bluuebunny/crossref_metadata_2025 huggingface dataset:
Selenium-catalyzed intramolecular atom- and redox-economical transformation ofo-nitrotoluenes into anthranilic acids
Authors: Li Yiming , Wang Yuhong , Yang Tilong , Lin Zhenyang , Jiang Xuefeng
Published: February 2021

Anthranilic acids (AAs) are significant basic chemicals used in pharmaceuticals, agrochemicals, dyes, fragrances,etc.
Benzimidazole-1,2,3-triazole hybrid molecules: synthesis and study of their interaction with G-quadruplex DNA
Authors: Singu Padma S. , Chilakamarthi Ushasri , Mahadik Namita S. , Keerti Bhamidipati , Valipenta Narasimhulu , Mokale Santosh N. , Nagesh Narayana , Kumbhare Ravindra M.
Published: February 2021

The benzimidazole-1,2,3-triazole hybrid 4f selectively interacted with G-quadruplex DNA over duplex DNA, inhibits cell cycle at the G2/M phase, inducing apoptosis, and may be a G-quadruplex DNA groove binder with anticancer activity.
Tuning protein adsorption on graphene surfaces via laser-induced oxidation
Authors: Sitsanidis Efstratios D. , Schirmer Johanna , Lampinen Aku , Mentel Kamila K. , Hiltunen Vesa-Matti , Ruokolainen Visa , Johansson Andreas , Myllyperkiö Pasi , Nissinen Maija , Pettersson Mika
Published: February 2021

Controlled guidance of proteins by light over graphene surfaces.
P-doped SnFe nanocubes decorated with PdFe alloy nanoparticles for ethanol fuel cells
Authors: Konwar Dimpul , Basumatary Padmini , Lee UnHo , Yoon Young Soo
Published: February 2021

In this study, a three-dimensional SnFeP nanocube was prepared, and then PdFe alloy was decorated over it. A single cell with PdFe/SnFeP@rGO anode catalysts exhibited promising power densities in ethanol and hydrogen fuels at low temperature.
An unprecedented π-electronic circuit involving an odd number of carbon atoms in a grossly warped non-planar nanographene
Authors: Escayola Sílvia , Poater Albert , Muñoz-Castro Alvaro , Solà Miquel
Published: February 2021

The formation of π-aromatic circuits along a grossly warped nanographene, C80H30, containing five- and seven-membered rings inserted into a six-membered mesh, reveals global π-circuits at the edge of the backbone.
Evolution of Pt truncated octahedra to nanodendrites during the synthesis in methanol–water solution
Authors: Zhou Di , Han Shaobo , Li Yong , Ta Na , Zhou Yan , Li Mingrun , Miao Shu , Shen Wenjie
Published: February 2021

The size and shape of the Pt particles could be tuned from truncated octahedra (3.5 nm) to nanodendrites (13.5 nm) by simply varying the CH3OH/H2O ratio in the synthetic solution.
First example of solid-state luminescent borasiloxane-based chiral helices assembled through N–B bonds
Authors: Podda Enrico , Coles Simon J. , Horton Peter N. , Lickiss Paul D. , Bull Okpara S. , Orton James B. , Pintus Anna , Pugh David , Carla Aragoni M. , Davies Robert P.
Published: February 2021

The reaction between differently substituted borasiloxanes and 2,5-bis(3-pyridylethynyl)thiophene provided the first example of luminescent borasiloxane-based chiral helices held together by N–B bonds.
One-carbon metabolism in cancer cells: a critical review based on a core model of central metabolism
Authors: Mazat Jean-Pierre
Published: February 2021

One-carbon metabolism (1C-metabolism), also called folate metabolism because the carbon group is attached to folate-derived tetrahydrofolate, is crucial in metabolism. It is at the heart of several essential syntheses, particularly those of purine and thymidylate. After a short reminder of the organization of 1C-metabolism, I list its salient features as reported in the literature. Then, using flux balance analysis, a core model of central metabolism and the flux constraints for an ‘average cancer cell metabolism’, I explore the fundamentals underlying 1C-metabolism and its relationships with the rest of metabolism. Some unreported properties of 1C-metabolism emerge, such as its potential roles in mitochondrial NADH exchange with cytosolic NADPH, participation in NADH recycling, and optimization of cell proliferation.
APOGEE spectroscopic evidence for chemical anomalies in dwarf galaxies: The case of M 54 and Sagittarius
Authors: Fernández-Trincado José G. , Beers Timothy C. , Minniti Dante , Moni Bidin Christian , Barbuy Beatriz , Villanova Sandro , Geisler Doug , Lane Richard R. , Roman-Lopes Alexandre , Bizyaev Dmitry
Published: February 2021

We present evidence for globular cluster stellar debris in a dwarf galaxy system (Sagittarius; Sgr) based on an analysis of high-resolution H-band spectra from the Apache Point Observatory Galactic Evolution Experiment (APOGEE) survey. We add [N/Fe], [Ti/Fe], and [Ni/Fe] abundance ratios to the existing sample of potential members of M 54; this is the first time that [N/Fe] abundances are derived for a large number of stars in M 54. Our study reveals the existence of a significant population of nitrogen- (with a large spread, ≳1 dex) and aluminum-enriched stars with moderate Mg depletions in the core of the M 54+Sgr system, which share the light element anomalies characteristic of second-generation globular cluster stars (GC), thus tracing the typical phenomenon of multiple stellar populations seen in other Galactic GCs at similar metallicity, confirming earlier results based on the Na-O anti-correlation. We further show that most of the stars in M 54 exhibit different chemical patterns evidently not present in Sgr field stars. Furthermore, we report the serendipitous discovery of a nitrogen-enhanced extra-tidal star that has GC second-generation-like chemical patterns for which both chemical and kinematic evidence are commensurate with the hypothesis that the star has been ejected from M 54. Our findings support the existence of chemical anomalies associated with likely tidally shredded GCs that are in dwarf galaxies in the Local Group. These findings motivate future searches for such bona fide stars along other known Milky Way streams.
NEUTRONIC BENCHMARK ON HOLOS-QUAD MICRO-REACTOR CONCEPT
Authors: Stauff N. E. , Lee C. H. , Shriwise P. , Wells A. , Filippone C.
Published: February 2021

