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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. [ABSTRACT FROM AUTHOR]

Available nuclear data for molybdenum included in the nuclear data libraries are not of sufficient quality for reactor physics or criticality safety issues and indeed information about uncertainties and covariance is either missing or leaves much to be desired. Therefore, IRSN and JAEA performed experimental measurements on molybdenum at the J-PARC (Japan Proton Accelerator Research Complex) facility in Japan. The aim was to measure capture cross section and transmission of natural molybdenum at the ANNRI (Accurate Neutron-Nucleus Reaction measurement Instrument) in the MLF (Material Life and science Facility) of J-PARC. The measurements were performed on metallic natural molybdenum samples with various thicknesses. A NaI detector, placed at a flight-path length of about 28 m, was used for capture measurements and a Li-glass detector (flight-path length of about 28.7 m) for transmission measurements. Following the data reduction process, the measured data are being analyzed and evaluated to produce more accurate cross sections and associated uncertainties. [ABSTRACT FROM AUTHOR]

The probability table method is widely used for continuous energy Monte Carlo calculation codes to treat the self-shielding effect in the unresolved resonance region. The ladder method is used to calculate the probability table. This method generates pseudo resonance structures using random numbers based on the averaged resonance parameters. The probability table affects the ladder number, i.e., the number of pseudo resonance structures. Since the higher the ladder number the longer it takes to create a cross-section library, reducing the value of the ladder number is an effective way to reduce the time spent in the creation of a cross-section library. However, the exact value of the most appropriate ladder number to use has not been found. Currently, users have to use a nuclear data processing system to manually set the ladder number to a large enough value in order to obtain a sufficiently converged probability table. In this study, the probability tables for all nuclides prepared in JENDL-4.0 are used to find the optimal ladder number. The results show that the probability table differences are small enough when the ladder number is 100. The impact of these differences on neutronics calculations is found to be small. [ABSTRACT FROM AUTHOR]

Pressurized water reactors are typically surrounded in the radial direction by neutron reflectors made from stainless steel and water. These reflectors decrease neutron leakage and provide protection of pressure vessel from fast neutrons damaging its integrity. Such a radial reflector influences multiplication factor of the core and distribution of neutron flux and fission power inside the core. All these effects can be analyzed by full-core simulations using macroscopic constants. Methodology for generation of the macroscopic constants for non-fuel regions will be tested for new stainless steel reflectors at the VR-1 reactor. Rods from SS 304l material will be used for construction of radial reflectors for the VR-1 reactor. They will be design to generate sufficient measurable response in selected core characteristics. The study is focused on core power distribution and reactivity worth of absorbing rods in a VR-1 reactor core. The core typically consists of about 20 IRT-4M fuel assemblies and seven absorbing rods UR-70. Replacing water surrounding the core by several reflector assemblies containing stainless steel will influence leakage and distribution of neutrons inside the core. The current analysis deals with local effects and employs the sensitivity study to discover the nature of reflectors' impact on the reactor core. These effects were studied even for several past VR-1 reactor core configurations. All calculations were carried out in Serpent2 Monte-Carlo code with various evaluated libraries: ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 data. [ABSTRACT FROM AUTHOR]

This paper describes the effect of input uncertainties on a set of integral parameters (kinf, nuclide compositions) associated with the validation of CASMO-5 against PIE data. The nuclear data under consideration are the cross-sections, fission spectrum and neutron multiplicities and fission yields. Various sources of covariance information are considered, novel ones (ENDFB-VIII.0, JEFF-3.3) as well as more widely distributed ones (JENDL-4.0, ENDF/B-VII.1, Scale 6.1 and Scale 6.2). All possible nuclide reaction pairs (cross sections, fission spectrum and averaged number of neutron per fission) have been perturbed, e.g. all isotopes available in both the respective covariance libraries and the CASMO-5 library. The evolution of the uncertainty estimates with exposure is complemented with sensitivity analysis to determine the main contributors to the uncertainty. The Pearson coefficient defined between the model output and a given input is used in this work to assess the part of the variance in the model output coming from the considered input uncertainty. It is a very promising measure of sensitivity as it is computationally cheap even though it assumes linearity of the output with respect to input perturbations. The evolution of the uncertainty with exposure, both in terms of trends and magnitude are however very different. Sensitivity analysis allows determining why the trends and magnitudes are different. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

