bluuebunny/crossref_metadata_2025 huggingface dataset:
A ROBUST SECOND-ORDER MULTIPLE BALANCE METHOD FOR TIME-DEPENDENT NEUTRON TRANSPORT SIMULATIONS
Authors: Variansyah Ilham , Larsen Edward W. , Martin William R.
Published: February 2021

A second-order “Time-Dependent Multiple Balance” (TDMB) method for solving neutron transport problems is introduced and investigated. TDMB consists of solving two coupled equations: (i) the original balance equation (the transport equation integrated over a time step) and (ii) the “balance-like” auxiliary equation (an approximate neutron balance equation). Simple analysis shows that TDMB is second-order accurate and robust (unconditionally free from spurious oscillation). A source iteration (SI) method with diffusion synthetic acceleration (DSA) is formulated to solve these equations. A Fourier analysis reveals that the convergence rates of the proposed iteration schemes for TDMB are similar to those of the common (SI + DSA) schemes for Backward Euler (BE); however, TDMB requires about twice the computational effort per iteration. To demonstrate the theory—accuracy, robustness, and convergence rate—and investigate the efficiency of TDMB, we present results from a discrete ordinates (Sn) research code. Results are discussed, and future work is proposed.
USING ARTIFICIAL NEURAL NETWORKS TO ACCELERATE TRANSPORT SOLVES
Authors: Tano Mauricio E. , Ragusa Jean C.
Published: February 2021

Discontinuous Finite Element Methods (DFEM) have been widely used for solving SN radiation transport problems in participative and non-participative media. In this method, small matrix-vector systems are assembled and solved for each cell, angle, energy group, and time step while sweeping through the computational mesh. In practice, these systems are generally solved directly using Gaussian elimination, as computational acceleration for solving this small systems are often inadequate. Nonetheless, the computational cost of assembling and solving these local systems, repeated for each cell in the phase-space, can amount to a large fraction of the total computing time. In this paper, a Machine Learning algorithm is designed to accelerate the solution of local systems. This one is based on Artificial Neural Networks (ANNs). Its key idea is training an ANN with a large set of solutions to random one-cell transport problems and, then, replacing the assembling and solution of the local systems by the feedforward evaluation of the trained ANN. It is observed that the optimized ANNs are able to reduce the compute times by a factor of ~ 3:6 per source iteration, while introducing mean absolute errors between 0:5 – 2% in transport solutions.
SUPER-HISTORY METHODS FOR ADJOINT-WEIGHTED TALLIES IN MONTE CARLO TIME EIGENVALUE CALCULATIONS
Authors: Filiciotto F. , Jinaphanh A. , Zoia A.
Published: February 2021

Time eigenvalues emerge in several key applications related to neutron transport problems, including reactor start-up and reactivity measurements. In this context, experimental validation and uncertainty quantification would demand to assess the variation of the dominant time eigenvalue in response to a variation of nuclear data. Recently, we proposed the use of a Generalized Iterated Fission Probability method (G-IFP) to compute adjoint-weighted tallies, such as kinetic parameters, perturbations and sensitivity coefficients, for Monte Carlo time (or alpha) eigenvalue calculations. With the massive use of parallel Monte Carlo calculations, it would be therefore useful to trade the memory burden of the G-IFP method (which is comparable to that of the standard IFP method for k-eigenvalue problems) for computation time and to rely on history-based schemes for such adjoint-weighted tallies. For this purpose, we investigate the use of the super-history method as applied to estimating adjoint-weighted tallies within the α-k power iteration, based on previous work on k-eigenvalue problems. Verification of the algorithms is performed on some simple preliminary tests where analytic solutions exist. In addition, the performances of the proposed method are assessed by comparing the super-history and the G-IFP methods for the same sets of benchmark problems.
AN IMPROVED DISTINCT ELEMENT METHOD FOR HIGH PACKING FRACTION STOCHASTIC MEDIA MODELING
Authors: Feng Zhiyuan , An Nan , Wang Kan
Published: February 2021

Due to the generality and flexibility of Monte Carlo method in geometric modeling, Monte Carlo method plays an important role in accurate simulation of random media. At present, rand om sequential addition method (RSA) and distinct element method (DEM) are more accurate and mature explicit modeling methods. The former approach has the problem of upper limit of packing fraction, which is suitable for stochastic geometry with lower filling rate. DEM method can fill random medium model with packing fraction higher than 60%, but DEM is not suitable for non-contact dispersed particles based on the interaction between particles. There fore, an improved DEM method is proposed to solve the problem of modeling non-contact p articles dispersed in the stochastic media with high packing fraction. The virtual surfaces are constructed outside of the outer layer of particles to make them in contact with each other. Thus, the particle system is suitable for DEM method. The construction of virtual surface does not affect the neutron transport process. The correctness of the improved DEM is verified by comparing the total filling particle number and calculation results ofkeffwith RSA method. At the same time, according to the distribution of filling particles, the improved DEM method fills the particles more uniformly.
THE McSAFE PROJECT - HIGH-PERFORMANCE MONTE CARLO BASED METHODS FOR SAFETY DEMONSTRATION: FROM PROOF OF CONCEPT TO INDUSTRY APPLICATIONS
Authors: Sanchez-Espinoza V. H. , Mercatali L. , Leppänen J. , Hoogenboom E. , Vocka R. , Dufek J.
Published: February 2021