The Holos-Quad micro-reactor concept is proposed by HolosGen LLC for civilian applications to generate 22 MWt with a lifetime of approximately 8 effective full power years (EFPYs). The design is based on a very innovative high-temperature gas-cooled reactor concept using four Subcritical Power Modules (SPMs) that fit into one commercial 40-foot transport ISO container. Neutronics benchmarks were developed based on a preliminary version of the Holos-Quad design to confirm feasibility of the neutronics design of this non-traditional high temperature gas-cooled (HTGR)-type micro-reactor concept. Calculations were performed using Monte Carlo codes (SERPENT and OpenMC) as well as the PROTEUS high-fidelity deterministic code for two exercises based on a unit cell model and a full core model. The results obtained showed good agreement with all of the evaluated parameters for the unit cell problem and the full core problem. SERPENT and OpenMC display consistently good agreement in eigenvalue within 150 pcm for the unit cell benchmark and less than 270 pcm for the full core benchmark. PROTEUS eigenvalues showed relatively larger differences but still reasonable agreement with the SERPENT solutions. For the depletion benchmark, the observed eigenvalue differences between SERPENT and OpenMC were within 300 pcm throughout the depletion up to 100 GWd/MT, ensuring that the two codes used equivalent input parameters such as the recoverable heat values for fission reactions. This benchmark exercise confirms the neutronic feasibility and core performance of the preliminary Holos-Quad design.
A RESONANCE CALCULATION METHOD USING ENERGY EXPANSION BASED ON A REDUCED ORDER MODEL: USE OF ULTRA-FINE GROUP SPECTRUM CALCULATION AND APPLICATION TO HETEROGENEOUS GEOMETRY
Authors: Kondo Ryoichi , Endo Tomohiro , Yamamoto Akio , Takeda Satoshi , Koike Hiroki , Yamaji Kazuya , Ieyama Koichi , Sato Daisuke
Published: February 2021

A Resonance calculation using energy Spectral Expansion (RSE) method has been recently proposed in order to efficiently treat complicated heterogeneous geometry and resonance interference effect. In the RSE method, ultra-fine group spectra are generated from ultra-fine group calculations in homogeneous geometry, and the spectra are expanded by the orthogonal basis on energy based on the singular value decomposition. Then the transport calculation for expansion coefficients is numerically performed, and the ultra-fine group spectra in the target heterogeneous regions are reconstructed by the expansion coefficients and the orthogonal basis. In this study, the RSE method is applied to multi-cell geometries including UO2, MOX and water cells, in which the resonance interference effect between UO2 and MOX fuel cells appears. The validity of the RSE method is confirmed through comparison with the reference effective multi-group cross sections obtained from the direct ultra-fine group calculation in the target heterogeneous geometry.
ANALYSES OF INTEGRAL AND MOX CRITICAL EXPERIMENTS TO QUALIFY PARAGON2 PREDICTIONS*
Authors: Ouisloumen Mohamed
Published: February 2021

This paper presents the qualification of the newly developed Westinghouse lattice physics code PARAGON2. PARAGON2 uses high energy resolution in the solution of the transport equation. The objective of this paper is to demonstrate that PARAGON2 accurately predicts the integral and critical experiments. The integral experiments are used to assess PARAGON2 predictions of fine neutronics parameters such as: resonance integrals and radial profiles of reactions rates, isotopics, and burnup for depleted pellets. The integral experiments considered are: the Hellstrand’s, TRX, and the PIE experiments. For critical experiments, this paper will focus only on VENUS-2 MOX critical experiment.

The results obtained for the integral experiments clearly show the good predictions of PARAGON2 with the resonance scattering model which are close to measurement. The PARAGON2 predicted capture reaction rates, temperature coefficients, burnup and isotopic profiles match the measured values both in shape and magnitude. VENUS-2 reactivity prediction is in excellent agreement with the critical measurement value. Also, the standard deviations of measured minus predicted pin-wise fission reaction rates are very good (i.e. ≤ 2%) for both individual assemblies and the whole core.
NEUTRON NOISE PATTERNS FROM COUPLED FUEL-ASSEMBLY VIBRATIONS
Authors: Viebach M. , Lange C. , Seidl M. , Bilodid Y. , Hurtado A.
Published: February 2021

The neutron flux fluctuation magnitude of KWU-built PWRs shows a hitherto unexplained correlation with the types of loaded fuel assemblies. Also, certain measured long-range neutron flux fluctuation patterns in neighboring core quadrants still lack a closed understanding of their origin. The explanation of these phenomena has recently revived a new interest in neutron noise research.