A new nuclear data library for the CASMO5 advanced lattice physics code has been generated based on the recently-released ENDF/B-VIII.0 evaluation. The ENDF/B-VIII.0 evaluation represents the state-of-the-art in nuclear data and features new evaluations from the CIELO project for 1H, 16O, 56Fe, 235U, 238U and 239Pu. A description of the library generation procedure used to process these data into the CASMO5 586 energy group structure is provided. Initial validation of the new ENDF/B-VIII.0-based library, referred to as the E8R0 library, is also presented and involves the comparison of predicted k–eff and fission rate distributions to measurements from various critical experiments. The critical experiments used in the initial validation of the E8R0 library consist of the B&W 1810 series, B&W 1484 series, DIMPLE S06A/B, and TCA reflector experiment with iron plates. The results from the initial validation indicate that the new E8R0 library provides a satisfactory performance in terms of CASMO5 predicted k–eff and fission distributions. [ABSTRACT FROM AUTHOR]

The PETALE experimental program in the CROCUS reactor intends to provide integral measurements on reactivity worth and dosimetry measurement to constrain nuclear data relative to stainless steel heavy reflectors. The experimental setup consists in eight successive plates of pure iron, pure nickel, pure chromium, or nuclear-grade stainless steel set at the close periphery of the core. The plates are interleaved with up to nine dosimeters that consist of thin activation foils with different possible materials to be sensitive to different ranges of the neutron spectrum. A precise measurement with a good estimation of the uncertainties and correlations is required, especially when comparing reaction rates, e.g. transmission measurement and/or spectral indices. The present work focuses on the validation of the dosimetry technics developed in preparation of this experimental program. Different aspects are discussed: monitors, efficiency calibration, self-absorption correction, self-shielding and nuclear data uncertainties. The different sources of uncertainties for the experiment-calculation comparisons are characterized, taking into account all the correlation between the different dosimeters. These correlations are a mandatory element for the aimed Bayesian assimilation in order to avoir overfitting when considering dosimeter providing a similar information. [ABSTRACT FROM AUTHOR]

Highly localized in-core measurements are necessary for the validation of neutron transport calculations with high spatial resolution. In the present work, a miniature neutron detector developed at EPFL in collaboration with PSI was used to carry out a set of thermal neutrons counting measurements in the zero-power CROCUS reactor core within a spatial range in order of mm. The miniature detector, positioned close to the core reflector, shows a gradient of +(4.29 ± 0.10)% in the count rate profile in the radial direction within 1.3 cm, with higher values pointing towards the core reflector because of the higher share of neutrons in the thermal range. On the contrary, in a control rod guide tube the count rate gradient is -(4.37 ± 0.10)% and it is directed towards the core center. The measured values are compared with the azimuthal trend of the normalized 6Li reaction rate calculated with an iterative three-steps method performed with the Monte Carlo code Serpent 2. These measurements proved the feasibility of resolving spatial effects in the mm-range and they represent a basis for further investigating highly spatially-resolved phenomena in the CROCUS core. [ABSTRACT FROM AUTHOR]

An evaluation of validity of theoretical pulse models (Fuchs-Hansen and Nordheim-Fuchs model) in reactor pulse modelling was performed by developing so called Improved Pulse Model. The effect of each of the five assumptions on the most important pulse physical parameters, maximal power, total released energy and full width at half maximum was studied. In the Improved Pulse Model the assumptions are disposed out with the improvements, where to account the delayed neutrons the six point kinetic equations are solved, the temperature dependences of the temperature reactivity coefficient of fuel and specific heat are taken into account, also the final ejected time of transient control rod from reactor core, whose value of reactivity varies in height and the heat dissipation from the fuel are considered. It is found that the theoretical models predict a higher maximum power, lower total released energy and full width at half maximum than the Improved Pulse Model. [ABSTRACT FROM AUTHOR]