The increasing use of Monte Carlo methods for core analysis is fostered by the huge and cheap computer power available nowadays e.g. in large HPC. Apart from the classical criticality calculations, the application of Monte Carlo methods for depletion analysis and cross section generation for diffusion and transport core simulators is also expanding. In addition, the development of multi-physics codes by coupling Monte Carlo solvers with thermal hydraulic codes (CFD, subchannel and system thermal hydraulics) to perform full core static core analysis at fuel assembly or pin level has progressed in the last decades. Finally, the extensions of the Monte Carlo codes to describe the behavior of prompt and delay neutrons, control rod movements, etc. has been started some years ago. Recent coupling of dynamic versions of Monte Carlo codes with subchannel codes make possible the analysis of transient e.g. rod ejection accidents and it paves the way for the simulation of any kind of design basis accidents as an alternative option to the use of diffusion and transport based deterministic solvers. The H2020 McSAFE Project is focused on the improvement of methods for depletion considering thermal hydraulic feedbacks, extension of the coupled neutronic/thermal hydraulic codes by the incorporation of a fuel performance solver, the development of dynamic Monte Carlo codes and the development of methods to handle large depletion problems and to reduce the statistical uncertainty. The validation of the multi-physics tools developed within McSAFE will be performed using plant data and unique tests e.g. the SPERT III E REA test. This paper will describe the main developments, solution approaches, and selected results.
CONTRIBUTION OF THERMAL-HYDRAULICS SIMULATION TO CRITICALITY ANALYSIS OF A DEBRIS BED NUMERICAL MOCK-UP
Authors: Boulard P. , Jaboulay J-C. , Martinez J-M. , Zoia A.
Published: February 2021

In this work, we evaluate the impact of simulating the thermal-hydraulics feedback in criticality risk assessment in a benchmark configuration corresponding to a simplified debris bed. We have recently discussed the coupling of the multi-phase thermal-hydraulics code MC3D with a reactivity evaluation scheme based on multi-point kinetics, which paves the way towards the assessment of the system behavior in terms of energy release. In this work, we analyze the system reactivity as a function of the thermal-hydraulics state of the system. For our investigation, we have considered a simplified cylindrical region and assumed that the fuel particles in the mixture consist of spheres of small diameter. The power has been adjusted to be representative of nuclear decay heat. For most of the examined configurations, the maximum reactivity lies below zero, which shows that the occurrence of a criticality event is likely to be excluded, thanks to the contribution of the decay heat of the fuel.
SERPENT2-SUBCHANFLOW-TRANSURANUS PIN-BY-PIN DEPLETION CALCULATIONS FOR A PWR FUEL ASSEMBLY
Authors: García Manuel , Tuominen Riku , Gommlich Andre , Ferraro Diego , Valtavirta Ville , Imke Uwe , Van Uffelen Paul , Mercatali Luigi , Sanchez-Espinoza Victor , Leppänen Jaakko , Kliem Sören
Published: February 2021

This work presents the results for a coupled neutronic-thermalhydraulic-thermomechanic pin-level depletion calculation of a PWR fuel assembly using Serpent2-SUBCHANFLOW-TRANSURANUS. This tool is based on a semi-implicit depletion scheme with pin-by-pin feedback, mesh-based field exchange and an object-oriented software design. The impact of including fuel-performance capabilities is analyzed, with focus on high-burnup effects. The treatment of the Doppler feedback to the neutronics is examined as well, in particular the use of radial fuel-temperature profiles or radially averaged values.
HEXAGONAL PWR-CORE MODELING AND SIMULATION WITH APPLICATION OF NECP-BAMBOO
Authors: Zhang Cheng , Cao Liangzhi , Li Yunzhao , Hua Guowei
Published: February 2021

In this paper, the modeling and simulation of the PWRs loaded with hexagonal fuel assemblies has been implemented with the NECP-Bamboo code. NECP-Bamboo, consisting of a 2D lattice code named Bamboo-Lattice and a 3D steady-state core code named Bamboo-Core, was primitively designed for the PWRs loaded with the rectangular fuel assemblies. As the capability extension for PWRs with hexagonal fuel assemblies, four aspects of improvement have been implemented in NECP-Bamboo. Firstly, the Constructive Solid Geometry (CSG) has been implemented in Bamboo-Lattice for the lattice modeling. Secondly, the explicit modeling of the reflector assembly has been applied to provide more reliable few-group constants, compared with the conventional 1D model for the reflector assembly. Thirdly, the assembly-homogenization capability has been extended to the hexagonal assembly. Fourthly, the diffusion solver in Bamboo-Core based on the Variational Nodal Method (VNM) has been extended to handle hexagonal geometry. With application of the capability-extended NECP-Bamboo, the modeling and simulations for the VVER-1000 benchmark loaded with MOX fuel has been implemented. It can be observed that the numerical results provided by NECP-Bamboo can agree well with corresponding results by the Monte-Carlo code.
HIGH-FIDELITY EX-CORE CAPABILITIES IN VERA
Authors: Pandya Tara , Royston Katherine , Davidson Eva , Evans Tom , Godfrey Andrew , Henderson Shane , Gentry Cole , Stimpson Shane , Collins Benjamin
Published: February 2021

