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This paper presents a graph neural network (GNN) technique for low-level reconstruction of neutrino interactions in a Liquid Argon Time Projection Chamber (LArTPC). GNNs are still a relatively novel technique, and have shown great promise for similar reconstruction tasks in the LHC. In this paper, a multihead attention message passing network is used to classify the relationship between detector hits by labelling graph edges, determining whether hits were produced by the same underlying particle, and if so, the particle type. The trained model is 84% accurate overall, and performs best on the EM shower and muon track classes. The model's strengths and weaknesses are discussed, and plans for developing this technique further are summarised. [ABSTRACT FROM AUTHOR]

Physics event generators are essential components of the data analysis software chain of high energy physics experiments, and important consumers of their CPU resources. Improving the software performance of these packages on modern hardware architectures, such as those deployed at HPC centers, is essential in view of the upcoming HL-LHC physics programme. In this paper, we describe an ongoing activity to reengineer the Madgraph5_aMC@NLO physics event generator, primarily to port it and allow its efficient execution on GPUs, but also to modernize it and optimize its performance on vector CPUs. We describe the motivation, engineering process and software architecture design of our developments, as well as the current challenges and future directions for this project. This paper is based on our submission to vCHEP2021 in March 2021, complemented with a few preliminary results that we presented during the conference. Further details and updated results will be given in later publications. [ABSTRACT FROM AUTHOR]

This paper reports on the development of a collection of dynamical models of one-dimensional peak profile functions used to fit dynamic structure factors S (Q, ħω) of Quasi Elastic Neutron Scattering (QENS) data. The objective of this development is to create a maintainable and interoperable Python library with models reusable in other projects related to the analysis of data from Quasi Elastic Neutron Scattering experiments. The ambition is that the library also will serve as a platform where scientists can make their models available for others. We illustrate how the library can be used by newcomers to the field as well as by experts via different examples. These examples, provided as Jupyter notebooks, show how the QENS models can be integrated in the whole QENS data processing pipeline. [ABSTRACT FROM AUTHOR]

This paper presents a Computational Fluid Dynamics (CFD) model on the effect of capillary pressure on the retention behaviour of a granular material. The model proposes an unprecedented CFD insight into the onset of liquid menisci at the inter-particles contact under varying hydraulic conditions. The present work models the material grains as smooth spherical particles that define a porous network filled by two interstitial fluids: air and silicon oil. The numerical model has been subsequently validated against experimental measurements of the degree of saturation at different capillary pressures taken by Dullien et al. [F.A. Dullien, C. Zarcone, I.F. MacDonald, A. Collins, R.D. Bochard. J. Colloid Interface Sci. 127, 2 (1989)] in a system of smooth glass beads flooded with silicon oil. Results from the numerical simulations confirm the good capability of the model to reproduce the experimental retention behaviour of the granular material. Finally, the present paper laid the basis for future CFD studies on the effect of various factors (e.g. hydraulic hysteresis, surface roughness and/or grain shape) on the capillary pressure acting at the interparticle contact. [ABSTRACT FROM AUTHOR]

It is well recognised that particle breakage in granular materials is affected by stress level, stress path, initial density, and particle size distribution (PSD), amongst others. Furthermore, it has been shown that breakage has a significant influence on the stress-strain behaviour of soils. This paper compares a commonly used breakage parameter with grading entropy coordinates. Such coordinates enable for the representation of any PSD as a single point in a Cartesian coordinate plane. Hence, the evolution of PSD changes may be easily tracked. This paper aims to demonstrate that grading entropy coordinates are as (or more) effective than other breakage parameters, whilst providing additional insight. On the basis of limited data it is shown that grading entropy coordinates are able to capture the dependence of breakage on stress level, stress path and initial PSD. [ABSTRACT FROM AUTHOR]

Particulate matter offers a broad spectrum of phenomena, well known in the granular matter community. This fact is not only related to the individual or collective behaviour that grains can present depending on the external boundary conditions and/or internal interactions but, to the wide range of particle sizes. This feature confers the possibility of studying a particular phenomenon in different size scales. The main goal of this paper is to describe the detachment of particles from surfaces, where particles go from tens of microns to several millimetres, in a variety of scenarios. All the studied systems presented here have in common the need to determine the critical conditions to lose the equilibrium balance, initiating incipient motion. Solving the right time moment for particle resuspension, consideration of the local interactions between particle, surface and external applied forces can be achieved from different modelling approaches, from deterministic dynamics to statistical mechanics. This paper focuses on some of these approaches performed recently on grains through different size scales. [ABSTRACT FROM AUTHOR]

The paper discusses the goals and objectives of creating digital twins of the production system of a machine-building enterprise. The data structure of the information model of the production system of a machine-building enterprise, which is the basis for building a digital twin, is presented. The paper shows the main approaches to managing a production system based on the construction of its digital twin. It is revealed that along with traditional approaches to PS management by forming recommendations in terms of PS engineering and its operation, the choice of the most rational PS management algorithms that take into account the peculiarities of the production process organization and ensure the formation of production schedules that take into account PS reliability indicators has a great potential. It is proposed to use a specialized language of DPML to describe the information model of the PS through the "product-process-resource" paradigm, which ensures the coordination of the processes of forming recommendations in terms of engineering and operation of the PS, as well as the choice of the most rational algorithm for managing the PS. [ABSTRACT FROM AUTHOR]