The contribution at hand investigates the idea that a synchronized coolant-driven vibration of major parts of the fuel-assembly ensemble leads to these phenomena. Starting with an assumed mode of such collective vibration, the resulting effects on the time-dependent neutron-flux distribution are analyzed via a DYN3D simulation. A three-dimensional representation of the time-dependent bow of all fuel assemblies is taken into account as a nodal DYN3D feedback parameter by time-dependent variations of the fuel-assembly pitch. The impact of its variation on the cross sections is quantiﬁed using a cross-section library that is generated from the output of corresponding CASMO5 calculations.

The DYN3D simulation qualitatively reproduces the measured neutron-flux fluctuation patterns. The magnitude of the fluctuations and its radial dependence are comparable to the measured details. The results imply that collective fuel-assembly vibrations are a promising candidate for being the key to understand long-known fluctuation patterns in KWU built PWRs. Further research should elaborate on possible excitation mechanisms of the assumed vibration modes.
VALIDATION OF THE WIMS/PANTHER EMBEDDED SUPERCELL METHOD
Authors: Bryce Paul , Hosking Glynn , Knight Martin , Lindley Ben , Powney David , Schneidesch Christophe , Slosse Nicolas , Taylor Tom
Published: February 2021

The WIMS/PANTHER Embedded Supercell Method (ESM) provides a signiﬁcant improvement in prediction accuracy for radial power distributions for PWR reactors compared to the standard “two-step” approach, without the need for a signiﬁcant increase in computational resource. Recent papers at PHYSOR conferences have outlined the details of the method and demonstrated its operation, and the accuracy improvements possible, by means of benchmarking calculations.

This paper applies the method to a 4-loop PWR in the U.K, and three PWRs (3-loop and 2-loop) in Belgium. Comparisons are made against measured data from the start-of-cycle physics testing performed for each cycle, and power-shape measurements collected during the cycle using a conventional “two-step” nodal reactor solution, and with the ESM. All results will be presented with the JEF2.2 nuclear data library, for ease of comparison between the methods and previously reported results, although the effects of more modern evaluations will be commented upon.

The benchmark calculations referred to above studied a challenging MOX/UO2 benchmark core akin to an SMR. The four reactors studied here include conventional UO2 only core designs and cycles with UO2/MOX mixed cores. A variety of boron-and gadolinium-based burnable absorbers are also present. The data is used to show that the method both operates successfully for real reactor problems, and delivers improvements in the prediction accuracy of measured parameters.
APPLICATION OF STUDSVIK’S CMS5 CODE SYSTEM TO ACCIDENT TOLERANT FUEL CORE DESIGN AND ANALYSIS
Authors: Grandi Gerardo , Ferrer Rodolfo , Bahadir Tamer
Published: February 2021

The possible deployment of Accident Tolerant Fuels (ATF) for currently-operating Light Water Reactors (LWR) has prompted interest in the use of Studsvik’s CMS5 code system to support the analysis of such advanced ATF core designs. Various ATF concepts have been proposed; for example, uranium silicide (U3Si2) fuel, together with iron-based (FeCrAl) cladding. The purpose of this work is to showcase the application of the CMS5 code system, which includes the CASMO5 advanced lattice physics code and the SIMULATE5 three-dimensional nodal simulator, to the analysis of a U3Si2/FeCrAl ATF concept. Given that the CMS5 code system was designed from inception to enable the analysis of advanced core designs, only minor changes to the CASMO5 lattice physics code and SIMULATE5 core simulator are necessary. The current CASMO5 586 energy-group nuclear data library provides all the necessary data to support the generation of homogenized data for downstream use by SIMULATE5 for ATF. The SIMULATE5 nodal code, which features a simplified fuel pin model, requires updating various thermophysical properties corresponding to the U3Si2/SiC ATF fuel and the gaseous conductance models. An equilibrium core for the Integral Inherently Safe (I2S) LWR design developed by the Georgia Institute of Technology was selected. The results of the CMS5 simulation were compared with those in the literature and were found to be in good agreement, giving us confidence that the CMS5 package can be used in the modeling of LWR systems with ATF technology.
ONE- AND TWO-NODE HYBRID LEAKAGE-CORRECTED 3-D CMFD ALGORITHM FOR FAST AND ACCURATE PIN-WISE PWR ANALYSIS
Authors: Kim Jaeha , Yu Hwanyeal , Kim Yonghee
Published: February 2021

The HCMFD (Hybrid Coarse-Mesh Finite Difference) with GPS (GET Plus SPH) method and its 3-D application are investigated in this paper. In the HCMFD algorithm, a parallel computing for a pin-by-pin core calculation can be performed very effectively with a non-linear local-global iterative scheme. For an effective parallel computing, the one-node CMFD is used to solve a global eigenvalue problem. The conventional two-node CMFD is used for solving local fixed source problems with incoming current boundary conditions. The GPS method is a leakage correction method to correct the pin-wise XSs of the conventional GET-based two-step procedure. In the GPS method, the XS-dependent SPH factors are parameterized as a function of the pin-wise albedo information, current-to-flux ratio (CFR). With updated XS-wise SPH factors, the pin-wise XSs are corrected in order to improve the accuracy of the conventional two-step core analysis. The GPS method is implemented to an in-house pin-by-pin diffusion solver with the HCMFD algorithm to keep their strong points together. In this paper, several 3-D variant cores of KAIST-1A benchmark problem were chosen to demonstrate the combination of HCMFD and GPS (HCMFD-GPS) method.
A STUDY OF LEAKAGE-CORRECTED TWO-STEP METHOD BASED ON THE NODAL EQUIVALENCE THEORY FOR FAST REACTOR ANALYSIS
Authors: Jang Seongdong , Kim Yonghee
Published: February 2021