High-fidelity nuclear reactor calculations have become increasingly important when considering lifetime extensions of the current nuclear fleet. The need for performing large calculations has led to development of advanced, novel methods for faster and more efficient computing system use. Under the Consortium for Advanced Simulation of Light Water Reactors (CASL), we have enabled the capability to perform high-fidelity ex-core calculations in the Virtual Environment for Reactor Applications (VERA) by coupling with the Shift Monte Carlo (MC) radiation transport package. The codes are coupled inmemory, allowing for the pin-by-pin fission source from the core simulator to be used by Shift. This unique capability allows the user to obtain both in-core and ex-core quantities by running a single simulation. Examples of ex-core calculations that can be performed include multicycle vessel fluence, detector response during reactor start-up and operation, and coupon fluence. One main advantage of the VERA ex-core capability is its flexibility and ease of use; users can select default settings with the standard VERA input for typical calculations or create their own ex-core geometry for specific cases. Also, Shift takes advantage of hybrid deterministic-MC methods to reduce variance and computational time. This paper details the full suite of VERA ex-core capabilities and provides input examples, simulation results, and computing resource use suggestions. These new capabilities have the potential to impact a wide user group in the nuclear community by enhancing and enabling high-fidelity light water reactor (LWR) ex-core calculations.
MULTIPHYSICS MODELING OF PRECURSORS IN MOLTEN SALT FAST REACTORS USING PROTEUS AND Nek5000
Authors: Feng B. , Fei T. , Shaver D. , Jung Y. , Fang J. , Rahaman R. , Lee C. , Smith M. , Shemon E.
Published: February 2021

The goal of this work was to calculate the impact of the delayed neutron precursor drift in fast spectrum Molten Salt Reactors (MSRs) using coupled solutions from the neutronics code PROTEUS and the computational fluid dynamics code Nek5000. Specifically, using a multiphysics approach to solve the effective delayed neutron fraction (βeff) or delayed neutron precursor distribution for reactors with flowing fuel salts would provide valuable information for transient simulations and safety assessments. Given the multiple options for the flux solution and geometric resolution/fidelity in PROTEUS, two approaches were developed and applied to various test cases: PROTEUS-NODAL/Nek5000 and PROTEUS-SN/Nek5000. For the former, the precursors are tracked in the built-in precursor drift model in PROTEUS-NODAL, whereas in the latter, Nek5000 directly tracks the precursors. Both approaches were used to solve a single test channel problem and showed excellent agreement in the calculated βeff. Separately, a 3D hourglass-shaped core was modeled using the PROTEUS-SN/Nek5000 approach. This problem was designed to demonstrate the capability of the discrete ordinates (SN) solver and Nek5000 to model complex core designs with axially varying geometries and the ability for Nek5000 to track the precursors and calculate the resulting βeff. In addition, the Nek5000 calculations revealed the presence of recirculation zones in the hourglass design, which could lead to significant temperatures in the fuel salt and surrounding materials. These first coupled solutions show why these approaches may be necessary for not only predicting the precursor drift effect in fast MSRs but also for reactor design and performance assessments.
HIGH-FIDELITY MULTI-PHYSICS PIN-BY-PIN MODEL OF A SVEA-96 OPTIMA2 ASSEMBLY WITH TORT-TD/CTF
Authors: Seubert A. , Périn Y. , Henry R.
Published: February 2021

This paper describes the development of a 3-d high-fidelity multi-physics pin-by-pin neu-tronic and sub-channel thermal hydraulic model of a SVEA-96 Optima2 fuel assembly for the coupled code system TORT-TD/CTF. No pin power reconstruction is applied. Detailed descriptions are given on (a) the generation of few-group parameterized pin cell-homogenized cross section data using HELIOS, (b) the 3-d neutronic pin-by-pin and ther-mal-hydraulic sub-channel model of a single SVEA-96 Optima2 assembly and its qualifica-tion by Monte Carlo models using Serpent, (c) coupled steady-state results for a single as-sembly and (d) coupled steady-state and transient simulations of a 4-by-4 mini core model of SVEA-96 Optima2 assemblies in a partially rodded state.
EFFICIENT MULTIPHYSICS ITERATIONS IN MPACT WITH PARTIALLY CONVERGENT CMFD
Authors: Shen Qicang , Kochunas Brendan , Downar Thomas
Published: February 2021

Partial convergence of CMFD can help to stabilize multiphysics iteration schemes. In this paper, an efficient multiphysics iteration scheme with near-optimal partially convergent CMFD implemented in MPACT is presented. In the new scheme, the feedback intensity of the problem is automatically estimated, and the relative convergence of CMFD solver is adjusted accordingly. Numerical results show that MPACT with near-optimal partially convergent CMFD can have almost the same convergence rate in problems with feedback as those without feedback. For the problems tested here the run time may be reduced by more than 20% and up to 49% compared with that of current MPACT.
DESIGN AND INVESTIGATION OF WEIGHT BUNDLE SIMULATOR FOR INDIAN PHWR USING APDL - A THERMAL ASPECT
Authors: Bhadauria Madhuri , Kumar Ravi , Das Arup K.
Published: February 2021

