SFR Core Research Articles

Nodal method for handling irregularly deformed geometries in hexagonal lattice cores

The hexagonal nodal code RENUS has been enhanced to handle irregularly deformed hexagonal assemblies. The underlying RENUS methods involving triangle-based polynomial expansion nodal (T-PEN) and corner point balance (CPB) were extended in a way to use line and surface integrals of polynomials in a deformed hexagonal geometry. The nodal calculation is accelerated by the coarse mesh finite difference (CMFD) formulation extended to unstructured geometry. The accuracy of the unstructured nodal solution was evaluated for a group of 2D SFR core problems in which the assembly corner points are arbitrarily displaced. The RENUS results for the change in nuclear characteristics resulting from fuel deformation were compared with those of the reference McCARD Monte Carlo code. It turned out that the two solutions agree within 18 pcm in reactivity change and 0.46% in assembly power distribution change. These results demonstrate that the proposed unstructured nodal method can accurately model heterogeneous thermal expansion in hexagonal fueled cores.

Corium pool behaviour in the core catcher of a sodium-cooled fast reactor

Within the framework of the severe accident mitigation strategy for GEN-IV reactors, the behaviour of molten corium in the SFR core catcher is a subject of significant interest. This article aims to define the main considerations within the core catcher during a prospective severe accident scenario, in particular with regard to the rate of heat transfer to the core catcher structures. On the basis of the core catcher design adopted at the end of the ASTRID project, a simplified model is proposed to conservatively evaluate the quasi-steady state behaviour in a stratified oxide-metal molten corium pool with mechanically stable crusts at the interfaces between the stratified layers and at the core catcher walls. The corium oxide layer is subject to significant internal heat generation due to the radioactive decay heat. The proposed model implies that the core catcher should withstand reasonable worst case estimates for the thermal load from the corium pool, avoiding sodium boiling and thermal ablation of the core catcher structures. Thermal ablation is only predicted in the cases of an order of magnitude under-prediction of the heat transfer coefficients at the lateral boundary of the corium oxide layer, or for local heat flux values in excess of 3.7 times the surface average. Further investigations will be dedicated to the revision of the more conservative model assumptions, in particular regarding the magnitude of the heat flux from high aspect ratio encloses with large internal Rayleigh numbers. A refined transient version of the model is intended for implementation within the PROCOR platform developed by the CEA.

Modelling of a passive hydraulic lock concept for the mitigation transfer tubes in a SFR core

To improve the safety of the integrated Sodium-cooled Fast Reactor, recent safety complementary devices can prevent or mitigate the consequences of a hypothetical severe accident consecutive to a loss of reactor cooling sequence. In particular, the M−TT device is set by empty transfer tubes, in place of some fuel sub-assemblies present in the reactor core, using an innovative passive hydraulic lock. It is based on the concept of hydraulic diode controlled by the flowrate of primary pumps. This system is dedicated to provide an easy way to discharge corium towards the core catcher in the hypothetical case of severe accident. In bonus, concerning the prevention of severe accident, it opens a direct flow path between the hot pool, equipped with decay heat removal exchangers, and the cold pool. This can lead to an easier passage of colder sodium towards the cold pool to promote core cooling by natural convection in case of the loss of the primary flow. This paper is devoted to the evaluation of the assessment of the hydraulic lock function and of the small magnitudes of the perturbations induced by our device on the thermal–hydraulic behavior of the reactor during normal operations. Through an analytical analysis and an up-scaling numerical approach, ranging from the local scale (CFD) to the system one, we claim that the lock function is kept during the normal situations at various power regimes. The leak flow represents no more than 1% of the primary pump flow. In addition, no important reactor thermal–hydraulic perturbation is brought by adding transfer tubes in the core. This is true for normal steady states, whatever the power is, but also for a full-power loss of forced flow accidental transient leading to the natural convection cooling.

Neutronic Analysis of Start-Up Tests at China Experimental Fast Reactor

The China Experimental Fast Reactor (CEFR) is a small, sodium-cooled fast reactor with 20 MW(e) of power. Start-up tests of the CEFR were performed from 2010 to 2011. The China Institute of Atomic Energy made some of the neutronics start-up-test data available to the International Atomic Energy Agency (IAEA) as part of an international neutronics benchmarking exercise by distributing the experimental data to interested organizations from the member states of the IAEA. This benchmarking aims to validate and verify the physical models and neutronics simulation codes with the help of the recorded experimental data. The six start-up tests include evaluating criticality, control-rod worth, reactivity effects, and neutron spectral characteristics. As part of this coordinated research, the IAEA performed neutronics calculations using the Monte Carlo codes Serpent 2 and OpenMC, which can minimize modeling assumptions and produce reference solutions for code verification. Both codes model a three-dimensional heterogeneous core with an ENDF/B-VII.1 cross-section library. This study presents the calculation results with a well-estimated criticality and a reasonably good estimation of reactivities. The description and analysis of the core modeling assumptions, challenges in modeling a dense SFR core, results of the first phase of this project, and comparative analysis with measurements are presented.