The conventional two-step method based on the generalized equivalence theory (GET) cannot be directly applied to the fast reactor analysis since the assumption of the space-energy separability is not very valid due to a relatively long neutron mean free path. This study aims to develop a leakage-corrected two-step method for the fast reactor analysis with the aid of the albedo-corrected parameterized equivalence constants (APEC) method. The critical idea of the APEC method is to correct the homogenized group constants (HGCs) including discontinuity factors (DFs) during the nodal calculation through predetermined APEC functions. The APEC functions are functionalized in terms of the normalized leakage parameters such as a current-to-flux (CFR) ratio so that they can correct the cross-sections (XSs) and discontinuity factors by reflecting the in-situ neutron leakage information of the nodal analysis. The feasibility of the APEC-corrected two-step method was investigated by solving 5-group diffusion equations for a two-dimensional sodium-cooled fast reactor with a 6-triangle finite difference method. The 5-group HGCs for fuel assemblies were determined by using a continuous-energy Monte Carl code, and the conventional assembly discontinuity factors are also introduced for each hexagonal fuel assembly. First of all, it was demonstrated that the simple FDM scheme could reproduce the reference nodal quantities with the GET. And the APEC functions are formulated using the reference solutions to evaluate the feasibility of simple APEC functional for both XSs and DFs. Then, a smaller color-set problem was defined to determine practical APEC functions for the original benchmark, and various numerical evaluations are performed in terms of the k-eff value and nodal power distribution.
USE OF THE SIMPLIFIED PN EQUATIONS AND TRANSPORT CORRECTIONS IN THE WIMS/PANTHER EMBEDDED SUPERCELL METHOD
Authors: Taylor Tom , Knight Martin , Bryce Paul
Published: February 2021

The WIMS/PANTHER Embedded Supercell Method (ESM) provides a significant improvement in prediction accuracy in radial power distributions for pressurised water reactors compared to the standard “two-step” approach, without the need for a significant increase in computational resource.

A companion paper presents validation of the ESM as previously presented, using PANTHER pin-by-pin diffusion to correct interface errors arising from the standard two-step approach. However, in principle any reference method can be used to solve the embedded supercells and correct the basic solution – this is a significant advantage of the ESM.

A paper presented at PHYSOR 2016 demonstrated that use of diffusion theory introduces significant error relative to transport theory only in the high energy range, due to the discontinuous fission source when modelling the interface between two types of pincell.

This paper investigates further improvement of the PANTHER solution through use of simplified PN in the fast energy groups. This solution method can be implemented without the need for any significant change to the calculation route and further improves agreement with a transport reference for a small computational cost.

The remaining error in the solution is examined by derivation of an effective diffusion coefficient from a heterogeneous transport reference. This suggests a simple characterisation of the discrepancy, which can be easily corrected.

Results are presented in comparison to WIMS for supercells and the KAIST benchmark. For the KAIST small core benchmark assembly power errors relative to a WIMS fine group transport reference are shown to be less than 0.5 %. Results are also presented in comparison to Monte Carlo for the Watts Bar benchmark.
MULTILEVEL CMFD ACCELERATION METHODS FOR WHOLE CORE REACTOR SIMULATION IN VERA
Authors: Cornejo Luke , Collins Benjamin , Stimpson Shane
Published: February 2021

Ongoing efforts are being made to improve the performance of MPACT as the deterministic neutron transport solver in the Virtual Environment for Reactor Analysis (VERA). As other parts of the code have been improved, the coarse mesh finite difference method (CMFD) has come to take up a significant portion of the runtime. Multilevel-in-energy CMFD and multilevel-in-space CMFD solvers have been used to improve CMFD solver performance. A new multilevel-in-space-and-energy CMFD solver is being introduced that combines components of these two methods. W-Cycles and partial W-Cycles are being investigated to further improve the efficiency of the multilevel-in-energy CMFD solver. The performance of these methods is demonstrated on full core reactor physics problems of interest to VERA.
COMPARISON OF CHEBYSHEV AND ANDERSON ACCELERATIONS FOR THE NEUTRON TRANSPORT EQUATION
Authors: Calloo Ansar , Le Tellier Romain , Couyras David
Published: February 2021

This work focuses on the k-eigenvalue problem of the neutron transport equation. The variables of interest are the largest eigenvalue (keff) and the corresponding eigenmode is called the fundamental mode. Mathematically, this problem is usually solved using the power iteration method. However, the convergence of this algorithm can be very slow, especially if the dominance ratio is high as is the case in some reactor physics applications. Thus, the power iteration method has to be accelerated in some ways to improve its convergence. One such acceleration is the Chebyshev acceleration method which has been widely applied to legacy codes. In recent years, nonlinear methods have been applied to solve the k-eigenvalue problem. Nevertheless, they are often compared to the unaccelerated power iteration. Hence, the goal of this paper is to apply the Anderson acceleration to the power iteration, and compare its performance to the Chebyshev acceleration.
COMPARING THE DISCONTINUOUS GALERKIN AND HIGH-ORDER DIAMOND DIFFERENCING METHODS FOR THE TRANSPORT EQUATION ON A LOZENGE-BASED HEXAGONAL GEOMETRY
Authors: Calloo Atyab A. , Hébert Alain
Published: February 2021