Under a postulated scenario of Loss of Coolant Accident (LOCA) with un-availability of emergency core cooling system (ECCS) for Indian Pressurized Heavy Water Reactor (IPHWR), the channel integrity needs to be assured. An experimental facility is presently being developed at Indian Institute of Technology Roorkee (IIT R) India to study such a severe event. In this study a CFD simulation of fuel bundle weight simulator is being carried out using ANSYS 19.0 in transient thermal analysis using ANSYS Parametric Design Language (APDL) solver. The test facility aims to estimate the thermal aspect of weight bundle simulator which is the source of heat generation for pressure tube. Thermo-mechanical deformation of pressure tube will depend on the heat source, therefore it is of great importance to check the thermal integrity of weight bundle fuel simulator. Power requirement, weight, and dimensions of weight bundle fuel simulator is similar to the actual fuel bundle used in 700 MWe IPHWR. Simulation was done for 3 % and 2% decay heat. From analysis it was found that the rate of temperature rise in pressure tube with 3 % decay heat reached a maximum of 2 0C/sec and average rate of temperature rise was around 1 0C/sec whereas with 2% decay heat, pressure tube maximum temperature rise was 1.5 0C/sec and average rate 0.8 0C/sec. Results obtained will be further used for designing of 700 MWe full length channel set-up.
ANALYSIS OF THE SUPERPHENIX START-UP TESTS WITH APOLLO-3: FROM ZERO POWER ISOTHERMAL CONDITIONS TO DYNAMIC POWER TRANSIENT ANALYSIS
Authors: Pierre Sciora , Elias Garcia , Gerald Rimpault , Jean-Baptiste Droin , Vincent Pascal
Published: February 2021

Sodium-cooled Fast Reactors (SFRs) remain a potential candidate to meet future energy needs. In addition, the SFRs experimental feedback is considerable, for instance, the French research program has considered experimental facilities including the Superphénix which has emerged as a transition to commercial deployment. In this paper a set of tests from the Superphénix start-up are reanalyzed with new tools, considering APOLLO-3 and TRIPOLI-4 (respectively deterministic and stochastic codes) for neutron physics evaluation, GERMINAL-V2 for the fuel irradiation behavior and CATHARE-3 for the thermal-hydraulics modelling. Neutron physics evaluations are performed for the main control rod worth and the Doppler Effect, both measured under isothermal conditions at Superphénix start-up. A good agreement is obtained for these tests, which were purely neutronic tests. Next, the core temperature distribution is evaluated at nominal conditions, where larger discrepancies are observed. However, these deviations are related to the measurement of the fuel assemblies, which have a larger than expected uncertainty. Finally a transient, consisting of a negative reactivity insertion, is analyzed to assess the dynamic core behavior. A good agreement is obtained during the reactivity insertion, however the thermal-hydraulic model has to be improved, namely the vessel model, which is considered as a 0-D volume.
SCIENTIFIC WORKFLOWS FOR MCNP6 AND PROTEUS WITHIN THE NEAMS WORKBENCH
Authors: Dominesey Kurt A. , Kowal Peter J. , Eugenio Jonathan A. , Ji Wei
Published: February 2021

In this work we integrate MCNP and PROTEUS, two high-fidelity reactor physics codes, into the Nuclear Energy Advanced Modeling and Simulation (NEAMS)Workbench. Specifically, we enhance the user interface of each code by developing full-featured editor services. Next, we re-define MCNP textual outputs in HDF5 formats, at great convenience to the user. Further, these HDF5 outputs are (optionally) coerced to the OpenMC format, enabling powerful and common post-processing workflows. For compatible use of MCNP and PROTEUS, we develop a model unification workflow, converting PROTEUS to MCNP-compatible meshes and generating a corresponding template MCNP deck. Finally, we investigate cooperative use of MCNP and PROTEUS for hybrid Monte Carlo and deterministic variance reduction through Consistent Adjoint-Driven Importance Sampling (CADIS), demonstrated with a fixed-source shielding problem. Ultimately, we find our integration effort dramatically streamlines pre- and post-processing with MCNP and PROTEUS and enables transformative hybrid Monte Carlo and deterministic workflows.
THE APPLICABILITY RANGE OF THE MODIFIED SOURCE MULTIPLICATION (MSM) METHOD TESTED IN THE FAST VENUS-F REACTOR
Authors: Krása A , Kochetkov A , Messaoudi N , Vittiglio G , Wagemans J
Published: February 2021

The MSM method is an experimental technique for determination of reactivity of a sub-critical reactor. It consists of one dynamic measurement followed by two static measurements, which use an extraneous neutron source. For the data analysis, the core averaged kinetic parameters need to be calculated as well as a spatially-dependent correction factor that corrects for the point kinetics approximation. In order to test the range for which the method is valid and to demonstrate the reliability of the correction factor calculations in a fast reactor, a dedicated experimental campaign was performed in the fast lead-bismuth VENUS-F reactor. The reactivity of a dozen of sub-critical configurations was measured with the MSM method using ten 235U fission chambers. The detectors were located at various distances from the active zone and from the extraneous neutron source, leading to a large range of values of a correction factor (calculated with the Monte Carlo MCNP5 code) used in the data analysis.
BURNUP CALCULATIONS OF THE JSI TRIGA REACTOR FUEL AND COMPARISON WITH MEASUREMENTS
Authors: Anže Pungerčič , Dušan Čalič , Snoj Luka
Published: February 2021

