bluuebunny/crossref_metadata_2025 huggingface dataset:
RECONSTRUCTING THE AXIAL VOID VELOCITY PROFILE IN BWRs FROM MEASUREMENTS OF THE IN-CORE NEUTRON NOISE
Authors: Pázsit Imre , Hursin Mathieu , Nylén Henrik , Dykin Victor
Published: February 2021

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.
ANALYSIS OF THE BEAVRS BENCHMARK USING CASMO5/SIMULATE5 WITH JENDL-4.0 AND ENDF/B-VII.1
Authors: Tatsuya Fujita , Tomohiro Sakai
Published: February 2021

The BEAVRS benchmark was analyzed using the CASMO5/SIMULATE5 in order to compare the measurement data and the calculation results based on the JENDL-4.0 and ENDF/B-VII.1 and investigate the difference between those calculation results. For the hot zero power (HZP) physics test, the calculation results showed good agreement with the measurement data for both of cycles 1 and 2. For cycle 1, the calculation results of the isothermal temperature coefficient (ITC) differed from the measurement data by approximately 1 pcm/℉, and the same tendency has been reported in previous studies. For the cycle operation, the calculation results of the boron letdown agreed well with the measurement data. On the other hand, some calculation results of the axial detector signals had a large difference from the measurement data, which is supposedly attributed to the discrepancy of the axial offset (AO) caused by the authors’ approximation for the control rod and shutdown bank positions. In terms of the comparison between the JENDL-4.0 and ENDF/B-VII.1, although approximately 15 ppm difference of the boron letdown in the cycle operation was observed, no significant difference was seen for other core parameters, thus, the influence of the two nuclear data library was small on the present results.
A SUBGROUP METHOD BASED ON THE EQUIVALENT DANCOFF-FACTOR CELL TECHNIQUE COMPARED WITH THE FINE-STRUCTURE METHOD IN APOLLO3®
Authors: Mao Li , Zmijarevic Igor , Sanchez Richard , Castaing Nicolas Gerard
Published: February 2021

We compare a newly implemented subgroup method, which is based on the equivalent Dancoff-factor cell technique, with the fine structure method in APOLLO3®. The new method obtains precise reaction rates and consequently is precise in predicting multipli-cation factor. It can reduce the time in the self-shielding calculation by a factor of 38 compared with the subgroup method based on the multi-cell approximation in a PWR Gd-UOX assembly calculation. The fine structure method obtains larger errors in reac-tion rates than the subgroup method, but the error compensation sometimes leads to small errors in multiplication factors. This work demonstrates the precision and the efficiency of the new method.
APPLICATION OF CALCULATED MSM FACTORS USING TRIPOLI4® SEQUENCE ON BORON LINED PROPORTIONAL COUNTER ROD WORTH MEASUREMENT
Authors: Couybes J. , Nicolas S. , Chabert L. , Manifacier L. , Barbosa T. , Gerlier R.
Published: February 2021

The topic addressed deals with the determination of adjoint parameters for instrumentation relevance. This is a crucial subject for comprehension of subcritical levels in the frame of safety analysis. Indeed, such states require interpretation and raw data cannot be processed as such. To do so, the transcription of core reactivity through instrumentation located in the reactor periphery is considered with the use of MSM factors [1],[2]. We implement this method inside a TRIPOLI4® [3] sequence in order to establish predictive mapping of MSM factors and figure out optimal position for instrumentation location at the beginning of reactor operations.

Firstly, MSM factors are introduced, along with the designer point of view for geometry construction based on ROOT package [4]. At this point, the methodology of TRIPOLI4® calculation is explained in detail, including the sequencing associated to and how the Green Functions are performed within TRIPOLI4®.

In this second part and within the verification framework, the previous method is extended to a “fictitious core” developed in TechnicAtome for Monte Carlo [5] calculation and for different core pattern loadings. After the completion of these numerical validations gained on a High Performing Cluster, the method is then expanded to critical mock up [6] and challenged to recent experimental results for validation. The comparisons end up with a good agreement between predictive calculation and experimental values of reactivity worth.

Finally the document ends with a mid-term projection for outlooks and improvements, for ensuring an enhancement of the safety approach. Several items are discussed especially, fine tuning for the spatial meshing (regarding instrumentation size) and the impact on TRIPOLI4® Monte Carlo code with the development of new features. Then, the authors focus on sensitivity effect concerning delayed neutron spectrum and kinetics parameters. As a conclusion, this paper proposes to validate the method exposed in the near future, using experimental data coming from many years of critical mock up operations.
EFFECT OF FISSION SOURCE SPECTRUM ON MONTE CARLO CALCULATION OF EX-CORE QUANTITIES
Authors: Davidson Eva E. , Pandya Tara M. , Royston Katherine E. , Evans Thomas M. , Godfrey Andrew T. , Henderson Shane C. , Wolfram Gary , Risner Joel M.
Published: February 2021

The Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) offers unique capabilities to combine highfidelity in-core radiation transport with temperature feedback using MPACT and CTF with a follow-on fixed source transport calculation using the Shift Monte Carlo code to calculate ex-core quantities of interest. In these coupled calculations, MPACT provides a fission source to Shift for the follow-on radiation transport calculation. In past VERA releases, MPACT passed a spatially dependent source without the energy distribution to Shift. Shift then assumed a235U Watt spectrum to sample the neutron source energies. There were concerns that, in cases with burned or mixed oxide (MOX) fuel near the periphery of the core, the assumption of a235U Watt spectrum for the source neutron energies would not be accurate for studying ex-core quantities of interest, such as pressure vessel fluence or detector response. Therefore, two additional options were implemented in VERA for Shift to sample neutron source energies: (1) a nuclide-dependent Watt spectra for235U,238U,239Pu, and241Pu, and (2) to use the standard 51-energy group MPACT spectrum. Results show that the 51-group MPACT spectrum is not suitable for ex-core calculations because the groups have been fine-tuned for in-core calculations. Differences in relative detector response due to235U and nuclide-dependent Watt spectra sampling schemes were negligible; however, the use of nuclide-dependent Watt spectra for vessel fluence calculations was found to be important for fuel cycles with burned and fresh fuel.
MODELLING ASTRID-LIKE SODIUM-COOLED FAST REACTOR WITH SERPENT-DYN3D CODE SEQUENCE
Authors: Rydlewicz Wojciech , Fridman Emil , Shwageraus Eugene
Published: February 2021

This study explores the feasibility of applying the Serpent-DYN3D sequence to the analysis of Sodium-cooled Fast Reactors (SFRs) with complex core geometries, such as the ASTRIDlike design. The core is characterised by a highly heterogeneous configuration and was likely to challenge the accuracy of the Serpent-DYN3D sequence. It includes axially heterogeneous fuel assemblies, non-uniform fuel assembly heights and large sodium plena. Consequently, the influence of generation and correction methods of various homogenised, few-group crosssections (XS) on the accuracy of the full-core nodal diffusion DYN3D calculations is presented. An attempt to compare the approximate time effort spent on models preparation against the accuracy of the result is made. Results are compared to reference full-core Serpent MC (Monte Carlo) solutions. Initially, XS data was generated in Serpent using traditional methods (2D single assemblies and 2D super-cells). Full core calculations and MC simulations offered a moderate agreement. Therefore, XS generation with 2D fuel-reflector models and 3D single assembly models was verified. Super-homogenisation (SPH) factors for XS correction were applied. In conclusion, the performed work suggests that Serpent-DYN3D sequence could be used for the analysis of highly heterogeneous SFR designs similar to the studied ASTRID-like, with an only small penalty on the accuracy of the core reactivity and radial power distribution prediction. However, the XS generation route would need to include the correction with SPH factors and generation of XS with various MC models, for different core regions. At a certain point, there are diminishing returns to using more complex XS generation methods, as the accuracy of full-core deterministic calculations improves only slightly, while the time effort required increases significantly.
RESEARCH ON THE CROSS-SECTION GENERATING METHOD IN HTGR SIMULATOR BASED ON MACHINE LEARNING METHODS
Authors: Zeguang LI , Sun Jun , Wei Chunlin , Sui Zhe , Qian Xiaoye
Published: February 2021

With the increasing needs of accurate simulation, the 3-D diffusion reactor physics module has been implemented in HTGR’s engineering simulator to give better neutron dynamics results instead of point kinetics model used in previous nuclear power plant simulators. As the requirement of real-time calculation of nuclear power plant simulator, the cross-sections used in 3-D diffusion module must be calculated very efficiently. Normally, each cross-section in simulator is calculated in the form of polynomial by function of several concerned variables, the expression of which was finalized by multivariate regression from large number scattered database generated by previous calculation. Since the polynomial is explicit and prepared in advance, the cross-sections could be calculated quickly enough in running simulator and achieve acceptable accuracy especially in LWR simulations. However, some of concerned variables in HTGR are in large scope and also the relationships of these variables are non-linear and very complex, it is very hard to use polynomial to meet full range accuracy. In this paper, a cross-section generating method used in HTGR simulator is proposed, which is based on machine learning methods, especially deep neuron network and tree regression methods. This method first uses deep neuron networks to consider the nonlinear relationships between different variables and then uses a tree regression to achieve accurate cross-section results in full range, the parameters of deep neuron networks and tree regression are learned automatically from the scattered database generated by VSOP. With the numerical tests, the proposed cross-section generating method could get more accurate cross-section results and the calculation time is acceptable by the simulator.
DEVELOPMENT OF PIN-LEVEL NEUTRONICS/THERMAL-FLUID ANALYSIS COUPLED CODE SYSTEM FOR A BLOCK-TYPE HTGR CORE
Authors: Seungsu Yuk , Nam-il Tak , Jo Keun Chang
Published: February 2021