This paper presents an implementation and a comparison of two spatial discretisation schemes over a hexagonal geometry for the two-dimensional discrete ordinates transport equation. The methods are a high-order Discontinuous Galerkin (DG) finite element scheme and a high-order Diamond Differencing (DD) scheme. The DG method has been, and is being, studied on the hexagonal geometry, also called a honeycomb mesh – but not the DD method. In this research effort, it was chosen to divide the hexagons into (at least) three lozenges. An affine transformation is then applied onto said lozenges to cast them into the reference quadrilaterals usually studied in finite elements. In practice, this effectively means that the equations used in Cartesian geometry have their terms and operators altered using the Jacobian matrix of the transformation. This was implemented in the discrete ordinates solver of the code DRAGON5. Two 2D benchmark problems were then used for the verification and validation, including one based on the Monju 3D reactor benchmark. It was found that the diamond-differencing scheme seemed better. It converged much faster towards the solution at comparable mesh refinements for first-order expansion of the flux. Even if this difference was not present for second-order, DG was slower, about two to four times slower.
THE TWO-STEP APPROACH FOR WHOLE-CORE RESONANCE SELF-SHIELDING CALCULATION
Authors: Qin Shuai , Zhang Qian , Liang Liang , He Qingming , Wu Hongchun
Published: February 2021

A two-step approach is proposed to accomplish high-fidelity whole-core resonance self-shielding calculation. Direct slowing-down equation solving based on the pin-cell scale is performed as the first step to simulate different operating conditions of the reactor. Resonance database is fitted using the results from the pin-cell calculation. Several techniques are used in the generation of the resonance database to estimate multiple types of resonance effects. The second step is the calculation of practical whole-core problem using the resonance database obtained from the first step. The transport solver is embedded both at the first step and the second step to establish the equivalence relationship between the fuel rod in the practical problem and the pin-cell at the first step. The numerical results show that the new approach have capability to perform high-fidelity resonance calculations for practical problem.
VALIDATION OF WIMS11 FOR SMALL MODULAR REACTOR ANALYSIS
Authors: Smith Peter , Hosking Glynn , Tollit Brendan , Lindley Ben , Cox Andrew , Perry Ray , Ware Tim , Mason Robert , Stefanowska Magda
Published: February 2021

The WIMS (Winfrith Improved Multigroup Scheme) reactor physics code is actively being developed for whole core modelling of a range of Small Modular Reactor types including the Pressurized Water Reactor (PWR), High Temperature Reactor (HTR), and Liquid Metal Cooled Fast reactor (LMFR). These developments include the capability for whole core multiphysics modelling with neutronics and thermal hydraulic feedback, as well as methods to determine the power deposition from neutron and gamma heating. Flux solutions are obtained using a wide variety of deterministic methods including diffusion theory, SP3, and full transport with the method of characteristics and Sn discrete ordinates methods, as well as multi-group Monte Carlo methods. The SP3 method allows both steady state and time dependent transient solutions by solving the time dependent SP3 equations. A wide variety of nuclear data libraries are available with WIMS including data from the JEF3.3, ENDF/B-VII.0 and CENDL3.1 nuclear data evaluations.

This paper presents validation of the latest version of the WIMS code, WIMS11, for PWR and HTR systems. Comparisons are made against physics data obtained from the OECD/NEA PWR Watts Bar multi-physics benchmark and the IAEA HTR-10 benchmark, as well as neutron and gamma heating experiments that took place on the NESSUS reactor at Winfrith in the United Kingdom. In each case, validation of WIMS has been obtained by comparison either against measured data, or results provided by other benchmark participants that have been obtained with alternative deterministic or Monte Carlo methods.
SERPENT/SUBCHANFLOW COUPLED CALCULATIONS FOR A VVER CORE AT HOT FULL POWER
Authors: Ferraro Diego , García Manuel , Imke Uwe , Valtavirta Ville , Tuominen Riku , Bilodid Yuri , Leppänen Jaakko , Sanchez-Espinoza Víctor
Published: February 2021

An increasing interest on the development of highly accurate methodologies in reactor physics is nowadays observed, mainly stimulated by the availability of vast computational resources. As a result, an on-going development of a wide range of coupled calculation tools is observed within diverse projects worldwide. Under this framework, the McSAFE European Union project is a coordinated effort aimed to develop multiphysics tools based on Monte Carlo neutron transport and subchannel thermal-hydraulics codes. These tools are aimed to be suitable for high-fidelity calculations both for PWR and VVER reactors, with the final goal of performing pin-by-pin coupled calculations at full core scope including burnup. Several intermediate steps are to be analyzed in-depth before jumping into this final goal in order to provide insights and to identify resources requirements. As part of this process, this work presents the results for a pin-by-pin coupling calculation using the Serpent 2 code (developed by VTT, Finland) and the subchannel code SUBCHANFLOW (SCF, developed by KIT, Germany) for a full-core VVER model. For such purpose, a recently refurbished master-slave coupling scheme is used within a High Performance Computing architecture. A full-core benchmark for a VVER-1000 that provides experimental data is considered, where the first burnup step (i.e. fresh core at hot-full rated power state) is calculated. For such purpose a detailed (i.e. pin-by-pin) coupled Serpent-SCF model is developed, including a simplified equilibrium xenon distribution (i.e. by fuel assembly). Comparisons with main global reported results are presented and briefly discussed, together with a raw estimation of resources requirements and a brief demonstration of the inherent capabilities of the proposed approach. The results presented here provide valuable insights and pave the way to tackle the final goals of the on-going high-fidelity project.
TOWARDS THE VALIDATION OF NOISE EXPERIMENTS IN THE CROCUS REACTOR USING THE TRIPOLI-4 MONTE CARLO CODE IN ANALOG MODE
Authors: Pakari Oskari , Mancusi Davide , Petit Odile , Zoia Andrea , Lamirand Vincent , Pautz Andreas
Published: February 2021