Fuel burnup of the JSI TRIGA was calculated by simulating complete operational history consisting of 240 different core configurations from 1966 to 2020. At the moment we are unable to perform burnup measurements, e.g. gamma spectroscopy on burned fuel elements, hence we used weekly measured excess reactivity as a reference point of different core configurations to verify the calculated core reactivity. Changes in reactivity due to burnup were assumed to be linear and this assumption was verified for burnup intervals smaller than 3 MWd/kg(HM). The comparison was performed on 46 different core configurations with different type of fuel elements. The Serpent-2 calculations decently predict the rate of reactivity change on different cases, as 52 % of calculations are withing 1σ and 86.9 % within 2σ of the measurements for total number of 46 cases. Additional analysis was performed by comparing unit cell calculations of different fuel types. Four different types of TRIGA fuel were used to analyse burnup changes in LEU and HEU fuel, where positive reactivity feedback on burnup was observed for HEU fuel due to burnable absorbers. Serpent-2 and WIMSD-5B were compared on unit-cell basis where good agreement within 200 pcm of reactivity change for large burnup was observed. In addition neutron spectrum changes due to burnup were investigated using unit-cell calculations where 4 % increase of the thermal peak and 1 % decrease of fast peak of the spectrum was observed for typical fuel burnups of 20 MWd/kg(HM), which approximately represents JSI TRIGA burnup at this moment.
RESONANCE EVALUATION OF 239Pu IN THE RESONANCE REGION UP TO 4 keV
Authors: Leal Luiz , Leclaire Nicolas , Duhamel Isabelle
Published: February 2021

Resolved resonance evaluation of the 239Pu cross section in the energy range up 4 keV has been carried out with the SAMMY computer code. Existing resolved resonance evaluation such as ENDF/B-VIII.0 and JENDL4 data libraries for 239Pu is limited to 2.5 keV whereas above this energy the unresolved resonance methodology is used. High resolution transmission and fission data taken at the Oak Ridge Electron Linear Accelerator (ORELA) permitted extending the resonance region up to 4 keV. The thermal and average fission cross section values calculated with the resonance parameters derived in the evaluation fall close to that indicated in the suggested IAEA fission standards.
REACTOR NOISE ANALYSIS FOR A GRAPHITE-MODERATED AND -REFLECTED CORE IN KUCA
Authors: Sakon Atsushi , Nakajima Kunihiro , Takahashi Kazuki , Hohara Shin-ya , Sano Tadafumi , Fukaya Yuji , Hashimoto Kengo
Published: February 2021

In graphite-reflected thermal reactors, even a detector placed far from fuel region may detect a certain degree of the correlation amplitude. This is because mean free path of neutrons in graphite is longer than that in water or polyethylene. The objective of this study is experimentally to confirm a high flexibility of neutron detector placement in graphite reflector for reactor noise analysis. The present reactor noise analysis was carried out in a graphite-moderated and -reflected thermal core in Kyoto University Critical Assembly (KUCA). BF3 proportional neutron counters (1” dia.) were placed in graphite reflector region, where the counters were separated by about 35cm and 30cm -thick graphite from the core, respectively. At a critical state and subcritical states, time-sequence signal data from these counters were acquired and analyzed by a fast Fourier transform (FFT) analyzer, to obtain power spectral density in frequency domain. The auto-power spectral density obtained from the counters far from the core contained a significant degree of correlated component. A least-squares fit of a familiar formula to the auto-power spectral density data was made to determine the prompt-neutron decay constant. The decay constant was 63.3±14.5 [1/s] in critical state. The decay constant determined from the cross-power spectral density and coherence function data between the two counters also had a consistent value. It is confirmed that reactor noise analysis is possible using a detector placed at about 35cm far from the core, as we expected.
14 MeV NEUTRON IRRADIATION EXPERIMENTS - GAMMA SPECTROSCOPY ANALYSIS AND VALIDATION AUTOMATION
Authors: Stainer Thomas , Gilbert Mark R , Packer Lee W , Lilley Steven , Gopakumar Vignesh , Wilson Chris
Published: February 2021

An important area of research required for fusion reactor design is the study of materials under high energy neutron irradiation. Deuterium-Tritium (D-T) reactions release 14.1 MeV neutrons and material studies of such high energy neutrons focusing on transmutation and activation are paramount for fusion tokamak devices such as ITER and DEMO. In order to understand neutron damage and transmutation-induced radioactivity in fusion regime energies, a series of experimental campaigns were performed at the ASP facility based at Aldermaston in the UK, which uses a deuteron accelerator to bombard a tritiumloaded target and generate 14 MeV-neutron emission rates of up to 2.5 × 1011 s−1. In this work, a holistic treatment of the 11,000 gamma spectra (time series data) collected over five experimental campaigns is applied to identify radioisotopes and validate nuclear data and the inventory code, FISPACT-II. Whilst previous analysis has examined single spectra and foil irradiation’s using traditional, human-driven methods, this work applies novel methods using Artificial Neural Networks (ANN) and classification algorithms to allow a fully automated approach. Using such methods we show good broad agreement with FISPACT-II inventory simulations, and an overview of results are given as C/E values.
A COMPARISON OF OXIDE DECAY HEAT SIMULATIONS AND NUCLEAR DATA LIBRARIES WITH FUSION IRRADIATION EXPERIMENTS
Authors: Packer L W , Angelone M , Gilbert M R , Loreti S , Nobs C R , Pillon M , Sublet J-C , Vilkhivskaya O
Published: February 2021

We compare existing experimental decay heat data sets measured at the JAEA fusion neutron source (FNS), which employed a fast extraction rabbit system that in certain cases allowed the measurements to capture, at short cooling times, the decay profile of 16N in a range of oxides. Focussing on those data points and timescales that can be attributed to 16N, we compare measurements to simulations performed using the FISPACT-II inventory code together with evaluated nuclear data libraries. Making small corrections for other contributions at these short times, we derive integral cross section data estimates for 16O(n,p)16N from 12 oxide sample measurements and compare with previously obtained measurements in the IAEA EXFOR database and evaluations in the nuclear data libraries themselves.
USING PROMPT NEUTRON DECAY CONSTANT MEASUREMENTS TO OBTAIN ADDITIONAL KINETICS INFORMATION
Authors: McKenzie George , Grove Travis , Hutchinson Jesson
Published: February 2021