The Modified Source Multiplication method is used to determine an unknown reactivity level of a reactor from a known one if one has access to the detector counting for both levels when the reactor is fed by a constant neutron source like an Am-Be source. When available, an accelerator driven source, in continuous mode, can be useful as its intensity can be tunable and then adapted to the experimental conditions. However, in that case, the MSM technique must be extended to account for an external source whose intensity, energy and angular distributions can vary from one measurement to another. In this paper, this Modified Multi-Source Multiplication (MMSM) method is applied to measurements done during the FREYA project in the GUINEVERE facility, operated with the GENEPI-3C accelerator providing a mixture of (D,T) and (D,D) neutrons. The monitoring of these sources through the detection of the associated charged particles allows the calculation of the MMSM factors and the estimate of the reactivity values. The results are compared in different configurations with the reactivity obtained with an Am-Be source or in dynamic measurements performed with GENEPI-3C. Their excellent agreement shows the possibility of using such accelerator-based neutron sources for MSM measurements when they are correctly monitored. This is of great interest for deep sub-critical level characterization for which detector count rates per source neutrons might be low. [ABSTRACT FROM AUTHOR]

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

Three sub-critical (SC) core configurations were investigated in the VENUS-F zero power reactor coupled with the GENEPI-3C accelerator. The SC10 and SC12 were a mock-up of a MYRRHA start-up core and SC11 represented a more complex MYRRHA core loaded with various types of in-pile-sections. The sub-criticality of 11 variants of these VENUS-F cores was changed in several steps from -6$ down to -30$ using the safety and control rods. Their sub-criticalities were determined with the Source Jerk Integral (SJI) method using 11 fission chambers located all over the reactor. For the data analysis, the 8-group delayed neutron parameters from the JEFF-3.1.2 evaluated nuclear data library were used. Reliability and reproducibility of the experimental results were tested by repeating the measurements, swapping the detectors and varying the accelerator beam intensity, thus changing the detector count rates and verifying the validity of the dead time corrections. The obtained results are compared with MCNP calculations. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

In this work, we present the first dynamic calculations performed with the Monte Carlo neutron transport code TRIPOLI-4R with thermal-hydraulics feedback. For this purpose, the Monte Carlo code was extended for multi-physics capabilities and coupled to the thermal-hydraulics subchannel code SUBCHANFLOW. As a test case for the verification of transient simulation capabilities, a 3x3-assembly mini-core benchmark based on the TMI-1 reactor is considered with a pin-by-pin description. Two reactivity excursion scenarios initiated by control-rod movement are simulated starting from a critical state and compared to analogous simulations performed using the Serpent 2 Monte-Carlo code. The time evolution of the neutron power, fuel temperature, coolant temperature and coolant density are analysed to assess the multi-physics capabilities of TRIPOLI-4. The stabilizing e_ects of thermal-hydraulics on the neutron power appear to be well taken into account. The computational requirements for massively parallel calculations are also discussed. [ABSTRACT FROM AUTHOR]

Cold-startup and hot-standby reactivity accident tests conducted at the SPERT III E-core research reactor are analysed with the coupled neutron-kinetic/thermal-hydraulic code system DYN3D-ATHLET. Homogenised 2-group cross sections for DYN3D are thereby generated with the Monte Carlo neutron transport code Serpent 2 in combination with the ENDF/B-VII.1 cross section library. Results in terms of maximum power, energy release, and reactivity compensation are in good agreement with the experimental values. The time-dependent contributions to the reactivity feedback are investigated for both a cold-startup test and a hot-standby test. These findings prove the suitability of the combined application of the simulation codes to predict the reactor dynamic behaviour in the event of prompt-critical and super-prompt critical transients even for small reactor cores. Furthermore, static core characteristics of the SPERT III E-core reactor at cold-startup condition are analysed with using a static DYN3D model, a detailed Serpent reference model, and a simplified Serpent model consistent with the DYN3D model. The critical control rod position and the excess reactivities of both the control rods and the transient rod obtained with the Serpent reference model are consistent with the experimental values. For the same parameters, the DYN3D model is in good agreement with the Serpent simplified model. [ABSTRACT FROM AUTHOR]