Intrinsic neutron noise experiments offer a non-invasive manner to measure the prompt decay constant or reactivity of fissile systems. Using the fluctuations in the density of fission chains, one can infer the kinetics parameters via correlation analysis such as the Rossi-alpha method. The models allowing for the interpretation of these measurements typically rely on the assumption of the system behaving according to point kinetics. When dealing with systems where point kinetics fail to predict the true time correlation – such as heterogeneous or large cores – the direct simulation of fission chains using Monte Carlo methods appears as the only reliable candidate to provide reference predictions for the correlation functions. Monte Carlo methods using explicit fission model libraries are thus being examined as tools for prediction in noise analysis. In this work we illustrate the developments and simulation results of the analog transport capabilities of the TRIPOLI-4 Monte Carlo code coupled with the LLNL fission library FREYA, as applied to a set of neutron noise experiments carried out in the CROCUS zero-power reactor with emphasis on the identification of spatial effects. To validate the general capability of the code to predict noise correlations, we examine time distributions of the whole core fission and explicit detection reactions. We present the methodology to achieve a good agreement between experiments and simulations. We reproduced experimental results for relative α, within typical biases, and conclude on the general feasibility of the analog method. We further explore a decoupled core model and analyze it using the noise method. The results indicate an effective method to treat decoupled systems.
ANALYSIS OF TIME-EIGENVALUE AND EIGENFUNCTIONS IN THE CROCUS BENCHMARK
Authors: Nauchi Yasushi , Jinaphanh Alexis , Zoia Andrea
Published: February 2021

Time-dependent neutron transport in non-critical state can be expressed by the natural mode equation. In order to estimate the dominant eigenvalue and eigenfunction of the natural mode, CEA had extended the α-k method and developed the generalized iterated fission probability method (G-IFP) in the TRIPOLI-4® code. CRIEPI has chosen to compute those quantities by a time-dependent neutron transport calculation, and has thus developed a time-dependent neutron transport technique based on k-power iteration (TDPI) in MCNP-5. In this work, we compare the two approaches by computing the dominant eigenvalue and the direct and adjoint eigenfunctions for the CROCUS benchmark. The model has previously been qualified for keffs and kinetic parameters by TRIPOLI-4 and MCNP-5. The eigenvalues of the natural mode equations by α-k and TDPI are in good agreement with each other, and closely follow those predicted by the inhour equation. Neutron spectra and spatial distributions (flux and fission neutron emission) obtained by the two methods are also in good agreement. Similar results are also obtained for the adjoint fundamental eigenfunctions. These findings substantiate the coherence of both calculation strategies for natural mode.
APPLICATION OF RESPONSE MATRIX METHOD TO TRANSIENT SIMULATIONS OF NUCLEAR SYSTEMS
Authors: Mickus Ignas , Roberts Jeremy A. , Dufek Jan
Published: February 2021

Until recently, reactor transient problems were exclusively solved by approximate deterministic methods. The increase in available computing power made it feasible to approach the transient analyses with time-dependent Monte Carlo methods. These methods offer the first-principle solution to the space-time evolution of reactor power by explicitly tracking prompt neutrons, precursors of delayed neutrons and delayed neutrons in time and space. Nevertheless, a very significant computing cost is associated with such methods. The general benefits of the Monte Carlo approach may be retained at a reduced computing cost by applying a hybrid stochastic-deterministic computing scheme. Among such schemes are those based on the fission matrix and the response matrix formalisms. These schemes aim at estimating a variant of the Greens function during a Monte Carlo transport calculation, which is later used to formulate a deterministic approach to solving a space-time dependent problem. In this contribution, we provide an overview of the time-dependent response matrix method, which describes a system by a set of response functions. We have recently suggested an approach where the functions are determined during a Monte Carlo criticality calculation and are then used to deterministically solve the space-time behaviour of the system. Here, we compare the time-dependent response matrix solution with the transient fission matrix and the time-dependent Monte Carlo solutions for a control rod movement problem in a mini-core reactor geometry. The response matrix formalism results in a set of loosely connected equations which offers favourable scaling properties compared to the methods based on the fission matrix formalism.
INLINE THERMAL AND XENON FEEDBACK ITERATIONS IN MONTE CARLO REACTOR CALCULATIONS
Authors: Gill Daniel F.
Published: February 2021

In this work, we describe a method for converging nonlinear feedback during the convergence of the neutron fission source in a Monte Carlo reactor simulation. This approach involves updating feedback physics during discard batches in the Monte Carlo simulation rather than fully (or partially) converging the neutronics prior to the nonlinear update. This approach is demonstrated for a single PWR pin with thermal feedback and with both thermal and xenon feedback. Converging these feedbacks inline with the fission source is shown to have the benefit of reducing numerical instability by effectively underrelaxing the tallied quantities in the Monte Carlo simulation and improving computational performance by converging feedback within (or near to) the number of discard batches required to converge the fission source even without any feedback.
NUMA-AWARE DATA MANAGEMENT FOR NEUTRON CROSS SECTION DATA IN CONTINUOUS ENERGY MONTE CARLO NEUTRON TRANSPORT SIMULATION
Authors: Denoyelle Nicolas , Tramm John , Yoshii Kazutomo , Perarnau Swann , Beckman Pete
Published: February 2021