Prompt neutron decay constant measurements are typically used to infer the prompt neutron lifetime in a chain-reacting system, but these measurements include much more information about the kinetics of the system. This work focuses specifically on the Rossi-α method of measuring the prompt neutron decay constant. The Rossi-α system of equations consists of five unknown parameters. With five unknowns and three equations, the system of equations solved by the Rossi-α measurement is under-determined. It has been postulated that a well posed problem can be formulated using independent measurement configurations on a single system. The external source strength, the effective delayed neutron fractions, and Diven’s parameter do not typically change for different configurations of a single experiment. If good detector geometry is chosen, the detector efficiency does not change for multiple measurements of a single experiment using Rossi-α measurements. By making these assumptions, there are only two unique variables. This knowledge can be used to solve for most of the unique parameters using two independent, unique Rossi-α measurements.
FAST ROSSI-ALPHA MEASUREMENTS OF PLUTONIUM USING ORGANIC SCINTILLATORS
Authors: Hua M.Y. , Bravo C.A. , MacDonald A.T. , Hutchinson J.D. , McKenzie G.E. , Grove T.J. , Goda J.M. , McSpaden A.T. , Clarke S.D. , Pozzi S.A.
Published: February 2021

In this work, Rossi-alpha measurements were simultaneously performed with a 3He-based detection system and an organic scintillator-based detection system. The assembly is 15 kg of plutonium (93 wt% 239Pu) reflected by copper and moderated by lead. The goal of Rossi-alpha measurements is to estimate the prompt neutron decay constant, alpha. Simulations estimate keff = 0:624 and α = 52:3 ± 2:5 ns for the measured assembly. The organic scintillator system estimated α = 47:4 ± 2:0 ns, having a 9.37% error (though the 1.09 standard deviation confidence intervals overlapped). The 3He system estimated α = 37 μs. The known slowing down time of the 3He system is 35-40 μs, which means the slowing down time dominates and obscures the prompt neutron decay constant. Subsequently, the organic scintillator system should be used for assemblies with alpha much less than 35 μs.
VVER-1000 BENCHMARK INTERPRETATION WITH MONTE CARLO CODE MCS
Authors: Setiawan Fathurrahman , Lemaire Matthieu , Lee Hyunsuk , Zhang Peng , Lee Deokjung
Published: February 2021

An interpretation of the NEA-1517/82 benchmark from the SINBAD shielding database has been conducted with the MCS Monte Carlo code developed at the Ulsan National Institute of Science and Technology (UNIST) and the ENDF/B-VII.1 nuclear data library. The NEA-1517/82 benchmark corresponds to experiments on a VVER-1000 critical mock-up (thermal reactor with hexagonal fuel lattice) inside the LR-0 research reactor operated by the Nuclear Research Institute (NRI) in the Czech Republic. A new 3D model of the VVER-1000 mock-up core is developed for MCS based on the SINBAD documentation. The model includes the top and bottom parts of fuel pins, the spacer grids and core components: baffle, barrel, downcomer, tank, reactor pressure vessel (RPV) and concrete block used as biological shielding. The quality of the model is verified first by code/code comparison of MCS against MCNP6 for criticality and power distributions (pin-by-pin and axial power). The validation of MCS results is then performed against six critical cases, 260 measured pin powers and benchmark calculations of the axial power profile. Finally, a comparison of calculated and measured neutron spectra inside the mock-up core is presented as a preliminary study for upcoming works on the deep-penetration shielding capability of MCS.
VERA-CS VALIDATION OF CRITICAL EXPERIMENTS
Authors: Novellino Vincent , Palmtag Scott
Published: February 2021

VERA is a suite of multiphysics codes which uses MPACT to model neutron transport in light water reactors (LWRs) [1]. In this paper, we validate MPACT by modeling critical experiments conducted at the IPEN/MB-01 and B&W facilities. We modeled critical loading experiments with a variety of different fuel pins and materials placed in the core. The experiments were modeled in two dimensions using MPACT and an axial buckling term. Default mesh parameters exist in MPACT for modeling larger reactor cores, and a mesh convergence study was performed to find appropriate mesh parameters for modeling the smaller critical reactors. The keff results show a consistent bias and small standard deviation for the IPEN/MB-01 reactor and a small bias and small standard deviation for the B&W facility. Overall, the results show that MPACT performs well for modeling small critical reactors.
A SIMPLIFIED SMAHTR BENCHMARK PROBLEM SET
Authors: Reed K. Lisa , Rahnema Farzad
Published: February 2021

The Small, Modular Advanced High Temperature Reactor (SmAHTR) is a preconceptual design for a fluoride salt-cooled small modular reactor (SMR) [1]. In this paper, a stylized 2D benchmark problem set has been created based on SmAHTR. Certain gaps and considerations in burnable poison and control rod content were unspecified/undetermined in the preconceptual design, but those gaps were filled for the stylized problem set. With those features, this problem set could then be used for benchmarking neutron transport methods as well as its low order methods in 2D single assembly and full core configurations.