Recently, the coupling between computer codes that simulate different physical phenomena has attracted for more accurate analysis. In the case of high-temperature gas-cooled reactor (HTGR), the coupling between neutronics and thermal-fluid analysis is necessary because of large change of temperature in the reactor core. Korea Atomic Energy Research Institute (KAERI) has developed the coupled code system between a reactor physics analysis code CAPP and a thermal-fluid system safety analysis code GAMMA+ for a block-type HTGR. The CAPP/GAMMA+ coupled code system provides more accurate block-wise distribution data than CAPP or GAMMA+ stand-alone analysis. However, the block-wise distribution data has the limitation in order to predict safety parameters such as the maximum temperature of the nuclear fuel. It is necessary to calculate refined distribution, for example, pin-level (fuel compact level) distribution. In this study, we tried to solve this problem by coupling CAPP and a high-fidelity thermal-fluid analysis code CORONA. CORONA can perform a high-fidelity thermal-fluid analysis of Computational Fluid Dynamics (CFD) level by dividing a block-type HTGR core into small lattices. On the other hand, CAPP can provide a pin power distribution. It is expected that the refined, more accurate distribution data for a block-type HTGR can be obtained by coupling these two codes. This paper presents the development of coupled code system between CAPP and CORONA, and then it is tested on a simple HTGR column problem with encouraging results.
RESONANCE SELF-SHIELDING CALCULATION FOR PLATE-TYPE FUEL ASSEMBLIES
Authors: Xia Fan , Wu Hongchun , Li Yunzhao , Yang Jiewei
Published: February 2021

Owing to its ability to handle arbitrary geometry and high computation efficiency, subgroup method is a widely used resonance self-shielding method. Ordinarily, the subgroup parameters are generated from homogenous resonance integral tables, and then can be used to calculate heterogeneous problems via the equivalence theory. However, it cannot provide accurate results especially for the plate-type assemblies with strong heterogeneity. What’s more, the fuel enrichment in plate-type assemblies is relatively much higher than the common rod-type ones. As a result, the resonance interference effect in plate-type fuels is particularly intense. To solve this problem and to provide accurate effective self-shielded cross-sections for plate-type fuel assemblies, 1-D plate problems are used to generate subgroup parameters. In order to deal with resonance interference, the resonance interference factors are generated by equivalent homogenous problems solved by 0-D hyperfine group solver. To avoid the computational burden caused by too many hyperfine group calculations, the importance of resonance isotopes is calculated in advance to select important resonance isotopes. Important and non-important ones are handled by different ways respectively. JRR-3M plate-type assemblies are used to test the newly method. Numerical results show that the relative errors of effective self-shielded cross-sections are generally less than 1.5% compared with the reference.
METHODS FOR CALCULATION OF SELF-SHIELDED CROSS SECTIONS OF FULLY CERAMIC MICRO-ENCAPSULATED FUEL DOUBLE HETEROGENEOUS SYSTEM
Authors: He Qingming , Yin Wen , Liu Zhouyu , Zu Tiejun , Cao Liangzhi , Wu Hongchun
Published: February 2021

Fully ceramic micro-encapsulated (FCM) fuel is a kind of fuel that employs tri-structural isotropic (TRISO) particles to enhance safety. The FCM fuel assembly is a double heterogeneous system. The conventional self-shielding calculation methods cannot treat the DH effect. In this paper, three methods based on equivalent homogenization of the TRISO particle and the matrix are studied and compared: the hyper-fine energy group cross sections (XSs) homogenization based hyper-fine energy group method (HHM), the hyper-fine energy group XSs homogenization based subgroup method (HSM) and the subgroup XSs homogenization based subgroup method (SSM). These methods are implemented in a high-fidelity neutronics code NECP-X. Numerical results show that these methods are able to treat the double heterogeneity of the FCM fuel. The precision of the HHM and HSM is higher than that of the SSM.
ITERATIVE AND PARALLEL PERFORMANCE ANALYSIS OF NON-BLOCKING COMMUNICATION ALGORITHMS IN THE MASSIVELY PARALLEL NEUTRON TRANSPORT CODE PIDOTS
Authors: Yessayan Raffi , Azmy Yousry Y. , Zerr R. Joseph
Published: February 2021

The PIDOTS neutral particle transport code utilizes a red/black implementation of the Parallel Gauss-Seidel algorithm to solve the SN approximation of the neutron transport equation on 3D Cartesian meshes. PIDOTS is designed for execution on massively parallel platforms and is capable of using the full resources of modern, leadership class high performance computers. Initial testing revealed that some configurations of PIDOTS’s Integral Transport Matrix Method solver demonstrated unexpectedly poor parallel scaling. Work at Idaho and Los Alamos National Laboratories then revealed that this inefficiency was a result of the accumulation of high-cost latency events in the complex blocking communication networks employed during each PIDOTS iteration. That work explored the possibility of minimizing those inefficiencies while maintaining a blocking communications model. While significant speedups were obtained, it was shown that fully mitigating the problem on general-purpose platforms was highly unlikely for a blocking code. This work continues that analysis by implementing a deeply interleaved non-blocking communication model into PIDOTS. This new model benefits from the optimization work performed on the blocking model while also providing significant opportunities to overlap the remaining un-mitigated communication costs with computation. Additionally, our new approach is easily transferable to other similarly spatially decomposed codes. The resulting algorithm was tested on LANL’s Trinity system at up to 32,768 processors and was found at that processor count to effectively hide 100% of MPI communication cost – equivalently 20% of the red/black phase time. It is expected that the implemented interleaving algorithm can fully support far higher processor counts and completely hide communication costs up ~50% of total iteration time.
PARAMETRIC STUDY OF PARALLEL BLOCK JACOBI / SOURCE ITERATION HYBRID METHODS IN 2-D CARTESIAN GEOMETRY AND CONSTRUCTION OF THE INTEGRAL TRANSPORT MATRIX METHOD MATRICES VIA GREEN’S FUNCTIONS
Authors: Hoagland Dylan S. , Azmy Yousry Y.
Published: February 2021