Due to the high confinement of the closed fuel assemblies in Sodium-cooled Fast Reactors (SFR), the liquid sodium filling the space between these assemblies may vaporize in case of an important mechanical excitation (cavitation process). That phenomenon would therefore induce a specific force feedback to the assemblies leading to a specific neutronic response of the core. Neutronic of SFR is indeed sensible to mechanical deformation. Hence the necessity to have a good prediction of the displacement of these mechanical structures in the framework of Fluid-Structure Interaction (FSI). To study this kind of transient at a reactor scale, we propose in this paper a coupling procedure between the fluid and the structures. The fluid behaviour is investigated by Merkle's three-equations model and simulated within a coarse mesh. The structures are modelled by a mass-spring system subjected to the fluid forces. In order to stabilize the numerical code coupling, a relaxation process is added and some results of this computational work are presented. [ABSTRACT FROM AUTHOR]

We present an extension of the Analytic Discrete-Ordinates method to time-dependent transport in finite media. The application of this technique to time-dependent transport is primarily accomplished through the use of a Laplace transform. In the case of finite media, a system of equations arises from enforcing boundary conditions. Instead of directly solving this system, we construct a solution in terms of a Neumann series. We then show that terms can be neglected when numerically evaluating the inverse Laplace transform such that the series reduces to a finite sum. With this extension, we use convergence acceleration to generate a high-precision benchmark. [ABSTRACT FROM AUTHOR]

Several methods have recently been developed to solve multiphysics transients using the Improved Quasi-Static method and its derivatives. In order to address some perceived drawbacks of these methods, we have developed a new method for solving multiphysics transient calculations using the Predictor-Corrector Quasi-Static method. Our method involves computing reactivity feedback parameters during each transport timestep in order to enable reactivity feedback on the small timescale used by the point kinetics phase of the calculation. The advantage of this approach is that the transport solver does not need to store local flux information between time steps, potentially making it more appropriate for use with a Monte Carlo solver. We have demonstrated the accuracy of our method by solving a simple model problem that exhibits difficult multiphysics behavior. Additionally, we have compared our results against another recently published multiphysics coupling scheme, confirming that our approach does not negatively affect the accuracy of the transient solution. [ABSTRACT FROM AUTHOR]

The distribution of the power inside a reactor core can be described by the time dependent multigroup neutron diffusion equation. One of the approaches to integrate this time-dependent equation is the modal method, that assumes that the solution can be described by the sum of amplitude function multiplied by shape functions of modes. These shape functions can be computed by solving a _-modes problems. The modal method has a great interest when the distribution of the power cannot be well approximated by only one shape function, mainly, when local perturbations are applied during the transient. Usually, the shape functions of the modal methods are updated for the time-dependent equations with a constant time-step size to obtain accurate results. In this work, we propose a modal methodology with an adaptive control time-step to update the eigenfunctions associated with the modes. This algorithm improves efficiency because of time is not spent solving the systems to a level of accuracy beyond relevance and reduces the step size if they detect a numerical instability. Step size controllers require an error estimation. Different error estimations are considered and analyzed in a benchmark problem with a out of phase local perturbation. [ABSTRACT FROM AUTHOR]

Decay heat (DH) is the heat produced through a radioactive decay of fission products during or after a reactor operation. It is known as the second largest source of power in the core after fission. Being such a strong contributor to reactor power, it should be accurately determined at any time of reactor operation. Currently, there are two main approaches for DH estimation used in reactor simulation codes. One approach is based on careful inventorying of all produced target nuclides and their individual contributions to total power. Alternatively, the other popular approach is based on collapsing all target fission products into a small number of groups similar to delayed neutron estimation techniques. However, the last (multigroup) method currently has limitations when used in some transient scenarios such as transients occurred in fresh fuel. In this study, the multigroup method was further developed for reducing limitations while retaining the advantage in computation speed. Then, it was implemented into Reactor Analysis code for Steady state and Transient (RAST-K) and tested against other codes. As a result, the improved method was found capable of determining DH power at all tested stages of reactor operation under any tested operation scenario. In particular, the test simulations using the improved method showed better results in those scenarios that were under accuracy limitations of the original multigroup method. Overall, the quality of transient calculations in RAST-K was improved when using the newly implemented DH module. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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

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. [ABSTRACT FROM AUTHOR]