For this benchmark set, continuous energy Monte Carlo calculations were performed. Those calculations provided keff values of 0.9459 (± 11 pcm) and 1.1436 (± 12 pcm) in the full core configuration with all the control rods fully inserted and withdrawn, respectively. The single assembly calculations yielded an eigenvalue, kinf, of 0.9987 (± 15 pcm) and 1.2117 (± 15 pcm) with all of the control rods either inserted or removed, respectively. In the full core configuration, the worth of all the control rods and burnable poison particles were determined to be 197.6 (± 0.16) mk and 311.6 (± 0.23) mk, respectively. The corresponding results in the single assembly configurations are 213 (± 0.21) mk and 337.4 (± 0.20) mk, respectively. A near-critical configuration was also determined for the reactor by inserting control rods in some assemblies, thus providing a case with a keff value of 0.9909 (± 12 pcm).
EXTENDING CTF MODELING CAPABILITIES TO SFRs AND VALIDATION AGAINST SHRT TESTS
Authors: Aly Ahmed , Abarca Agustin , Avramova Maria , Ivanov Kostadin
Published: February 2021

The utilization of liquid metals as coolants for fast reactors brings several economical and practical advantages that lead to a sustainable future for nuclear energy. Molten sodium is used as a coolant in Sodium Fast Reactors (SFRs). Sodium is relatively cheaper than other metal coolants. It requires lower pumping power, causes less neutron moderation and it is non-corrosive to the fuel cladding. The SFR hexagonal subassemblies are relatively smaller than Light Water Reactors (LWRs) subassemblies. The differences in the geometrical design of SFRs compared to LWRs lead to different physical behavior of the coolant. Several models and correlations particular to sodium were implemented in thermal-hydraulics (TH) computer codes in order to model the coolant behavior accurately. CTF is a subchannel TH code that was designed and validated for LWRs. In this work, the capabilities of the code were extended to SFRs by incorporating sodium coolant properties and correlations for friction factor, flow mixing coefficient and conduction heat transfer. The code was then validated against selected steady state data from the Experimental Breeder Reactor II Shutdown Heat Removal Tests SHRT-17 and SHRT-45R. CTF was used to simulate the instrumented subassemblies XX09 and XX10. The results demonstrate the capability of CTF to model SFRs. Code validation is currently being extended to the transient phases of the SHRT experiments.
NEXUS FRAMEWORK FOR WHOLE-CORE FUEL PERFORMANCE: CURRENT APPLICATIONS AND FUTURE TRENDS
Authors: Peakman Aiden , Bennett Thomas , Fitzgerald Kerr , Gregg Robert , Rossiter Glyn
Published: February 2021

Current industry practice in fuel licensing often relies on thermo-mechanical modeling of a fuel rod with an artificially constructed bounding power history. The benefit of this approach is that it is computationally efficient; however, the drawbacks are that 1) such an approach is not always conservative, for instance when modelling phenomena related to late onset pellet-clad gap closure; and 2) it can poorly estimate available safety margins for fuel operating at high local power densities and/or to high burnup. For these reasons NNL developed an in-house whole-core fuel performance framework – NEXUS – to enable modelling of all fuel rods in the core using the ENIGMA fuel performance code and computed power histories from core simulation packages (currently limited to PARCS or SIMULATE). One of the main objectives was to create a tool that was both computationally efficient and user friendly. The former was achieved by making use of parallelisable architecture, while the latter was achieved by minimising necessary user input and providing tools for easy interrogation of the fuel performance output.

NEXUS has been applied to several LWR operational scenarios, which we summarise in this paper, including steady-state operation of an ABWR, and a rod ejection accident in a small modular soluble boron free PWR and a GWe-class PWR. We also summarise current development activities related to integrating NNL’s in-house fuel performance Monte Carlo uncertainty analysis software CASINO into the NEXUS framework.
SURROGATE MODEL OPTIMIZATION OF A ‘MICRO CORE’ PWR FUEL ASSEMBLY ARRANGEMENT USING DEEP LEARNING MODELS
Authors: Whyte Andy , Parks Geoff
Published: February 2021

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.
A METHOD FOR STUDY OF NUCLEAR FUEL CYCLE SCENARIOS WITH AN INTENSIVE ELECTRICITY PRODUCTION DECLINE IN A SHORT PERIOD
Authors: Zhou Weifeng , Krivtchik Guillaume , Blaise Patrick
Published: February 2021

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.
“CHAPTER 5” DIFFUSION THEORY
Authors: Ganapol B. , Tsvetkov P.
Published: February 2021

In many nuclear reactor physics texts (excluding Elmer Lewis’s recent text however), “Chapter 5” is dedicated to diffusion theory; hence, the title of this submission. Here, we will investigate analytical solutions to the most basic form of the monoenergetic 1D stationary diffusion equation. The intuitive approach taken radically departs from the usual method of solving the diffusion equation. In particular, we consider a general setting such that the method accommodates all solutions to the monoenergetic diffusion equations in 1D plane and curvilinear geometries. This is not your father’s diffusion theory and, for this reason, we anticipate it will eventually become the classroom standard.
BETA EFFECTIVE SENSITIVITY TO NUCLEAR DATA WITHIN THE APOLLO3® NEUTRONIC PLATFORM
Authors: Valocchi G. , Archier P. , Tommasi J.
Published: February 2021