The calculation of macroscopic neutron cross-sections is a fundamental part of the continuous-energy Monte Carlo (MC) neutron transport algorithm. MC simulations of full nuclear reactor cores are computationally expensive, making high-accuracy simulations impractical for most routine reactor analysis tasks because of their long time to solution. Thus, preparation of MC simulation algorithms for next generation supercomputers is extremely important as improvements in computational performance and efficiency will directly translate into improvements in achievable simulation accuracy. Due to the stochastic nature of the MC algorithm, cross-section data tables are accessed in a highly randomized manner, resulting in frequent cache misses and latency-bound memory accesses. Furthermore, contemporary and next generation non-uniform memory access (NUMA) computer architectures, featuring very high latencies and less cache space per core, will exacerbate this behaviour. The absence of a topology-aware allocation strategy in existing high-performance computing (HPC) programming models is a major source of performance problems in NUMA systems. Thus, to improve performance of the MC simulation algorithm, we propose a topology-aware data allocation strategies that allow full control over the location of data structures within a memory hierarchy. A new memory management library, known as AML, has recently been created to facilitate this mapping. To evaluate the usefulness of AML in the context of MC reactor simulations, we have converted two existing MC transport cross-section lookup “proxy-applications” (XSBench and RSBench) to utilize the AML allocation library. In this study, we use these proxy-applications to test several continuous-energy cross-section data lookup strategies (the nuclide grid, unionized grid, logarithmic hash grid, and multipole methods) with a number of AML allocation schemes on a variety of node architectures. We find that the AML library speeds up cross-section lookup performance up to 2x on current generation hardware (e.g., a dual-socket Skylake-based NUMA system) as compared with naive allocation. These exciting results also show a path forward for efficient performance on next-generation exascale supercomputer designs that feature even more complex NUMA memory hierarchies.
AN ASSESSMENT OF THE SPATIAL VARIATION OF ISOTOPIC RATIOS IN A CANDU REACTOR FOR NUCLEAR TREATY MONITORING
Authors: Burkhardt Aaron W. , Bevins James E. , Baxter Stephen H.
Published: February 2021

A 3-D Quarter-Core CANDU-6 is modeled using Serpent 2 for nuclear treaty monitoring. The spatial variation of flux spectra and isotopic concentrations is analyzed to determine the potential isotopic distribution of key radionuclides from standard reactor operations relevant to non-proliferation. The initial results of the model show a 46% difference in overall flux magnitude throughout the core as well as a 2-30% difference in discrete energy flux. The coupled production rate (magnitude) and spectral differences can contribute to significant spatial variations in isotope ratios throughout the core. Initial results indicate 239Pu/240Pu ratios vary by as much as 51% across a single CANDU-6 fuel bundle at final burnup. The model is currently being used to develop an accurate representation of spent fuel to perform spatial isotopic analysis across the entire CANDU-6 core.
FAST REACTOR MULTIPHYSICS AND UNCERTAINTY PROPAGATION WITHIN WIMS
Authors: Lindley Ben , Tollit Brendan , Smith Peter , Charles Alan , Mason Robert , Ware Tim , Perry Ray , Lavarenne Jean , Davies Una , Gregg Robert
Published: February 2021

For liquid metal-cooled fast reactors (LMFRs), improved predictive modelling is desirable to facilitate reactor licensing and operation and move towards a best estimate plus uncertainty (BEPU) approach. A key source of uncertainty in fast reactor calculations arises from the underlying nuclear data. Addressing the propagation of such uncertainties through multiphysics calculations schemes is therefore of importance, and is being addressed through international projects such as the Sodium-cooled Fast Reactor Uncertainty Analysis in Modelling (SFR-UAM) benchmark. In this paper, a methodology for propagation of nuclear data uncertainties within WIMS is presented. Uncertainties on key reactor physics parameters are calculated for selected SFR-UAM benchmark exercises, with good agreement with previous results. A methodology for coupled neutronic-thermal-hydraulic calculations within WIMS is developed, where thermal feedback is introduced to the neutronic solution through coupling with the ARTHUR subchannel code within WIMS and applied to steady-state analysis of the Horizon 2020 ESFR-SMART project reference core. Finally, integration of reactor physics and fuel performance calculations is demonstrated through linking of the WIMS reactor physics code to the TRAFIC fast reactor fuel performance code, through a Fortran-C-Python (FCP) interface. Given the 3D multiphysics calculation methodology, thermal-hydraulic and fuel performance uncertainties can ultimately be sampled alongside the nuclear data uncertainties. Together, these developments are therefore an important step towards enabling propagation of uncertainties through high fidelity, multiphysics SFR calculations and hence facilitate BEPU methodologies.
METHODOLOGY FOR HIGH-FIDELITY DETERMINISTIC MODELLING OF SWISS LWR FUEL ASSEMBLIES
Authors: Bernal A. , Pecchia M. , Rochman D. , Vasiliev A. , Ferroukhi H.
Published: February 2021

The main goal of this work is to perform pin-by-pin calculations of Swiss LWR fuel assemblies with neutron transport deterministic methods. At Paul Scherrer Institut (PSI), LWR calculations are performed with the core management system CMSYS, which is based on the Studsvik suite of codes. CMSYS includes models for all the Swiss reactors validated against a database of experimental information. Moreover, PSI has improved the pin power calculations by developing models of Swiss fuel assemblies for the Monte Carlo code MCNP, with the isotopic compositions obtained from the In-Core Fuel Management data of the Studsvik suite of codes, by using the SNF code. A step forward is to use a neutron code based on fast deterministic neutron transport methods. The method used in this work is based on a planar Method of Characteristics in which the axial coupling is solved by 1D SP3 method. The neutron code used is nTRACER. Thus, the methodology of this work develops nTRACER models of Swiss PWR fuel assemblies, in which the fuel of each pin and axial level is modelled with the isotopic composition obtained from SNF. This methodology was applied to 2D and 3D calculations of a Swiss PWR fuel assembly. However, this method has two main limitations. First, the cross sections libraries of nTRACER lack some of the isotopes obtained by SNF. Fortunately, this work proves that the missing isotopes do not have a strong effect on keff and the power distribution. Second, the 3D models require high computational memory resources, that is, more than 260 Gb. Thus, the nTRACER code was modified, so now it uses only 8 Gb, without any loss of accuracy. Finally, the keff and power results are compared with Monte Carlo calculations obtained by Serpent.
FEW-GROUP CROSS SECTIONS LIBRARY BY ACTIVE LEARNING WITH SPLINE KERNELS
Authors: Szames E. , Ammar K. , Tomatis D. , Martinez J.M.
Published: February 2021