The assumption that void fraction, and by extension coolant density, is uniform in the radial direction is a common approximation used in lattice physics simulations. In this study, models without uniform radial void fraction are used and lattice criticality and pin powers are investigated in two ways. One way uses hypothetical models that reflect extreme radial void distributions; modifications such as uniform radial pin enrichment and the removal of gadolinium rods are included in these models as well. Experimentally-determined boiling water reactor radial void distributions are also replicated in neutronics models using Serpent 2. It is found in the hypothetical models that the presence of gadolinium rods has a large effect on the interaction between lattice criticality and radial void distribution. It was also found that considering experimental radial void fraction distributions had the largest effect on the pin power of the rods containing gadolinium. Furthermore, it is observed that considering realistic radial void distributions, in general, decreased lattice criticality. The reason can be attributed that to the passive negative-feedback design of light water reactors. These are useful findings because calculation of more accurate peaking factors can lead to efficient and yet safer reactor operation. [ABSTRACT FROM AUTHOR]

While the literature has numerous examples of Monte Carlo and computational fluid dynamics (CFD) coupling, most are hard-wired codes intended primarily for research rather than as standalone, general-purpose codes. In this work, we describe an open source application, ENRICO, that allows coupled neutronic and thermal-hydraulic simulations between multiple codes that can be chosen at runtime (as opposed to a coupling between two specific codes). In particular, we outline the class hierarchy in ENRICO and show how it enables a clean separation between the logic and data required for a coupled simulation (which is agnostic to the individual solvers used) from the logic/data required for individual physics solvers. ENRICO also allows coupling between high-order (and generally computationally expensive) solvers to low-order "surrogate" solvers; for example, Nek5000 can be swapped out with a subchannel solver. ENRICO has been designed for use on distributed-memory computing environments. The transfer of solution fields between solvers is performed in memory rather than through file I/O.We describe the process topology among the different solvers and how it is leveraged to carry out solution transfers. We present results for a coupled simulation of a single light-water reactor fuel assembly using Monte Carlo neutron transport and CFD. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

In this paper we present a computational exercise for shut-down dose rate calculations for the JET tokamak using the in-house developed JSIR2S code package as part of its validation. The computation is performed in two parts: neutron transport and transport of secondary gamma radiation. In order to calculate neutron activation reaction rates with sufficiently low variance, hybrid variance reduction techniques using the ADVANTG code have been utilized. Probability based sampling of secondary source particles was performed. Calculated gamma dose rates after shut down are compared with dose rate measurements performed on site using ionization chambers. The C/E agreement for 1st octant is between 0.8 to 1 while statistically meaningfull results for the 2nd octant are yet to be obtained. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Advances in computational methods have given rise to the study and simulation of different aspects of reactor behavior. As such, topics associated with high computational costs become feasible candidates for further investigation and one of them is reactor space-time kinetics (STK). Until recently, STK simulation and point kinetics approximation were limited to deterministic codes, with Monte Carlo codes being too costly to start with. However, recent developments in this area have allowed the use of certain methods in stochastic codes. One such technique is based on the Transient Fission Matrix (TFM) model, a hybrid method that uses a system response obtained through Monte Carlo and stored in fission and time matrices as input for deterministic calculations. The result enables a view of the STK in terms of neutron propagation probability and propagation time across the system. The TFM method was applied to a simple coupled core configuration to generate a numerical benchmark. The Serpent 2 Monte Carlo code was used for the stochastic part of the calculation. The configuration consists of two fuel assemblies placed in a light water tank, with a water blade of varying width between them. TFM, flux and fission results were obtained for varying water blade widths, ranging between 0 cm and 20 cm. The data is then used to analyze the behavior of the system, as well as the effects of the coupling between the two assemblies. As the assemblies move further apart, the system slowly transitions from two tightly coupled assemblies that essentially form a single core, to two almost independent cores. This study enables to produce a benchmark for future calculations and predefine an innovative way of designing high dominant ratio configurations, required for tackling Monte Carlo residual problems. An actual experimental program could be led in ad hoc zero power reactor (ZPR), such as the KUCA reactor of Kyoto University. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