The collateralized debt obligations and credit default swaps applications are shown in this paper. The industry obligations secondary market risk estimation methods are considered in this work. The new methods taking into account statistically significant parameters for industrial credit derivatives portfolio are offered for single-name investment risks numerical experiments realization. The mathematical estimation of tranche were shown. The single and multiple name default obligations necessary mathematical modeling methods and formulae for the industrial materials manufacturers derivative credit tools market are shown. It is determined that the portfolio of synthetic debt tools is made of the given parameters. The task of a loss derivative tranches mathematical estimation is solved. Late defaults raise the equity tranches payment required sums with high spreads, early defaults reduce. Also the functional characteristics required for an estimation huge debts problem solving are partly considered in this paper. The problem of the default modeling for market tools and numerical simulation of the obligations influence on conditions of current bistability mode are shown here. Some credit derivatives of industrial manufacturers are demonstrated in the modeling process of default as an example. It is found that the model is an additional factor help us to estimate the default opportunity. [ABSTRACT FROM AUTHOR]

The early detection of anomalies through the analysis of the neutron noise recorded by incore and ex-core instrumentation gives the possibility to take proper actions before such problems lead to safety concerns or impact plant availability. The study of the neutron fluctuations permits to detect and differentiate anomalies depending on their type and possibly to characterize and localize such anomalies. This method is non-intrusive and does not require any external perturbation of the system. To effectively use the neutron noise for reactor diagnostics it is essential to accurately model the effects of the anomalies on the neutron field. This paper deals with the development and validation of a neutron noise simulator for reactors with different geometries. The neutron noise is obtained by solving the frequency-domain two-group neutron diffusion equation in the first order approximation. In order to solve this partial differential equation a code based on a high order finite element method is developed. The novelty of this simulator resides on the possibility of dealing with rectangular meshes in any kind of geometry, thus allowing for complex domains and any location of the perturbation. The finite element method also permits automatic refinements in the cell size (h-adaptability) and in its polynomial degree (p-adaptability) that lead to a fast convergence. In order to show the possibilities of the neutron noise simulator developed a perturbation in a hexagonal two-dimensional reactor is investigated in this paper. [ABSTRACT FROM AUTHOR]

This paper describes the application of the Virtual Environment for Reactor Applications, VERA, under development by the Consortium for Advanced Simulation of LWRs (CASL) to the core physics analysis of the AP1000 PWR. The AP1000 PWR features an advanced first core with radial and axial heterogeneities which provides significant enhancements compared to traditional first cores, allowing best fuel usage and short transition to the equilibrium cycle with subsequent fuel reloads. These core advanced features can pose some challenges to the core physics tools making application of VERA to the AP1000 PWR first core especially relevant to qualify VERA performance. This paper focuses on the qualification efforts at hot zero power conditions, where Monte-Carlo reference solutions have also been established. In particular, the paper focuses on the comparison of the predictions obtained with VERA for the four AP1000 units that recently started up with the measured values. It is shown that there is excellent agreement between VERA and the key reactor physics parameters measured during the AP1000 startup. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

To demonstrate that nuclear reactors can be built faster and more economically than they have been in the past, the US Department of Energy Office of Nuclear Energy is sponsoring the development of a small nuclear reactor called the Transformational Challenge Reactor (TCR) [1–2]. An important part of the design and licencing process of a new reactor is validation of the software used to analyze the reactor using established reactor physics benchmarks. This paper discusses validation of the neutronics software used to model four preliminary designs of the TCR core [2]. Because the TCR core design uses innovative technology and methods, comparable established benchmarks are limited or do not exist. For this effort, established benchmarks from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP) [3] were considered to be suitable for this design based on analysis using the SCALE/TSUNAMI-computed similarity indices to determine the amount of shared uncertainty between the design and each selected benchmark experiment. This paper addresses the challenges faced in benchmarking a unique reactor for licensing and construction, a task that will become more common as a new generation of innovative nuclear reactors are designed and built. [ABSTRACT FROM AUTHOR]

This paper overviews the initial results of a new project at the Oak Ridge National Laboratory, supported via an internal seed funding program, to develop a novel computational capability for model validation: MAPPER. MAPPER will eliminate the need for empirical criteria such as the similarity indices often employed to identify applicable experiments for given application conditions. To achieve this, MAPPER uses an information-theoretic approach based on the Kullback-Leibler (KL) divergence principle to combine responses of available or planned experiments with application responses of interest. This is accomplished with a training set of samples generated using randomized experiment execution and application of high-fidelity analysis models. These samples are condensed using reduced order modeling techniques in the form of a joint probability distribution function (PDF) connecting each application response of interest with a new effective experimental response. MAPPER's initial objective will be to support confirmation of criticality safety analysis of storage facilities which require known keff biases for safe operation. This paper reports some of the initial results obtained with MAPPER as applied to a set of critical experiments for which existing similarity-based methods have been shown to provide inaccurate estimates of the biases. [ABSTRACT FROM AUTHOR]