Parallel Block Jacobi (PBJ) [1] is an asynchronous spatial domain decomposition with application in solving the neutron transport equation due to its extendibility to massively parallel solution in unstructured spatial meshes (grids) without the use of the computationally complex and expensive sweeps required by the Source Iteration (SI) method in these applications. [2] However, PBJ iterative methods suffer a lack of iterative robustness in problems with optically thin cells, [1] which we have previously demonstrated to be a consequence of PBJ’s asynchronicity. To mitigate this effect, we have developed multiple PBJ / SI hybrid methods which employ a PBJ method (Parallel Block Jacobi - Integral Transport Matrix Method (PBJ-ITMM) or Inexact Parallel Block Jacobi (IPBJ)) along with SI. [3,4] In this work, we perform a parametric study to determine performance of numerous PBJ / SI hybrid methods as a function of multiple problem parameters. This parametric study reached 5 main conclusions: 1) our hybrid approach is more effective with PBJ-ITMM than with IPBJ, 2) for PBJ-ITMM, there is a hybrid method that mitigates the aforementioned iterative slowdown in optically thin cells without diminishing the method’s potential parallelism in unstructured grids, 3) this hybrid method is most effective in problems with large, continuous regions of very thin cells, 4) the best performing hybrid method consistently executes within a factor of ten slower than current state-of-the-art acceleration methods that are not efficiently extendable to the massively parallel regime, and 5) both PBJ-ITMM and IPBJ are observed to be viable approaches for our desired applications. In the pursuit of implementing PBJ-ITMM in unstructured grids, we conclude with a description of the Green’s Function ITMM Construction (GFIC) algorithm, which allows for the ITMM matrices to be constructed using the pre-existing SI sweep algorithm already present in unstructured grid SN transport codes.
STUDY OF THE EIGENVALUE SPECTRA OF THE NEUTRON TRANSPORT PROBLEM IN PN APPROXIMATION
Authors: Saracco P. , Abrate N. , Burrone M. , Dulla S. , Ravetto P.
Published: February 2021

The study of the steady-state solutions of neutron transport equation requires the introduction of appropriate eigenvalues: this can be done in various different ways by changing each of the operators in the transport equation; such modifications can be physically viewed as a variation of the corresponding macroscopic cross sections only, so making the different (generalized) eigenvalue problems non-equivalent. In this paper the eigenvalue problem associated to the time-dependent problem (α eigenvalue), also in the presence of delayed emissions is evaluated. The properties of associated spectra can give different insight into the physics of the problem.
IMPROVING A POSTERIORI SPATIAL DISCRETIZATION ERROR ESTIMATION FOR SN TRANSPORT BY SINGULAR CHARACTERISTICS TRACKING
Authors: Hart Nathan H. , Azmy Yousry Y.
Published: February 2021

Previously, we have developed a novel spatial discretization error estimator, the “residual source” estimator, in which an error transport problem, analogous to the discretized transport equation, is solved to acquire an estimate of the error, with a residual term acting as a fixed source. Like all error estimators, the residual source estimator suffers inaccuracy and imprecision in the proximity of singular characteristics, lines across which the solution is irregular. Estimator performance worsens as the irregularities become more pronounced, especially so if the true solution itself is discontinuous.

This work introduces a modification to the residual approximation procedure that seeks to reduce the adverse effects of the singular characteristics on the error estimate. A partial singular characteristic tracking scheme is implemented to reduce the portion of the error in the numerical solution born by irregularities in the true solution. This treated numerical solution informs the residual approximations.

The partial singular characteristic tracking scheme greatly enhances the numerical solution for a problem with prominent singular characteristics. The residual approximation and resultant residual source error estimate are likewise improved by the scheme, which only incurs the computational cost of an extra inner iteration.
TOWARDS CAD-BASED GEOMETRY MODELLING WITH THE RANDOM RAY METHOD
Authors: Shriwise Patrick C. , Tramm John R. , Davis Andrew , Romano Paul K.
Published: February 2021

The Advanced Random Ray Code (ARRC) is a high performance computing application capable of high-fidelity simulations of full core nuclear reactor models. ARRC leverages a recently developed stochastic method for neutron transport, known as The Random Ray Method (TRRM), which offers a variety of computational and numerical advantages as compared to existing methods. In particular, TRRM has been shown to be capable of efficient simulation of explicit three dimensional geometry representations without assumptions about axial homogeneity. To date, ARRC has utilized Constructive Solid Geometry (CSG) combined with a nested lattice geometry which works well for typical pressurized water reactors, but is not performant for the general case featuring arbitrary geometries.