The whole-core transport code nTRACER has made many advances in recent years. Several innovative cross section treatment methods were developed, a new axial transport solver was introduced for stabilizing the 2D/1D scheme, and substantial computational enhancements were achieved using NVIDIA CUDA and Intel Math Kernel Library (MKL). In addition, gamma transport solver was implemented to predict the power distributions more physically, and the flexibility of the restart calculation was improved using an offline processing code nTIG (nTRACER Input Generator). This paper is the compilation of the recent progresses in nTRACER developments. [ABSTRACT FROM AUTHOR]

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

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. [ABSTRACT FROM AUTHOR]

Molten salt reactors as liquid-fuelled reactors are flexible in terms of operation or design choices, but they are very different in terms of design, operation and safety approach compared to solid-fuelled reactors. Such reactors call for a new definition of their operating procedures and safety approach. Dedicated developments and studies have been performed in the frame of the European SAMOFAR project of Horizon2020 and in parallel in France involving CNRS, CORYS and Framatome to develop a system code called LiCore adapted to such reactors, corresponding to a basic-principle power plant simulator. The neutronic model LiCore, at the centre of the simulator, corresponds to an improved point-kinetics model to take into account the specificities of a MSR, notably the circulation of the delayed neutron precursors out of the core. Coupled to a simple piston model for the fuel motion in the core, this code can perform calculations faster than real time to simulate the behaviour of the fuel circuit. Transient calculations performed with LiCore are presented, together with comparisons first to a simple point-kinetics model and then to 3D calculations with the TFM-OpenFOAM coupled code. Finally, the LiCore code has recently been integrated in the ALICES platform, the integrated simulation toolset designed by CORYS for the development, maintenance and operation of major simulator such as power plant simulators. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Multi-physics coupled simulations have become increasingly important during the last two decades being one of the major field of application in the nuclear technology. The nuclear reactors themselves are complex systems whose responses are driven by interactions between neutron kinetics, thermal-hydraulics, heat transfer, mechanics and chemistry. Probably, in a nuclear system, the most complex and important feedback effect takes place between the core neutron kinetics and thermal-hydraulics. The development of coupled thermal-hydraulic -neutron kinetics codes is a recurrent field of research for the nuclear industry. This contribution, developed in the Consortium for Nuclear Power (CNP) framework, has the objective of develop a dynamic coupling, using TCP/IP based socket communication, between the thermal-hydraulic system code T-TRACE, Tractebel-ENGIE version of the latest US NRC TRACE release, and the multi-group 3-D nodal diffusion and core physics code PANTHER, developed and maintained by EDF Energy (UK). As a first step of the development, a fully temporally explicit coupling scheme has been developed between TRACE and PANTHER based on a boundary conditions exchange at the core level at each temporal iteration. The OECD TMI MSLB benchmark has been selected as verification scenario for testing the ongoing developing T-TRACE/PANTHER coupled code. The developed coupled code is benchmarked code-to-code against TRACE/PARCS and T-RELAP5/PANTHER. [ABSTRACT FROM AUTHOR]

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

High temperature gas cooled reactors (HTGR) are a candidate for timely Gen-IV reactor technology deployment because of high technology readiness and walk-away safety. Among HTGRs, pebble bed reactors (PBRs) have attractive features such as low excess reactivity and online refueling. Pebble bed reactors pose unique challenges to analysts and reactor designers such as continuous burnup distribution depending on pebble motion and recirculation, radiative heat transfer across a variety of gas-filled gaps, and long design basis transients such as pressurized and depressurized loss of forced circulation. Modeling and simulation is essential for both the PBR's safety case and design process. In order to verify and validate the new generation codes the Nuclear Energy Agency (NEA) Data bank provide a set of benchmarks data together with solutions calculated by the participants using the state of the art codes of that time. An important milestone to test the new PBR simulation codes is the OECD NEA PBMR-400 benchmark which includes thermal hydraulic and neutron kinetic standalone exercises as well as coupled exercises and transients scenarios. In this work, the reactor multiphysics code MAMMOTH and the thermal hydraulics code Pronghorn, both developed by the Idaho National Laboratory (INL) within the multiphysics object-oriented simulation environment (MOOSE), have been used to solve Phase 1 exercises 1 and 2 of the PBMR-400 benchmark. The steady state results are in agreement with the other participants' solutions demonstrating the adequacy of MAMMOTH and Pronghorn for simulating PBRs. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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