This paper investigates the applicability of surrogate model optimization (SMO) using deep learning regression models to automatically embed knowledge about the objective function into the optimization process. This paper demonstrates two deep learning SMO methods for calculating simple neutronics parameters. Using these models, SMO returns results comparable with those from the early stages of direct iterative optimization. However, for this study, the cost of creating the training set outweighs the benefits of the surrogate models. [ABSTRACT FROM AUTHOR]

Nuclear fuel cycle scenario study can be of benefit for decision-making in nuclear industry development. However, due to the lack of knowledge about the future, many scenario studies are subject to uncertainties. As a result, some parameters may be disrupted. Energy production of an entire nuclear fleet of interest is such a parameter. Its disruptive decline has a strong impact on the results of scenario studies. Resilience study against disruption of decline in energy production is required to anticipate possible failures of scenario studies. In such a resilience study, a large number of scenario trajectory simulations with different modes of decline are expected. It is too difficult to set the values of all scenario parameters in each trajectory manually. As a solution, a method is proposed in this paper to reduce the number of input parameters. A set of rules has been implemented as an intermediate layer between the decision-makers and the scenario code to drive the evolution of a nuclear fleet. In this case, a trajectory can be fully characterized by a few parameters. This method has been applied to a simplified academic nuclear fleet with two different modes of decline in energy production. The simulation results showed that the developed method is feasible. One should note that the scenario model in this paper is only used for academic purposes and does not correspond to any industrial strategy or policy. [ABSTRACT FROM AUTHOR]

Verification and validation (V&V) results of source term calculation capability implemented in the nodal diffusion code RAST-K are presented in this paper. An isotope inventory prediction method is presented in this work which is implemented with RAST-K and the lattice code STREAM. STREAM generates cross-section and provides number density information by history branch calculations. RAST-K supplies the power history and three history indexes (boron concentration, moderator temperature and fuel temperature). The main feature of the newly implemented spent nuclear fuel (SNF) characterization is the direct consideration of three-dimensional (3D) core simulation conditions by using operation history information. As a result of this, it could reduce the computation time. The implemented SNF analysis capability have two main functions. The first is to predict isotope inventory by Lagrange non-linear interpolation method, using power history correction factors. The second is to calculate the radiological response activity, decay heat, and neutron/gamma source strengths. The V&V of these two functions are thus presented herein. The isotope inventory prediction is validated with measured data from ten SNF samples of Takahama-3 and six samples of Calvert Cliffs-1 pressurized water reactors (PWR). Eighteen decay heat measurements of Ringhals Unit 3 PWR fuel assemblies are then employed to validate the decay heat calculation results. In addition, STREAM is employed in a code-to-code comparison for verification. The fuel assemblies cover the burnup range 14.3 - 47.25 GWd/tU, initial enrichment of 2.1 - 4.11 235U w/o and cooling time of 3.96 to 20.01 years. The comparison to STREAM shows the accuracy of the RAST-K SNF and prediction of the decay heat is within 4%. Overall, this paper demonstrates that RAST-K SNF calculation can be applied to the back-end cycle source term analysis. [ABSTRACT FROM AUTHOR]

The paper deals with design of new SCRAM experimental device at Training reactor VR-1. Crucial works cowers the detector development, transient calculations and principal design of the device. Conceptual design will be prepared in 2019–2020. This paper presents current work status. [ABSTRACT FROM AUTHOR]

Recently a joint project has been carried out between the Paul Scherrer Institut, the Ecole Polytechnique Federale de Lausanne and swissnuclear, an industrial partner, in order to determine the axial void distribution in a channel installed in the reflector of the zero power research reactor CROCUS, using neutron noise techniques. The main objective of the present paper is to report on the validation of the results against an alternative measurement technique using gamma-ray attenuation and simulations with the TRACE code. For the gamma-ray attenuation experiments, the channel used in CROCUS is installed out of the core in a Plexiglass water tank. The source and detector are fixed and the channel is moved axially to keep the geometry of the source/detector arrangement untouched. This is key to measure the void effect by gamma attenuation due to the low contrast of this technique. The paper compares the experimental results obtained with both techniques, with the outcomes of simulations carried out with the TRACE code. Even though the quantitative void fraction estimations are not consistent, the trends obtained with the simulation and experimental techniques are the same. The discrepancies between the various experimental techniques and the simulation outcomes are related to the heterogeneous distribution of the water-air mixture in the radial sections of the channel. [ABSTRACT FROM AUTHOR]