To facilitate simulation of arbitrarily complex geometries in ARRC efficiently, we propose performing transport directly on Computer-Aided Design (CAD) models of the geometry. In this study, we utilize the Direct-Accelerated Geometry Monte Carlo (DAGMC) toolkit which tracks particles on tessellated CAD geometries using a bounding volume hierarchy to accelerate the process, as a replacement for ARRC’s current lattice-based accelerations. Additionally, we present a method for automatically subdividing the large CAD regions in the DAGMC model into smaller mesh cells required by random ray to achieve high accuracy. We test the new DAGMC geometry implementation in ARRC on several test problems, including a 3D pincells, 3D assemblies, and an axial section of the Advanced Test Reactor. We show that DAGMC allows for simulation of complex geometries in ARRC that would otherwise not be possible using the traditional approach while maintaining solution accuracy.
LOAD BALANCING ANGULAR ADAPTIVITY ON ENERGY DEPENDENT REACTOR PROBLEMS
Authors: Nikiteas Ioannis , Dargaville Steven , Smith Paul N. , Smedley-Stevenson Richard P. , Pain Christopher C.
Published: February 2021

This paper presents a novel parallel algorithm capable of solving problems with multiple energy groups, such as reactor core problems, using energy dependent angular hadaptivity coupled with Dynamic Load Balancing (DLB). We use the C5G7 NEA benchmark as an example with our in-house code FETCH2 and compare runtimes and Continuous Degrees of Freedom (CDOFs) between energy independent angular adapting runs with and without DLB. Our results demonstrate a 32% reduction in CDOFs and a 38% reduction in runtime in favour of the energy dependent angular adaptivity runs with DLB when compared to energy independent angular adaptivity.
VERIFICATION OF THE OpenMC HOMOGENIZED MYRRHA-1.6 CORE MODEL
Authors: Hernandez-Solis Augusto , Molla Yohannes , Malambu Edoaurd , Stankovskiy Alexey , Van den Eynde Gert
Published: February 2021

The OpenMC code is being employed both as a multi-group nodal macroscopic cross-section generator and a reference multi-group Monte Carlo (MGMC) solution. The aim is to do a neutronic benchmark verification study versus a deterministic model (based on the MYRRHA-1.6 core) performed by the PHISICS simulator. MYRRHA, a novel research accelerator driven system concept that is also foreseen to work as a critical configuration, offers a rich opportunity of testing state-of-the art methods for reactor physics analysis due to its strong heterogeneous configuration utilized for both thermal and fast spectra irradiation purposes. The original core configuration representing MYRRHA-1.6 and formed by 169 assemblies, was launched in OpenMC for producing a homogenous nodal model that, when executed in its multi-group Monte Carlo mode, it produced a keff that differs in almost 500 pcm from the original case. This means that in the future, such approximation should correct the nodal cross-sections to preserve the reaction rates in order to match those ones from the heterogeneous model. Nevertheless, such MGMC mode of operation offered by OpenMC could be exploited in order to verify deterministic core simulators. By inputting the same nodal multi-group cross-section model into the transport solver of the PHISICS toolkit, the neutronic benchmark showed a difference of 171 pcm in eigenvalue while comparing it to its OpenMC MGMC counterpart. Also, local multi-group and energy-integrated nodal profiles of the neutron flux showed a maximum relative difference between methodologies of 15% and 1%, respectively. This means that the MGMC capabilities offered by OpenMC can be employed to verify other deterministic methodologies.
OPTIMISATION OF MONTE CARLO BURNUP SIMULATIONS
Authors: Dufek Jan , Mickus Ignas
Published: February 2021

We show here that computing efficiency of Monte Carlo burnup simulations depends on chosen values of certain free parameters, such as the length of the time steps and the number of neutron histories simulated at each Monte Carlo criticality run. The efficiency can thus be improved by optimising these parameters. We have set up a simple numerical model that made it possible for us to test a large number of combinations of the free parameters, and suggest a way to optimise their selection.
INVESTIGATION ON DETERMINISTIC TRUNCATION TO CONTINUOUS ENERGY MONTE CARLO NEUTRON TRANSPORT CALCULATION
Authors: Kim Inhyung , Kim Yonghee
Published: February 2021

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.
BEAVRS BENCHMARK EVALUATION WITH CASMO5 AND SIMULATE5
Authors: Bahadir Tamer
Published: February 2021