This work deals with the representation of homogenized few-groups cross sections libraries by machine learning. A Reproducing Kernel Hilbert Space (RKHS) is used for different Pool Active Learning strategies to obtain an optimal support. Specifically a spline kernel is used and results are compared to multi-linear interpolation as used in industry, discussing the reduction of the library size and of the overall performance. A standard PWR fuel assembly provides the use case (OECD-NEA Burn-up Credit Criticality Benchmark [1]).
A SUBCHANNEL COARSENING METHOD FOR Serpent2-SUBCHANFLOW APPLIED TO A FULL-CORE VVER PROBLEM
Authors: García Manuel , Ferraro Diego , Valtavirta Ville , Tuominen Riku , Imke Uwe , Mercatali Luigi , Sanchez-Espinoza Victor , Leppänen Jaakko
Published: February 2021

This work presents a methodology to coarsen subchannel models to accelerate largescale Serpent2-SUBCHANFLOW calculations. The method is based on first building a fully detailed subchannel model and then combining subchannels into larger channels with condensed hydraulic parameters. To quantify the accuracy and performance of this scheme, a full-core VVER-1000 problem is simulated with Serpent2-SUBCHANFLOW using two types of coarsened models, as well as a standard subchannel model as reference.
TIME-DEPENDENT HOMOGENIZATION FOR PRESSURIZED HEAVY-WATER REACTORS
Authors: Schwanke Peter , Nichita Eleodor
Published: February 2021

A new time-dependent homogenization approach that accounts for inter-assembly leakage has recently been proposed. The new technique extends Generalized Equivalence Theory (GET) to transient simulations through the use of time-dependent, leakage-corrected discontinuity factors that are calculated at each time step by means of a global-local iterative approach to account for the effect of neighbouring nodes so that highly heterogeneous cores are more accurately modelled than when employing single-node, zero-node-boundary-current Assembly Discontinuity Factors (ADFs). The technique has been previously tested for a one-dimensional, two-energy-group, BWR-like benchmark. The present work expands the analysis to a one-dimensional, two-energy-group, Pressurized Heavy-Water Reactor (PHWR) configuration. The PHWR configuration consists of 22 fuel nodes bounded on either side by two nodes of heavy-water (D2O) reflector. Each fuel node spans 28.575 cm and is a one-dimensional stylized representation of a 37-element, natural uranium fuel bundle with D2O coolant residing in a pressure tube that in turn resides in a calandria tube surrounded by D2O moderator. A simple transient induced by instantaneous half-core voiding of the D2O coolant is studied. Three types of calculations are performed: A reference, heterogeneous-node, fine-mesh calculation, a standardly-homogenized-node calculation and a GET-homogenized-node (using ADFs) calculation. The root-mean-square percent errors introduced by standard homogenization and ADF-based homogenization for kinetics calculations in PHWR cores are found to be 4% and 5%, respectively, after 0.5 s. This suggests that the use of a time-dependent homogenization method is desirable, and its use is shown to reduce the RMS errors to a maximum of 0.003% over the course of the transient. The conclusion is that although PHWR cores are not extremely heterogeneous, the accuracy of transient modelling for PHWRs is improved when using time-dependent homogenization over conventional ADFs and that the newly-developed time-dependent homogenization method promises to offer substantial improvements in accuracy for transient results with particular relevance to safety analyses.
ARTIFICIAL INTELLIGENCE DESIGN OF NUCLEAR SYSTEMS EMPOWERED BY ADVANCED MANUFACTURING*
Authors: Sobes Vladimir , Hiscox Briana , Popov Emilian , Delchini Marco , Archibald Richard , Hauck Cory , Laiu Paul , Betzler Ben , Terrani Kurt
Published: February 2021

Implicit in any engineering design is an underlying optimization problem, although the exact objective function to be optimized is rarely stated explicitly. Nuclear systems optimization is as old as the discipline of nuclear engineering. Advanced manufacturing in the nuclear industry has opened the door for the re-examination of optimization in a way in which it was not possible before, namely, determining the optimal geometry for a given objective function. A trivial example is the sphere as the shape that minimizes the volume (or mass) of bare fissile material in a critical configuration. However, the problem becomes less trivial under even the simplest of multiphysics considerations. In this work, we develop the solution methodology for finding the minimum volume geometric configurations under the multiphysics constraints of 1,500 pcm excess reactivity and maximum fuel temperature of 618°C under forced-flow cooling conditions. Constraining the solution geometry only to right cylinders, surprisingly yields two disjoint solution regions. Flat, wide (disk-like) cylinders and tall, narrow (rod-like) cylinders both satisfy the constraints and yield very similar minimal volumes. However, the ultimate pursuit of this work is truly arbitrary geometry.