Accidents prevention and mitigation of the related consequences are mandatory for the safety assessment of the Generation IV Sodium-cooled Fast Reactors (SFRs). In order to reach these goals, the accurate evaluation of the most challenging transients, which may lead to severe accidents, is fundamental. In particular, an accurate evaluation of the degradation of the thermo-mechanical properties of the fuel pins due to irradiation, affecting the course of accidental transient conditions, is necessary. In our studies, the reference calculation route for the analysis of a severe accident in SFRs is based on the sequential use of the SAS-SFR and the SIMMER-III/IV mechanistic codes for the initiation (up to hexcan failure of a fuel sub-assembly) and the late (with core melting) phases of the scenario, respectively. Nevertheless, the module devoted to the SAS-SFR/SIMMER data transfer shows some limitations when innovative pins and core designs are considered. Having this in mind, the present paper proposes an innovative approach based on the coupling of a fuel performance code FEMAXI and the SIMMER code. When applied to severe accident studies of new reactors designs, such an approach is expected to improve the consistency of the SIMMER results as well as to extend its range of application. The FEMAXI/SIMMER coupling is described in the paper. Further, the results of the code validation against the CABRI-2 E9 experimental test, performed at the CABRI reactor using an OPHELIE-6 annular fuel pin irradiated in the PHENIX reactor, are shown and discussed. In the E9 test, a slow power ramp non-failure transient, corresponding to control rod withdrawal conditions, was investigated. New FEMAXI/SIMMER results are compared also to stand-alone SAS-SFR and SIMMER evaluations. [ABSTRACT FROM AUTHOR]

There were two similar series of burst waiting time experiments carried out on Godiva-II and Caliban respectively and analyzed for neutron initiation study, unfortunately both the strength of spontaneous fission neutron source given in the published papers are not consistent with the spontaneous fission data of U-235 and U-238. In this paper, the discrepancies of spontaneous fission data of U-235 and U-238 from different references are reviewed and clarified and the correct strength of spontaneous fission neutron source for Godiva-II and Caliban are calculated respectively, then the multiplication equivalence method of arbitrary source is introduced and the equivalent fundamental-mode(EFM) co-efficiencies of spontaneous fission neutron source for Godiva-II and Caliban are calculated with MC software, lastly the strength of neutron source for these two HEU metal burst reactor in EFM are deduced. The EFM neutron strength of this two burst reactors are much less than proposed values used for burst waiting investigation in published papers respectively, and with the new data, the inconsistency of burst waiting time to the published results is presented and discussed. [ABSTRACT FROM AUTHOR]

This paper summarizes the latest improvements and lessons learned from the modeling and simulation of the transient test reactor at Idaho National Laboratory using the MAMMOTH reactor physics application. MAMMOTH is a MOOSE-based, Finite Element Method application that specializes in the analysis of the spatial dynamics behavior of nuclear reactors. Since early 2018 several transient tests have been conducted at TREAT, thus providing the opportunity to apply and benchmark modern modeling and simulation tools. MAMMOTH was used to provide predictions of the power coupling factor between the core and the experiment for various experiments. Even though the power coupling factor predictions agree very well with the experimental data, within the bounds of the experimental uncertainty, one shortcoming was the underprediction of the total energy deposited in the core and experiment. Determination of the sources for this discrepancy is ongoing, but several key problems have been identified and resolved, thus providing valuable insights for future research. This paper discusses several of these lessons learned. First, the heat capacity data for the TREAT fuel has some significant problems due to limitations of the measurement techniques used circa 1960s. The sensitivity of the peak power and the total energy deposition to various representations of the heat capacity is approximately 5%. Second, the effects of the biological shield and thermal column on the modeling of the core are non-negligible, since they affect the mean generation time and the effective reflection of neutrons back into the core, which is suspected to be important during the core heat up. Matching the reactor period resolves the fact that the reduced spatial domain used in the MAMMOTH model underpredicts the mean generation time. The neutron reflection from these regions is marginally improved with the use of an albedo boundary condition. Third, modeling of the control rod movement with a multi-scheme method is introduced and its current limitations are exposed. Fourth, we explore the effects of using a homogenized model with Superhomogenization equivalence and how that differs from fully heterogeneous simulations. Finally, the energy condensation effects for this graphite core are significant. Solutions with 10 and 26 energy groups show the benefits of using a finer coarse group structure. [ABSTRACT FROM AUTHOR]

The heat pipe cooled reactor is a solid-state reactor using heat pipes to passively transfer heat generated from the reactor, which is a potential and near-term space nuclear power system. This paper introduces the coupling scheme between the continuous energy Reactor Monte Carlo (RMC) code and the finite element method commercial software ANSYS. Monte Carlo method has the advantages of flexible geometry modeling and continuous-energy nuclear cross sections. ANSYS Parametric Design Language (APDL) is used to determine the detailed temperature distributions and geometric deformation. The on-the-fly temperature treatment of cross sections was adopted in RMC code to solve the memory problems and to speed up simulations. This paper proposed a geometric updating strategy and reactivity feedback methods for the geometric deformation of the solid-state core. The neutronic and thermal-mechanical coupling platform is developed to analyze and further to optimize the heat pipe cooled reactor design. The present coupling codes analyze a 2D central cross-section model for MEGAPOWER heat pipe cooled reactor. The thermal-mechanical feedback reveals that the solid-state reactor has a negative reactivity feedback (~1.5 pcm/K) while it has a deterioration in heat transfer due to the expansion. [ABSTRACT FROM AUTHOR]

