applicability of the Einstein relation has been examined in organic semiconductors.
Multiphysics simulation of plasma channel formation during micro-electrical discharge machining
Authors: Raza Sohaib , Nirala Chandrakant Kumar
Published: February 2021

A 2D axisymmetric plasma model for micro-electrical discharge machining (μEDM) is developed, and the discharge phenomenon is discussed in this paper. Variations in different plasma properties, such as density, temperature, and collisions of the electrons bombarding the anode and cathode electrodes, were simulated to comprehensively explain the discharge process. The said properties of the plasma channel will be extremely helpful in determining the heat flux available at the tool and workpiece of μEDM. The governing equations of electrostatics, drift-diffusion, and heavy species transport were coupled together and solved simultaneously for computing the properties of the plasma channel in water vapor. The simulation describes the movement of electrons and ions in the inter-electrode gap during the discharge initiation under the applied electric field. The anode spot responsible for the material removal was formed much earlier compared to the cathode spot formed at the tool. Both the temperature and the density of the electrons were observed to be higher near the workpiece, compared to the tool electrode. The temperature of the electrons and the current density of the plasma obtained during the simulation will be useful to determine the heat flux responsible for the material removal. The non-equilibrium nature of the plasma sheath is responsible for the steep changes in the collisional power loss and higher capacitive power deposition near the workpiece electrode.
Theoretical analysis of annular thermoelectric generators made of functionally graded materials
Authors: Niu Wei , Cao Xiaoshan , Hu Yifeng , Wang Fangfang , Shi Junping
Published: February 2021

In this study, the temperature field distribution and thermoelectric conversion efficiency of annular thermoelectric (TE) generators with functionally graded materials (FGMs) are investigated. The power series method is employed for solving a linear differential equation with variable coefficients to obtain temperature distribution, which has been proved to have good convergence and high precision. Numerical studies are carried out on FGM TE generators with linear and exponential variation, respectively. Material parameters that vary dependently and independently are discussed. If the figure of merit parameter Z keeps constant, then increases in the maximal conversion efficiency are not obvious. If material parameters vary independently, then the Seebeck coefficient increases from the cold temperature end to the hot temperature end and the maximal efficiency reaches 17.86%, which increases by 25.86% over 14.19% of homogeneous materials. For temperature-dependent TE materials, the problem could also be simplified to a functionally graded problem by presupposing an initial temperature field distribution. All these results give theoretical guidance for optimization of annular TE generators by using functionally graded structures.
A computational insight into the relationship between side chain IR line shapes and local environment in fibril-like structures
Authors: Pinto Sandra M. V. , Tasinato Nicola , Barone Vincenzo , Zanetti-Polzi Laura , Daidone Isabella
Published: February 2021

Infrared spectroscopy is a widely used technique to characterize protein structures and protein mediated processes. While the amide I band provides information on proteins’ secondary structure, amino acid side chains are used as infrared probes for the investigation of protein reactions and local properties. In this paper, we use a hybrid quantum mechanical/classical molecular dynamical approach based on the perturbed matrix method to compute the infrared band due to the C=O stretching mode of amide-containing side chains. We calculate, at first, the infrared band of zwitterionic glutamine in water and obtain results in very good agreement with the experimental data. Then, we compute the signal arising from glutamine side chains in a microcrystal of the yeast prion Sup35-derived peptide, GNNQQNY, with a fibrillar structure. The infrared bands obtained by selective isotopic labeling of the two glutamine residues, Q4 and Q5, of each peptide were experimentally used to investigate the local hydration in the fibrillar microcrystal. The experimental spectra of the two glutamine residues, which experience different hydration environments, feature different spectral signals that are well reproduced by the corresponding calculated spectra. In addition, the analysis of the simulated spectra clarifies the molecular origin of the experimentally observed spectroscopic differences that arise from the different local electric field experienced by the two glutamine residues, which is, in turn, determined by a different hydrogen bonding pattern.
The influences of high temperature on tribological properties of Cu-based friction materials with a friction phase of SiO2/SiC/Al2O3
Authors: Si Lina , Liu Cheng , Yan Hongjuan , Wang Yanjie , Yang Ye , Zhang Shuting , Zhang Yuyan
Published: February 2021

Cu-based friction materials have better tribological properties, high mechanical strength, and excellent thermal conductivity; therefore, they are widely used as friction materials in high-speed rail brake systems. However, with the remarkable increase in running speed of high-speed trains, a great amount of energy and high temperature are induced in the brake system; therefore, new challenges are generated for brake pad materials. In this paper, Cu-based friction materials were prepared with the addition of various hard particles as the friction phase, in order to obtain a good high temperature tribological performance. The experimental results showed that the Cu-based friction materials with addition of hard particles have moderate mechanical strengths and proper tribological properties. The temperature effects on the tribological behaviors of friction materials were investigated. The coefficient of friction (COF) for friction materials with addition of hard particles increased with temperature, resulting from the integrated effects of friction film, graphite, and material softening. The variation trends of the wear loss and wear rate with temperatures are similar to those of COFs, which are significantly reduced with the addition of hard particles.
Unconventional anomalous Hall effect in magnetic topological insulator MnBi4Te7 device
Authors: Jia Bin , Zhang Shuai , Ying Zhe , Xie Hangkai , Chen Bo , Naveed Muhammad , Fei Fucong , Zhang Minhao , Pan Danfeng , Song Fengqi
Published: February 2021

The natural intrinsic magnetic topological insulator MnBi2Te4(Bi2Te3)n is a platform for studying intriguing transport phenomena and provides an essential chance for the fundamental understanding of the combination of magnetism and topology. Here, we fabricated MnBi4Te7 thin film devices and carried out the transport measurement. It shows the unconventional anomalous Hall effect in the devices with the hysteresis anomaly and hump-like behavior. The gate-controlled sign reversal is also an interesting feature. With the modulation of gate voltage and temperature, we suggest that the unconventional anomalous Hall effect is due to the coexistence of the antiferromagnetic component and ferromagnetic component, which may result from the inhomogeneous spatial variation. Our work provides a resourceful insight into the study of the material family.