In this paper, we present a sensitivity analysis of the beta effective to nuclear data for the UM17x17 experiment that has been performed in the EOLE reactor. This work is carried out using the APOLLO3® platform. Regarding the flux calculation, the standard two-step approach (lattice/core) is used. For what concerns the delayed nuclear data, they are processed to be directly used in the core calculation without going through the lattice one. We use the JEFF-3.1.1 nuclear data library for cross-sections and delayed data. The calculation of k-effective and beta effective is validated against a TRIPOLI4® one while the main sensitivities are validated against direct calculation. Finally, uncertainty propagation is performed using the COMAC-V2.0 covariance library.
A PRACTICAL EFFECT OF THE RECENT O16 CROSS SECTION AND MUBAR UNCERTAINTIES
Authors: Parsons D. Kent , Turner Scott A. , Jaegers Peter J.
Published: February 2021

With the recent release of ENDF/B VIII.0 data, additional covariance data was provided for many isotopes, including O16. The detail of elastic scattering and mubar covariance data for O16 increased dramatically between ENDF/B VII.1 and ENDF/B VIII.0. This new covariance data has been processed with NJOY2016 and investigation has begun on the effects of these new uncertainty data. The uncertainties are applied to multi-group scattering cross sections and P1 Legendre components in deterministic neutron transport.

A simple but typical application of shielding fission neutrons with concrete has been used to assess the practical effects of the new covariance data for O16. A somewhat surprising result is that the mubar uncertainty can have a significant effect on the calculated shielding and criticality results.
BURNUP SENSITIVITY CALCULATIONS WITH CBZ FOR LIGHT WATER REACTOR ASSEMBLY PROBLEMS
Authors: Chiba Go
Published: February 2021

Sensitivities of k∞ and nuclides number densities during nuclear fuel burnup with respect to nuclear data are calculated with a reactor physics code system CBZ. Sensitivity calculations are carried out with the depletion perturbation theory applicable to nuclear fuel assemblies including burnable absorbers. Numerical results are presented both for BWR and PWR assemblies, and those demonstrate usefulness and effectiveness of burnup sensitivity calculation capabilities for LWR fuel assemblies.
ANSWERS® TOOLS FOR UNCERTAINTY QUANTIFICATION AND VALIDATION
Authors: Smith Paul N , Hanlon Dave , Dobson Geoff , Hiles Richard , Fry Tim , Stefanowska Magda , Richards Simon , Murphy Christophe
Published: February 2021

ANSWERS® is developing a set of uncertainty quantification (UQ) tools for use with its major physics codes: WIMS/PANTHER (reactor physics), MONK (criticality and reactor physics) and MCBEND (shielding and dosimetry). The Visual Workshop integrated development environment allows the user to construct and edit code inputs, launch calculations, post-process results and produce graphs, and recently uncertainty quantification and optimisation tools have been added.

Prior uncertainties due to uncertainties in nuclear data or manufacturing tolerances can be estimated using the sampling method or using the sensitivity options in the physics codes combined with appropriate covariance matrices. To aid the user in the choice of appropriate validation experiments, the MONK categorisation scheme and/or a similarity index can be used. An interactive viewer has been developed which allows the user to search through, and browse details of, over 2,000 MONK validation experiments that have been analysed from the ICSBEP and IRPhE validation sets. A Bayesian updating approach is used to assimilate the measured data with the calculated results. It is shown how this process can be used to reduce bias in calculated results and reduce the calculated uncertainty on those results. This process is illustrated by application to a PWR fuel assembly.
MODELING NUCLEAR DATA UNCERTAINTIES USING DEEP NEURAL NETWORKS
Authors: Radaideh Majdi I. , Price Dean , Kozlowski Tomasz
Published: February 2021

A new concept using deep learning in neural networks is investigated to characterize the underlying uncertainty of nuclear data. Analysis is performed on multi-group neutron cross-sections (56 energy groups) for the GODIVA U-235 sphere. A deep model is trained with cross-validation using 1000 nuclear data random samples to fit 336 nuclear data parameters. Although of the very limited sample size (1000 samples) available in this study, the trained models demonstrate promising performance, where a prediction error of about 166 pcm is found for keff in the test set. In addition, the deep model’s sensitivity and uncertainty are validated. The comparison of importance ranking of the principal fast fission energy groups with adjoint methods shows fair agreement, while a very good agreement is observed when comparing the global keff uncertainty with sampling methods. The findings of this work shall motivate additional efforts on using machine learning to unravel complexities in nuclear data research.
EVALUATION OF THE DOSE LEVELS IN THE IRRADIATION CHANNELS OF PU-BE NEUTRON SOURCES STORAGE DEVICE BY USING NEUTRON ACTIVATION ANALYSIS AND MCNP6 MODELLING
Authors: Lagzdina E. , Plukienė R. , Lingis D. , Garankin J. , Plukis A. , Germanas D. , Remeikis V.
Published: February 2021

PuBeneutrons sources of Educational Neutron Laboratory in the Center for Physical Sciences and Technology are used in several activities including neutron activation for educational purposes and laboratory irradiation experiments. In this work we address the dose rate estimation in the irradiation channels of reconstructed PuBe neutron sources storage device in the particular positions dedicated for neutron irradiation experiments using neutron activation analysis (NAA) and MCNP6 modelling.The results on neutron activation of V2O5, MnO2, Al and Na2CO3were analyzed and compared with MCNP6 model prediction results. The conclusions on the actual neutron flux energy distribution and model corrections are drawn. The absolute neutron intensity at dedicated irradiation points as well as neutron and gamma dose rates are obtained.