Neutron clustering is a recently identified problem with Monte Carlo eigenvalue calculations which can produce significantly erroneous results. Previous work by Sutton & Mittal (2017) considered neutron clustering as a problem of maintaining 'genetic diversity' within the neutron population. This paper proposes reducing the extent of neutron clustering by replacing fission neutrons in the source bank with uncorrelated neutrons, sampled from a uniform source distribution – effectively adding new neutron genealogies to the population. The efficacy of the method is demonstrated on a number of simple problems, showing improved behaviour of the Shannon entropy and neutron centre-of-mass. Although currently limited in scope, this paper intends to provide a route to reducing clustering effects in more general problems. [ABSTRACT FROM AUTHOR]

This paper presents the application and evaluation of a deterministic truncation of Monte Carlo (DTMC) solution method in a whole core reactor problem based on a continuous energy transport calculation. The DTMC method has been studied and developed as a systematic way to truncate the high-fidelity Monte Carlo (MC) solution to reduce the computational cost without compromising the essential reliability of the solution. Its fea-sibility and capability were preliminarily validated in several benchmark problems using a multi-group energy MC code. In this paper, further study has been conducted in the more practical application with the continuous-energy based MC calculation. The con-cept of the DTMC method is briefly described. Improvements to enhance the numerical stability and efficiency are specified in details. The DTMC method is applied to an SMR problem, in which reactor parameters are estimated to characterize the numerical per-formance and are compared to the standard MC method. Last, the computing time and corresponding figure-of-merit are evaluated. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Studsvik's approach to spent nuclear fuel analyses combines isotopic concentrations, fluxes, and cross-sections, calculated by the CASMO5 neutron transport and depletion code, with irradiation history data from the SIMULATE5 reactor core simulator and tabulated isotopic decay data. These data sources are used and processed by the SNF code to compute the spent nuclear fuel characteristics. Recent advances in the system, including cross-sections and decay data from ENDF/B-VIII.R0, are presented in this paper, together with validation results against decay heat power and isotopic compositions measurements. Measurements conducted at the Swedish interim storage facility, CLAB, are used for validation of the decay heat power, while comparisons to the results of the international program ARIANE are used to demonstrate the capability of CMS5/SNF to accurately predict isotopic compositions. The paper shows the results calculated with ENDF/B-VIII.R0, and the effect on the spent fuel characteristics is evaluated by comparisons to the earlier ENDF/B-VII.R1 results. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

The problem of determining the axial velocity profile from the in-core neutron noise measurements is revisited, with the purpose of developing an objective method for the determination of the void fraction. Until now it was assumed that in order to determine a realistic velocity profile which shows an inflection point and hence has to be at least a third order polynomial, one needs four transit times and hence five in-core detectors at various axial elevations. However, attempts to determine a fourth transit time by adding a TIP detector to the existing four LPRMs and cross-correlate it with any of the LPRMs were unsuccessful so far. In this paper we thus propose another approach, where the TIP detector is only used for the determination of the axial position of the onset of boiling. By this approach it is sufficient to use only three transit times. Moreover, with another parametrisation of the velocity profile, it is possible to reconstruct the velocity profile even without knowing the onset point of boiling, in which case the TIP is not needed. In the paper the principles are explained and the strategy is demonstrated by concrete examples. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

A conference paper on the role of monte carlo calculation in machine and radiation protection of high energy accelerators is discussed. It outlines beam losses, beam machine interaction and machine cases. The paper also examines particle interaction with accelerator material. Topics include beam loss monitor system, luminosity losses and synchrontron radiation.

This paper reviews select examples of recent targets produced by the group, including 171Yb wires, 62Ni pellets, and 25Mg foils. The targets discussed will cover the processes and equipment used for stable isotope manufacturing at ORNL and give a broader understanding of thin film preparation techniques, preparation of high-purity and specialty materials, and development of highly material-conservative methods. [ABSTRACT FROM AUTHOR]

Nuclear powder targets for neutron-induced cross section measurements were prepared by pressing. The choice of the production technique was related to the type of nuclear experiment and the quality of the powder. This paper describes the production process of unsintered compacts of 94Mo, 95Mo, 96Mo metal powders and 239PuO2 powder with a thickness as low as possible by uniaxial pressing of powders. Special attention goes to the difficulties encountered during the preparation process of 239PuO2 pellets. Investigation on the presence of impurities and so the quality of the PuO2 material was performed by scanning electron microscopy with energy dispersive X-ray spectroscopy, by inductively coupled plasma mass spectrometry, X-ray diffraction and by thermal analysis. [ABSTRACT FROM AUTHOR]

During Long Shutdown 2 (LS2) at CERN, the new Linac4 (L4) accelerator was connected to the PS Booster (PSB) to inject 160 MeV H− beam into the 4 superposed PSB rings. In order to achieve this, a completely new H− charge-exchange injection chicane system, with 200 µg/cm2 carbon stripping foil units, to convert the negative hydrogen ions into protons, by stripping off the electrons, has been installed. Beam commissioning of this system started in December 2020 and different types of foils have been used for this purpose. In parallel, stripping efficiency measurements, of different foil types, continued to take place with the stripping foil test stand in the L4 Transfer Line. This paper briefly recalls the final design, and installation of, the main components in the PSB injection region, before reporting on the important results obtained during measurements and commissioning with beam. [ABSTRACT FROM AUTHOR]