JHR is a new material testing reactor under construction at CEA Cadarache. Its high flux core contains 37 fuel assemblies loaded along concentric rings into alveolus of an aluminum matrix. For the operation of the reactor, twenty-seven of these fuel assemblies hovnst hafnium rods in their center while the other ones but also the beryllium radial reflector can accommodate experimental devices. In order to accurately predict its operating core characteristics but also its irradiation performance, a recently developed scheme based on the APOLLO3® platform is being developed which uses the sub-group method for spatial self-shielding, the 2D method of characteristics and the 3D unstructured conform MINARET Sn transport solver. A 2D model of JHR has been built and optimized for calculating, at the lattice step, the self-shielded and condensed cross sections thanks to the sub-group method and the method of characteristics. Results are benchmarked against a TRIPOLI-4® stochastic reference calculation. A more refined spatial mesh gives better results on fission rates and reactivity compared to the ones of the former APOLLO2 scheme. The classical 2-step calculations use the hypothesis of infinite lattice configuration, which is reasonable for the assemblies close to the center but not for peripheral ones. Hence, a new approach is being set up taking into account the surrounding of each assembly. The newly 3-step scheme uses the Sn solver MINARET and gives better results than the traditional 2-step scheme. This approach will be applied to a 3D modelling of the heterogeneous JHR core configurations incorporating experimental devices and enabling burn up calculations. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

This work concerns the numerical optimization of a multiphysics calculation scheme. The considered application is a 5x5 Pressurized Water Reactor (PWR) assemblies mini-core surrounded by radial and axial reflectors. The scenario adopted for the analysis is steady-state nominal conditions and fission products set to the equilibrium concentration. The neutronics is modelled at the pin-cell scale and the thermal-hydraulics at the subchannel level. Depending on the scenario, the damped fixed-point algorithm might not be sufficiently robust or efficient enough. For this reason, a technique based on the partial convergence of the solvers is tested. In every multiphysic iteration, a maximum number of iterations is imposed for both the neutronics and the thermal-hydraulics solvers. In combination with that, the solver restarts from the results of the last calculation. In this way, if the method is convergent, the initialization progresses towards the fixed-point solution. The results show that such a technique improves both the robustness and the speed of the algorithm. Within this approach, the range of relaxation factors that makes the algorithm converge is significantly broadened and the importance of this parameter on the global performance is reduced. The computing time also decreases by a factor between 10 and 20. Furthermore, this gain does not strongly depend on the exact imposed maximum number of iterations. Some preliminary observations are also reported in respect with the application of such a technique to the Anderson acceleration method. [ABSTRACT FROM AUTHOR]

The ANSWERS® WIMS reactor physics code is being developed for whole core multiphysics modelling. The established neutronics capability for lattice calculations has recently been extended to be suitable for whole core modelling of Small Modular Reactors (SMRs). A whole core transport, SP3 or diffusion flux solution is combined with fuel assembly resonance shielding and pin-by-pin differential depletion. An integrated thermal hydraulic solver permits differential temperature and density variations to feedback to the neutronics calculation. This paper presents new methodology developed in WIMS to couple the core neutronics to the integrated core thermal hydraulics solver. Two coupling routes are presented and compared using a challenging PWR SMR benchmark. The first route, called GEOM, dynamically calculates the resonance shielding and homogenisation with the whole core flux solution. The second coupling route, called CAMELOT, separates the resonance shielding and pincell homogenisation from the whole core solution via generating tabulated cross sections. Both routes can use the MERLIN homogenised pin-by-pin whole core flux solver and couple to the same integrated thermal hydraulic solver, called ARTHUR. Heterogeneous differences between the neutronics and thermal hydraulics are mapped via thermal identifiers for neutronics materials and thermal regions. The ability for the integrated thermal hydraulic solver to call an external code via a Fortran-C-Python (FCP) interface is also summarised. This flexible external coupling permits one way coupling to an external fuel performance code or two way coupling to an external thermal hydraulic code. [ABSTRACT FROM AUTHOR]

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

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. [ABSTRACT FROM AUTHOR]