Similarity Studies of Buoyancy Effects in Impinging Jets – Application to Sfr Core

Similarity Studies of Buoyancy Effects in Impinging Jets – Application to Sfr Core

A Monte Carlo study on burnup treatment in sodium-cooled reactor with Th fuel

Abstract Sodium Cooled Reactors is one of the Generation-IV plants selected to manage the long-lived minor actinides and to transmute the long-life radioactive elements. This study presents the comparison between two-designed SFR cores with 600 and 800 MWth total heating power. We have analyzed a conceptual core design and nuclear characteristic of SFR. Monte Carlo depletion calculations have been performed to investigate essential characteristics of the SFR core. The core calculations were performed by using the Serpent Monte Carlo code for determining the burnup behavior of the SFR, the power distribution and the effective multiplication factor. The neutronic and burn-up calculations were done by means of Serpent-2 Code with the ENDF-7 cross-sections library. Sodium Cooled Fast Reactor core was taken as the reference core for Th-232 burnup calculations. The results showed that SFR is an important option to deplete the minor actinides as well as for transmutation from Th-232 to U-233.

Evaluation of sodium boiling models using KNS-37 LOF experiments

Abstract The computational codes used in the evaluation of the ESFR-SMART reactor performance and specifically their sodium boiling models are assessed using two KNS-37 LOF experiments, i.e. L22 and L29 tests, where boiling onset and two-phase flow regime up to dry-out occurred. The well-equipped KNS-37 experimental facility provided very valuable information for understanding the physical phenomena occurring in a 37-pin subassembly under LOF conditions, as well as experimental data to be used for computational tools validation. NATOF-2D, SAS-SFR, TRACE, ASTEC-Na, CATHARE-2 & CATHARE-3 and NEPTUNE_CFD codes have been used in this exercise in order to compare the various boiling models and conclude on the advantages and limitations of them based on the comparison against the experimental data. Beyond boiling onset, the various sodium two-phase flow approaches determine the ability of the code to correctly represent the rewetting and voiding phases as well as cladding dry-out onset. A simulation performed by a CFD approach (NEPTUNE_CFD code) considering liquid-vapor interfaces by an interface-tracking method is also shown and compared with the others approaches. Conclusions on each code performance are presented where the improvements needed to solve the issues encountered are included. This paper provides a first step in the process to investigate the required evaluation of the sodium two-phase flow models able to assess the safety of new SFR core designs (e.g. low void cores) under accidental conditions such as ULOF transients.

NEUTRONIC MODELING OF A FAST CORE WITH MODERATING MATERIALS USING APOLLO3R CODE

In order to improve passive safety of Sodium-cooled Fast Reactors The French Alternative Energies and Atomic Energy Commission (CEA) has proposed a new core design called CADOR - an SFR core with enhanced Doppler reactivity feedback. One of its most important design features is the introduction of solid moderating materials inside each fuel assemblies to slightly decrease the average neutron energy. The article focuses on development and validation of a neutronics calculation scheme able to produce accurate results in case of CADOR and other fast cores with moderating materials. The study uses two different fuel assembly models moderated by metallic beryllium and zirconium hydride (ZrH2) respectively The study includes discussion of neutron scattering treatment and different ways of spatial homogenization and energy condensations. The results indicate that the accurate scattering treatment leads to much better estimation of Doppler constant, especially in case of ZrH2 moderated core. By using combined deterministic-Monte Carlo calculation scheme we are able to quantify the biases on global reactivity, reactivity feedbacks and control rod worth. We demonstrate that spatial homogenization plays a more important role in case of moderated CADOR assemblies and thus preserving certain level of heterogeneity within fuel assemblies can lower the calculation bias significantly.

NEUTRONIC MODELLING OF THE FFTF CONTROL ROD WORTH MEASUREMENTS WITH DIFFUSION CODES

This paper presents an assessment of three deterministic core simulators with the focus on the neutronic performance in steady-state calculations of small Sodium cooled Fast Reactor cores. The selected codes are DYN3D, PARCS and the novel multi-physics solver GeN-Foam. By using these codes, the multi-group diffusion solutions are obtained for the selected twenty control rod worth measurements performed during the isothermal physics tests of the Fast Flux Test Facility (FFTF). The identical set of homogenized few-group cross sections applied in the calculations is generated with the Serpent Monte Carlo code. The numerical results are compared with each other as well as with the measured values. The obtained numerical results, such as the multiplication factors and control rod worth values, are in good agreement as compared to the experimental data. Furthermore, a comparison of the radial power distributions is presented between DYN3D, PARCS and GeN-Foam. Ultimately, the power distributions are compared to the full core Serpent solution, demonstrating an adequate performance of the selected deterministic tools. In overall, this study presents a verification and validation of the neutronic solvers applied by DYN3D, PARCS and GeN-Foam to steady-state calculations of SFR cores.