GRE@T-PIONEeR - GRaduate Education Alliance for Teaching the PhysIcs and safety Of NuclEar Reactors - is a project funded by the Euratom – Horizon 2020 Framework Programme which aims at developing and providing specialised and advanced courses in computational and experimental reactor physics at the graduate level (MSc and PhD levels) and post-graduate level, as well as the staff members working in the nuclear industry. One of the work packages of GRE@T-PIONEeR is devoted to developing a specific course on the nuclear data pipeline processes and to present the role of nuclear data to play in calculations of innovative reactor systems. This course covers all steps in the nuclear data life cycle, starting from the measurements to their validation and final use in nuclear reactor calculations. Beyond the technical contents of the courses being developed, the paper describes the use of innovative pedagogical methods and active learning techniques, such as flipped classes, aimed at promoting student learning. [ABSTRACT FROM AUTHOR]

The thermal scattering laws (TSL) take into account the crystalline structure and atomic motions of isotopes bound in materials. This paper presents the CINEL code, which was developed to generate temperature-dependent TSL for solid, liquid and free gas materials of interest for nuclear reactors. CINEL is able to calculate TSL from the phonon density of states (PDOS) of materials under the Gaussian-Incoherent approximations. The PDOS can be obtained by using theoretical approaches (e.g., ab initio density functional theory and molecular dynamics) or experimental results. In this work, the PDOS presented in the ENDF/BVIII.0 and NJOY-NCrystal libraries were used for numerical validation purposes. The CINEL results are in good agreement with those reported in these databases, even in the specific cases of TSL with the newly mixed elastic format. The coding flexibility offered by Python using the JupyterLab interface allowed to investigate limits of physical models reported in the literature, such as a four-site model for UO2, anharmonic behaviors of oxygen atoms bound in a Fm3m structure, texture in Zry4 samples and jump corrections in a roto-translational diffusion model for liquid water. The use of graphic processing units (GPU) is a necessity to perform calculations in a few minutes. The performances of the CINEL code is illustrated with the results obtained on actinide oxides having a Fm3m structure (UO2, ThO2, NpO2 and PuO2), low enriched fuel (UMo), cladding (Zry4) and moderators (H2O with a specific emphasis on ice). [ABSTRACT FROM AUTHOR]

In this paper we present results of combining ring-polymer molecular dynamics with the LEAPR and THEMR modules of NJOY to generate thermal neutron scattering libraries for liquid para-hydrogen and ortho-deuterium. We present the methodology and show that it produces results that are in good agreement with data from both recent available measurements and previous theoretical studies. We also present some simple benchmark Monte-Carlo simulations compared with other available libraries. [ABSTRACT FROM AUTHOR]

NaOH was recently proposed as a moderator for the compact molten salt reactors, under development by Seaborg Technologies. In order to reliably predict the thermalisation of neutrons in such a reactor, the simulations must be based on modelling which accurately accounts for the low-energy neutron interactions in NaOH. This information is contained within thermal scattering libraries that are used during Monte-Carlo simulations. In this paper, we present the results of neutron scattering experiments carried out at VESUVIO, ISIS in order to facilitate the development thermal scattering libraries for solid NaOH. [ABSTRACT FROM AUTHOR]

Precise estimation of neutron thermalization in moderators relies on high-fidelity thermal scattering cross-section data governed by the thermal scattering law (TSL). The Institut de Radioprotection et de Sûreté Nucléaire (IRSN) has been working on the development of improved TSL for light water ice. Many polymorphic phases of light water ice exist depending on the thermodynamic conditions. The most common type of ice at standard pressure and temperature below water freezing point (273.15 K) is ice-Ih. It is essential to have high-resolution experimental double differential data for developing and/or validating TSL for moderator materials.Existing experimental double-differential scattering data for ice-Ih are extremely sparse and of limited quality. New high-quality double-differential measurements for ice-Ih over multiple temperatures and incident neutron energies would directly support the validation and improvement of ice-Ih TSL models for criticality safety applications. Series of time-of-flight (TOF) inelastic neutron scattering experiments on ice-Ih at temperatures starting at 271 K and down to 6 K, have been carried out at the SEQUOIA spectrometer at the Spallation Neutron Source (SNS) at the Oak Ridge National Laboratory (ORNL), United States. The experiments have been performed for incident neutron energies, Ei= 11, 55, 160, 250, and 600 meV, to explore different excitation energies in the vibrational phonon spectrum. This paper presents the thermodynamic conditions and the details of the TOF measurements on ice-Ih and the derived phonon spectrum from the experimentally measured double differential data. A study of the variation of the phonon spectrum of ice-Ih as a function of temperature is highlighted, and a preliminary TSL evaluation is developed based on the experimental phonon spectrum. [ABSTRACT FROM AUTHOR]