The MIT BEAVRS benchmark problem, which was primarily setup for the verification and validation of high-fidelity tools that have coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion models, has also found extensive usage in the reactor physics community for validating core analysis tools. The primary purpose of this paper is to provide an accurate, comprehensive evaluation of the BEAVRS benchmark with CASMO5 and SIMULATE5 codes. The CMS5 calculated results for low-power physics tests (hot zero power critical boron, control rod worth and isothermal temperature coefficients) and full power operation (boron let-down and flux map reaction rate distributions) are compared to plant measured data provided in the benchmark specification. The CMS5 model, using ENDF/BVII.1 nuclear data library, predicts HZP critical boron concentration for all-rods-out conditions within 10 ppm for Cycle-1, and 25 ppm in Cycle-2; the control rod worth is predicted with a difference of 0.7% ± 3.8%, where the maximum difference is less than 10%. For the core follow calculations at the hot full power condition, the average difference in predicting the critical boron concentration is less than 20 ppm. In addition, the radial and nodal reaction rate distributions are predicted with a mean difference of about 1.6% and 3.8%, respectively. The CMS5 calculations are repeated using the most recent ENDF/B-VIII.0 library. No significant difference is observed in predicting measured plant parameters with different nuclear data libraries. Additionally, the impact of various modeling options, which are typically employed with nodal diffusion codes, on the predictions of important core parameters are presented as part of the benchmark evaluation.
QUANTUM ANNEALING OPTIMIZATION OF A HEURISTIC SURROGATE MODEL FOR PWR FUEL LOADING
Authors: Whyte Andy , Parks Geoff
Published: February 2021

An efficient fuel arrangement must be generated by PWR operators every 6–18 months. This complex problem has been extensively researched with two broad approaches, heuristic and stochastic methods, becoming accepted. This initial study qualitatively introduces the concept of encoding full-core PWR fuel loading patterns in a form suitable for quantum annealing. The concepts of adiabatic quantum computers and quantum annealing are introduced, and a surrogate model encoding of a set of heuristics for loading pattern design produced in a form suitable for use in present-day quantum annealers. The simulated results show significant similarity to benchmark loading patterns.
FEW-GROUP CROSS SECTIONS MODELING BY ARTIFICIAL NEURAL NETWORKS
Authors: Szames E. , Ammar K. , Tomatis D. , Martinez J.M.
Published: February 2021

This work deals with the modeling of homogenized few-group cross sections by Artificial Neural Networks (ANN). A comprehensive sensitivity study on data normalization, network architectures and training hyper-parameters specifically for Deep and Shallow Feed Forward ANN is presented. The optimal models in terms of reduction in the library size and training time are compared to multi-linear interpolation on a Cartesian grid. The use case is provided by the OECD-NEA Burn-up Credit Criticality Benchmark [1]. The Pytorch [2] machine learning framework is used.
GEN-FOAM MODEL AND BENCHMARK OF DELAYED NEUTRON PRECURSOR DRIFT IN THE MOLTEN SALT REACTOR EXPERIMENT
Authors: Shi Jun , Fratoni Massimiliano
Published: February 2021

The effective delayed neutron fraction is an important reactor kinetics parameter. In flowing liquid-fuel reactors, this differs from the delayed neutron fraction because of the emission of delayed neutrons with a lower energy spectrum than prompt and the delayed neutron precursor (DNP) drift due to the fuel movement. In general, neglecting delayed neutron precursor drift leads to an over-estimation of the effective delayed neutron fraction. Nevertheless, the capability to simulate this peculiar phenomenon is not available in most reactor physics tools. In this project, a multi-physics approach to modeling DNP drift is developed using the GeN-Foam toolkit, and it benchmarked against available experimental data from the Molten Salt Reactor Experiment (MSRE). GeN-Foam couples a neutron diffusion solver with a thermal-hydraulics solver. Additionally, a new function was added for solving adjoint multi-group diffusion eigenvalue problems and calculating effective delayed neutron fraction. For benchmarking, an R-Z model of the MSRE was developed in GeN-Foam. The porous media model was applied, and cross sections were generated using the Monte Carlo code Serpent-2 with ENDF/B-VII.1 nuclear data library. In order to evaluate the impact of DNP drift, two steady-state conditions (stationary and flowing salt at 1200 gpm) were simulated. A reactivity change of -241 pcm was calculated using GeN-Foam for the MSRE between static and flowing fuel, which is in a good agreement with the experimental value of -212 pcm. The total effective delayed neutron fraction change was calculated to be -230 pcm vs. -304 pcm reported for the MSRE and analytical calculated during the experimental campaign. Three transient accidents were also analyzed.
NEUTRON DEPTH PROFILE CALCULATIONS USING ARTIFICIAL NEURAL NETWORKS
Authors: Hossny K. , Magdi S. , Nasr F. , Yasser Y. , Magdy A.
Published: February 2021