Accurately predicting the criticality of an experiment before interacting with the experimental components is very important for criticality safety. Radiation transport software can be utilized to calculate the effective neutron multiplication factor of a nuclear system. Because of the integral nature of the effective neutron multiplication factor, the value calculated contains various sources of nuclear-data induced uncertainty. The sensitivity analysis and data assimilation technique presented in this paper exhibit one possible method of identifying and reducing the effective neutron multiplication factor nuclear-data induced uncertainty. The results presented in this work show that it is possible to use relative sensitivity coefficients of the prompt neutron decay constant and the effective delayed neutron fraction to 239Pu nuclear data to reduce nuclear-data induced uncertainties in the effective neutron ultiplication factor. This work has been utilized by members of the Los Alamos National Laboratory project EUCLID (Experiments Underpinned by Computational Learning for Improvements in Nuclear Data) for optimally designing a new experiment, which will be used to reduce compensating errors in 239Pu nuclear data. [ABSTRACT FROM AUTHOR]

JENDL-5, the latest version of the Japanese evaluated nuclear data library, includes several sub-libraries to contribute to various applications. In this paper, we outline the evaluation and validation of the deuteron reaction sub-library developed mainly for the design of accelerator-based neutron sources and the alpha-particle reaction sub-library developed mainly for use in the back-end field. As for the deuteron sub-library, the data for 6,7Li, 9Be, and 12,13C from JENDL/DEU-2020 were partially modified and adopted. The data up to 200 MeV for 27Al, 63,65Cu, and 93Nb, which are important as accelerator structural materials, were newly evaluated based on the calculations with the DEURACS code. As for the alpha-particle sub-library, the data up to 15 MeV for 18 light nuclides from Li to Si isotopes were evaluated based on the calculations with the CCONE code, and then only the neutron production cross sections were replaced with the data of JENDL/AN-2005. Validation on neutron yield by Monte Carlo transport simulations was performed for both sub-libraries. As a result, it was confirmed that the simulations based on the sub-libraries showed good agreement with experimental data. [ABSTRACT FROM AUTHOR]

Nuclear data are used for a variety of applications, including criticality safety, reactor performance, and material safeguards. Despite the breadth of use-cases, the effective neutron multiplication factor, keff, of ICSBEP critical assemblies are primarily used for nuclear data validation; these are sensitive to specific energy regions and nuclides and are unable to uniquely constrain nuclear data. As a consequence, general-purpose nuclear data libraries, such as ENDF/B-VIII.0 [1], may have deficiencies that, while not apparent in criticality applications, negatively impact other applications, such as non-destructive analysis of special nuclear material [2, 3] and neutron diagnosed subcritical experiments [4]. Recent work by the Experiments Underpinned by Computational Learning for Improvements in Nuclear Data (EUCLID) project developed a machine learning tool, RAFIEKI, which uses random forests and the SHAP metric to determine which nuclear data contribute most to predicted bias between measured and simulated responses (e.g. keff). This paper contrasts RAFIEKI analysis applied to keff only against RAFIEKI analysis with keff paired with either LLNL pulsed sphere measurements or subcritical benchmarks. Two examples show that a) including pulsed sphere measurements substantially increases 9Be nuclear data importance to bias between 2 and 15 MeV, and b) including subcritical benchmarks has the potential for disentangling compensating errors between 240Pu (n,el) and (n,il) cross-sections between 0.1 and 10 MeV. These results show that RAFIEKI analysis applied to response sets that include, but go beyond, keff can aid nuclear data evaluators in identifying issues in nuclear data. [ABSTRACT FROM AUTHOR]

The status of the two neutronics international benchmark projects sanctioned by the Organisation for Economic Co-operation and Development Nuclear Energy Agency (OECD NEA), the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP), was last directly discussed with the international nuclear data community at the 14th International Conference on Nuclear Data for Science and Technology (ND2019) in Beijing, China. Since ND2019, the quantity of available integral benchmark experiment data has increased. The primary purpose of the ICSBEP and IRPhEP is to provide extensively peer-reviewed benchmark data to the international nuclear community in support of validation and testing of nuclear data and models. A total of 28 countries have contributed to the past and continued success of these projects as benchmark evaluations, technical reviews, and experimental data using their own time and resources: 26 to the ICSBEP and 25 to the IRPhEP. Key contributions to the handbooks over the past three years can only be highlighted within this paper. Full technical details and benchmark experiment descriptions can be located within the benchmark reports distributed within recent editions of the handbooks. [ABSTRACT FROM AUTHOR]

The use of calculated dosimetry metrics is a crucial element in predicting neutron damage to modern semiconductors under various conditions, e.g., control electronics in avionic systems, satellite sensors, or power output from solar panels. These dosimetry metrics have little value unless they are accompanied by a quantified uncertainty. This paper outlines a mathematical framework that captures the response models for most semiconductor damage metrics and addresses some of the challenges faced in quantifying the relevant sources of uncertainty. The energy-dependent correlations in the damage functions are a critical underpinning in propagating the uncertainty back to a measured quantity. Significant issues are associated with "model defect" in some of the models used, i.e., assumptions in the model form that are not easily considered in the uncertainty estimate. Other issues relate to fundamental differences between the experimental measured quantity and the calculated metric used to represent the damage mode. [ABSTRACT FROM AUTHOR]

In the last years, a new version of the Joint Evaluated Fission and Fusion File (JEFF) data library, namely JEFF-3.3, has been released with relevant updates in the neutron reaction, thermal neutron scattering and covariance sub-libraries. In the frame of the EU H2020 SANDA project, severale efforts have been made to enable the use of JEFF nuclear data libraries with the extensively tested and verified SCALE Code System. With this purpose, AMPX processing code has been applied to enable such application, allowing to provide insight into the interaction between the code and the new versions of JEFF data file. This paper provides an overview about the processing of JEFF-3.3 nuclear data library with AMPX for its application within the SCALE package. The AMPX-formatted cross-section library has been widely verified and tested using a comprehensive set of criticality benchmarks from ICSBEP, by comparing both with results provided by other processing and neutron transport codes and experimental. Processing of JEFF-3.3 covariances is also addressed along with their corresponding verification using covariances processed with NJOY. This work paves the way towards a successful future interaction between JEFF libraries and SCALE. [ABSTRACT FROM AUTHOR]

FUDGE (For Updating Data and Generating Evaluations) is an open-source code that supports reading, visualizing, checking, modifying, and processing nuclear reaction and decay data. For ease of use the front-end of FUDGE is written in Python while C and C++ routines are employed for computationally intensive calculations. FUDGE has been developed primarily at Lawrence Livermore National Laboratory (LLNL) with contributions from Brookhaven National Laboratory (BNL). It is used by the LLNL Nuclear Data and Theory (NDT) group to deliver high-quality nuclear data libraries to users for a variety of applications. FUDGE is also the world leader in converting data to the Generalized Nuclear Database Structure (GNDS) and working with GNDS data, including processing and visualizing. GNDS is a new extensible hierarchy that has been internationally adopted as the new standard for storing and using nuclear data libraries, replacing the previous standard ENDF-6. A new public release of FUDGE has recently been published on github. This paper gives an overview of nuclear data processing capabilities in FUDGE, as well as describing the latest release, new capabilities, future plans, and basic instructions for users interested in applying FUDGE to their nuclear data workflow. [ABSTRACT FROM AUTHOR]

Since 2016 the Nuclear Data Section of the International Atomic Energy Agency (NDS/IAEA) has encouraged the development of new codes for processing evaluated nuclear data for applications. The main motivation has been the need for improving processing methods and avoiding a common failure mode. The NDS/IAEA has supported the development of the processing system ACEMAKER as a natural extension of the PREPRO family for nuclear data processing with the aim of producing ACE-formatted files for Monte Carlo calculations. The new ACEMAKER modules can process any energy-angle distributions and gamma-production data. The system can generate ACE- formatted files for continuous-energy neutron and photonuclear data, thermal scattering laws and dosimetry cross sections, which can be used for Monte Carlo simulations. The paper summaries the ACEMAKER key features, ongoing verification and validation process, and planned future developments [ABSTRACT FROM AUTHOR]

FOOT (FragmentatiOn Of Target) is an applied nuclear physics experiment that aims at a full characterization of the nuclear fragmentation processes of interest for Particle Therapy and Radiation Protection in Space. The physics program foresees a set of measurements in direct and inverse kinematics using particle beams and targets with composition similar to human tissues and spacecraft shielding materials. The final goal of the experiment is the measurement of double differential cross sections with respect to angle and fragment energy in the 100-800 MeV/u range with a precision better than 5%. The FOOT Collaboration is currently completing the development of the apparatus and data acquisition campaigns with partial setups have already started. In this paper, an overview of the current status of the experiment is given, together with a summary of the preliminary results obtained from the first measurements with 16O beams. [ABSTRACT FROM AUTHOR]

Few data exist regarding the yields and time spectra of delayed neutrons from the photofission reaction on actinides. In this paper, we report the characteristics of both six-group and eight-group delayed neutron time models for the photofission of 238U induced by a Bremsstrahlung photon spectrum with an end-point energy at 9 MeV. This study was conducted using a 9 MeV linear electron accelerator housed at the SAPHIR platform located at CEA Paris-Saclay, France. A reference sample of depleted uranium was positioned inside a neutron detection block, previously designed and optimized by Monte Carlo simulation, made of high-density polyethylene covered by cadmium and including ten one meter-long 3He proportional counters. Based on the consistent set of half-lives proposed by Spriggs et al. as early as 1999 for neutroninduced fission, our description in eight delayed neutron groups for 238U at 9 MeV is the first published for photofission. These new results bring valuable information to the scarce data available in the literature and represent useful inputs for the development of photofission-based nuclear measurement systems as they meet a significant interest in the fields of nuclear data, radioactive waste package characterization and for security-related applications. [ABSTRACT FROM AUTHOR]

The intent of this paper is to present the resolved resonance evaluation of 16O in the energy range from thermal to 6 MeV. The newness of the present work is that recent cross-section data for the 16O(n,α) 13C reaction taken at the GELINA time-of-flight facility and transmission data obtained at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR) were included in the evaluation. The evaluation was carried out with the SAMMY code. The evaluation was used to calculate critical benchmark experiment sensitive to the 16O cross sections. [ABSTRACT FROM AUTHOR]