Neutron depth profiling (NDP) is a non-destructive technique used for identifying the concentration of impurity isotopes below the sample surface. NDP is carried out by detection of the emitted charged particles resulting from bombarding the sample with neutrons. NDP specifies the isotopic concentration versus the sample depth for a few micrometers below the surface. The sample is bombarded inside a research reactor using a thermal neutron beam. Charged particles like alpha particles or protons are produced from the neutron induced reactions in the sample. Each neutron isotopic interaction produces a certain Q, indicating a specific kinetic energy for the emitted charged particle. As the charged particle travels through the sample to eject the surface, it loses energy to atoms (electrons) on its path. The charged particle energy loss holds information regarding the number of atoms by which the emitted particle passed, thus indicating its original depth. The purpose of this work is to check the capability of Artificial Neural Networks (ANNs) in predicting the boron concentration profile across a boro-silicate sample of thickness 3.5 μm divided into 10 layers. Each layer included different boron concentration than the other. Also, the boron concentration had the values {0.1, 0.2, 0.3, 0.4, 0.5, 0.6, 0.7, 0.8, 0.9, 1}. Training, validation, and test data were generated synthetically using MCNP6 in which the boron concentrations varied in the layer number from one sample to another. MCNP6 model consisted of a silicon barrier detector, boro-silicate sample, chamber body and an idealized thermal neutron source. The detector, sample, and the source were located in a voided chamber. The samples were irradiated with a 0.025 eV monoenergetic thermal neutron beam from a monodirectional disk source. To cover the whole area of the samples, the thermal neutron beam had a radius of 3 cm. The silicon detector active volume was modelled as a 100 μm thick and 3 cm radius facing the sample directly. The sample, beam, and the detector were placed on the same axis. Ten ANN regression models were developed, one for each layer boron concentration prediction where the input for each model was the alpha spectrum read by the detector, while the output was the boron concentration for each layer. Results showed regression values higher than 0.94 for all of the developed models. ANNs proved its capability of predicting the boron profile form the alpha spectrum read by the detector regarding neutron depth profiling in a boro-silicate samples.
DEVELOPMENT OF A COUPLED SUBPLANE CAPABILITY IN MPACT
Authors: Stimpson Shane , Graham Aaron , Collins Benjamin
Published: February 2021

Recent efforts in the development of the deterministic transport code MPACT have been devoted to preparing the 2D/1D subplane method to be a production-level capability, as well as leveraging a multilevel coarse mesh finite difference (CMFD) approach to substantially reduce the runtime of target problems. For example, as compared to the previous default 2D/1D solver in MPACT on a standard quarter core model, the new solver reduces in core-hour requirements by ~5–6×.

Previous work focused solely on cases without multiphysics feedback, which is obviously important for analyzing the more realistic problems of operating reactors. The work presented in this article focuses on efforts to incorporate thermal hydraulics (TH) coupling through CTF by leveraging what are termed as subgrid solvers, which effectively treat material heterogeneities within subplane regions. Previous efforts have targeted using subgrid solvers for control rods and spacer grids; in this work, they are applied to account for the material property heterogeneities with regards to temperature/density distributions. This will allow the fidelity of coupling to be maintained while still reaping the performance benefits.

These new developments are demonstrated on two problems: (1) a single assembly case with feedback, known as Progression Problem 6a, and (2) a 3×3 cluster of assemblies with feedback based on Progression Problem 4a. The results demonstrate notable performance improvement potential for cases with TH feedback, but this approach is more dependent on the iteration process.
REDUCED-ORDER MODELLING OF PARAMETERIZED TRANSIENT FLOWS IN CLOSED-LOOP SYSTEMS
Authors: German Péter , Tano Mauricio , Ragusa Jean C. , Fiorina Carlo
Published: February 2021

In this paper, two Galerkin projection based reduced basis approaches are investigated for the reduced-order modeling of parameterized incompressible Navier-Stokes equations for laminar transient flows. The first approach solves only the reduced momentum equation with additional, physics-based approximations for the dynamics of the pressure field. On the other hand, the second approach solves both the reduced momentum and continuity equations. The reduced bases for the velocity and pressure fields are generated using the method of snapshots combined with Proper Orthogonal Decomposition (POD) for data compression. To remedy the stability issues of the two-equation model, the reduced basis of the velocity is enriched with supremizer functions. Both reduced-order modeling approaches have been implemented in GeN-Foam, an OpenFOAM-based multi-physics solver. A numerical example is presented using a two-dimensional axisymmetric model of the Molten Salt Fast Reactor (MSFR) and the dynamic viscosity as the uncertain parameter.

The results indicate that the one-equation model is slightly more accurate in terms of velocity, while the two-equation model, built with the same amount of modes for the velocity, is far more accurate in terms of pressure. The speed-up factors for the reduced-order models are 3060 for the one-equation model and 2410 for the two-equation model.
ANTARES: COUPLING PARCS WITH CATHARE-3
Authors: Forestier M. , Girault G. , Jacq F. , Sargeni A.
Published: February 2021

In recent years, the IRSN has launched a new project to couple the first 3D version of the thermal hydraulic code CATHARE-3 (system) with the 3D, neutronic nodal code PARCS (core): ANTARES (Advanced Neutronics and Thermal-hydraulic for the Analysis of the Reactor Safety). The purpose of this project is to increase the IRSN capability to couple different codes, to calculate the core power distribution in CATHARE-3 and to improve the thermal hydraulic boundaries conditions in PARCS. In this way, the IRSN diversifies its available tools to perform safety analysis with improved accuracy. The current technique usually adopted in France for the safety demonstrations is the so-called ‘conservative' approach, which consists of reducing all the feedback (Doppler and moderator effects) and in modifying some physical quantities in such a way to increase a power peak in an accidental transient. For this reason, these facilities (‘penalties’) have been implemented in ANTARES. In this paper we will give two examples of accidental transients that can be simulated with ANTARES: a REA (Rod Ejection Accident) and an inadvertent boron dilution event.
IMPLEMENTATION OF AN AUTONOMOUS REACTIVITY CONTROL SYSTEM IN A SMALL LEAD-COOLED FAST REACTOR
Authors: Dehlin Fredrik , Acharya Govatsa , Bortot Sara , Mickus Ignas
Published: February 2021

This paper describes the design, implementation and characterisation of an Autonomous Reactivity Control (ARC) system in a small modular lead-cooled fast reactor. The aim of this work was to demonstrate the applicability of the ARC system and to study its dynamic behaviour during an anticipated transient without scram. A simplified one-dimensional model was developed to calculate the heat transfer within the ARC system, and the reactivity worth as a function of the neutron poison’s insertion into the active core was obtained via static neutronic calculations. By coupling the aforementioned models, the ARC’s time-dependent reactivity was derived as a function of the coolant outlet temperature variation. This model was implemented into the BELLA multi-point dynamics code and transient simulations were run. A control rod ejection accident was studied leading to an unprotected transient overpower scenario, in which 350 pcm reactivity was inserted during one second. It was shown that the ARC system provides a forceful negative reactivity feedback and that steady-state temperatures after the transient were reduced by almost 300 K compared to an identical transient without its action. In this scenario, the ARC system managed to stabilise the coolant outlet temperature at a value 100 K above nominal conditions. The implementation of an ARC system provided the reactor with a passively actuated self-regulating reactivity control system able to insert large amounts of negative reactivity in a short amount of time.
INVESTIGATION OF A SELF-ACTUATED, GRAVITY-DRIVEN SHUTDOWN SYSTEM IN A SMALL LEAD-COOLED REACTOR
Authors: Acharya Govatsa , Dehlin Fredrik , Bortot Sara , Mickus Ignas
Published: February 2021

Passive safety systems in a nuclear reactor allow to simplify the overall plant design, beside improving economics and reliability, which are considered to be among the salient goals of advanced Generation IV reactors. This work focuses on investigating the application of a self-actuated, gravity-driven shutdown system in a small lead-cooled fast reactor and its dynamic response to an initiating event. The reactor thermal-hydraulics and neutronics assessment were performed in advance. According to a first-order approximation approach, the passive insertion of shutdown assembly was assumed to be influenced primarily by three forces: gravitational, buoyancy and fluid drag. A system of kinematic equations were formulated a priori and a MATLAB program was developed to determine the dynamics of the assembly. Identifying the delicate nature of the balance of forces, sensitivity analysis for coolant channel velocities and assembly foot densities yielded an optimal system model that resulted in successful passive shutdown. Transient safety studies, using the multi-point dynamics code BELLA, showed that the gravity-driven system acts remarkably well, even when accounting for a brief delay in self-actuation. Ultimately the reactor is brought to a sub-critical state while respecting technological constraints.
VALIDATION OF COOLANT THERMAL RESPONSE IN A TRANSIENT FINITE DIFFERENCE THERMAL TRANSPORT MODEL WITH APPLICATIONS TO FAST SPECTRUM REACTORS
Authors: Holdampf Sydney A. , Osborne Andrew G. , Deinert Mark R.
Published: February 2021

Transient behavior in nuclear reactors is important in accidents and with reactivity control systems that are driven by thermal feedback. Here, we describe a transient finite difference model for a pin cell system. The fidelity of the model is shown by validation against the thermocouple measurements of the CABRI BI1 experiment and the Safety Analysis System-Sodium Fast Reactor model of the experiment. In the BI1 experiment, a sodium-cooled mixed oxide fuel pin was subject to a loss of flow transient to coolant boiling within a sodium test loop positioned in the center of the CABRI research reactor. Comparisons to the initial steady-state coolant temperature profile, coolant temperature profile at twenty seconds into the transient, and at four axial locations within the coolant show agreement of the simple model with the experimental results better than or similar to those of the Safety Analysis System-Sodium Fast Reactor model. The model can be used to determine the thermal response times of coolant in fast reactors currently operating or in the design phase when subject to loss of flow accidents or other transients. Here, we investigate the difference in coolant thermal response for metal fueled and mixed oxide fueled sodium fast reactors when subject to transient overpower and loss of flow events. Additionally, we determine the effect of pin pitch on outlet coolant temperatures during the overpower event. Finally, we return to the CABRI experiment and show the importance of porosity in fuel temperature calculations.