EVALUATION OF THE ESFR END OF CYCLE STATE AND DETAILED ANALYSIS OF SPATIAL DISTRIBUTIONS OF REACTIVITY COEFFICIENTS

The ESFR-SMART project is the latest iteration of research into the behaviour of a commercial-size SFR core throughout its lifetime. As part of this project the ESFR core has been modelled by a range of different reactor physics simulation codes at its end of cycle state, and the important safety relevant parameters evaluated. These parameters are found to agree well between the different codes, giving good confidence in the results. A detailed mapping of the local sodium void worth is also performed due to the problems associated with the positive void coefficient seen in large SFR designs. The local void worth maps show that the use of zone-wise coefficients replicates the important reactivity feedbacks to a high degree, indicating their suitability for use in SFR simulations.

FAST (floating absorber for safety at transient) for the improved safety of sodium-cooled burner fast reactors

This paper presents floating absorber for safety at transient (FAST) which is a passive safety device for sodium-cooled fast reactors with a positive coolant temperature coefficient. Working principle of the FAST makes it possible to insert negative reactivity passively in case of temperature rise or voiding of coolant. Behaviors of the FAST in conventional oxide fuel-loaded and metallic fuel-loaded SFRs are investigated assuming anticipated transients without scram (ATWS) scenarios. Unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS), unprotected transient overpower (UTOP) and unprotected chilled inlet temperature (UCIT) scenarios are simulated at end of life (EOL) conditions of the oxide and the metallic SFR cores, and performance of the FAST to improve the reactor safety is analyzed in terms of reactivity feedback components, reactor power and maximum temperatures of fuel and coolant. It is shown that FAST is able to improve the safety margin of conventional burner-type SFRs during ULOF, ULOHS, UTOP and UCIT.

WITHDRAWN: Modelling of a passive hydraulic lock concept for the Mitigation Transfer Tubes in a SFR core

WITHDRAWN: Modelling of a passive hydraulic lock concept for the Mitigation Transfer Tubes in a SFR core

Analysis of the methods for group constants generation for calculation of a large SFR core using Serpent 2 and CriMR codes

This work aimed at generating homogenized group constants using the Serpent code and then using the CriMR diffusion code to model the large SFR OECD 3600 MWth MOX core. The results were compared with a full core reference Monte Carlo solution by Serpent. Reactivity feedback parameters were also considered. Generating the group constants from separate fuel assemblies allows for simultaneously carrying out calculations and then using the results as input in diffusion codes rather than waiting so long for a 3D full core Monte Carlo calculation to be completed. From the results of the integral parameters we see a close agreement in the calculation codes. The differences can be attributed to the errors that could arise from generating the constants from individual sub-assemblies. The differences in the underlying physics and approximations used in development of the codes could also be a factor. Another way the errors could be reduced is by checking to see that the sub-assembly configurations used in the non-multiplying zones are as close as possible to the real layout in a full 3D core.

Trends on major actinides from an integral data assimilation

The objective of this work is to revisit integral data assimilation for a better prediction of the characteristics of SFR cores. ICSBEP, IRPhE and MASURCA critical masses, PROFIL irradiation experiments and the FCA-IX experimental programme (critical masses and spectral indices) with well-mastered experimental technique have been used. As calculations are performed without modelling errors (with as-built geometries) and without approximations with the TRIPOLI4 MC code, highly reliable C/E are achieved. Assimilation results suggest a 2.5% decrease for 238U capture from 3 keV to 60 keV, and a 4-5% decrease for 238U inelastic in the plateau region. For this energy range, uncertainties are respectively reduced to 1-2% and to 2-2.5% for 238U capture and 238U inelastic respectively. The increase trends on 239Pu capture cross section of around 3% in the [2 keV-100 keV] energy range come from a low PROFIL 240Pu/239Pu ratio C/E. For 240Pu capture cross section, the increase trend of around 4% in the [3 keV-100 keV] energy range goes in the same direction as the recent ENDF/B.VIII evaluation though at a much lower level. The nuclear data uncertainty associated to SFR ASTRID critical mass is reduced to 470 pcm.

Coupled thermalhydraulic-neutronic stability extended criterion in a SFR core

A coupled thermalhydraulic-neutronic extended criterion developed in order to assess the conditions of static instability of the sodium flow in a SFR core during the beginning of an unprotected loss of flow (ULOF) is presented in the first part of this paper. This extended criterion takes into account the evolution of the pressure drop in a sub-assembly (SA) as well as the evolution of its power when the reactor is affected by a flow rate (or power) perturbation by starting from a steady-state. The considered steady states are typical of quasi-steady states reached by a SFR during a ULOF. The temporal evolution of the flow and the temporal evolution of the power can be represented by a system of two differential equations whose linear stability analysis has enabled to define the aforementioned extended criterion.Then, in the second part of the paper, a verification of this extended criterion is performed by simulating the behavior of a GEN IV SFR of 1500 MWth. Four various steady states of the reactor have been investigated with a simulation tool devoted to unprotected loss of flow simulations (MACARENa). First simulations have enabled to calculate the various physical parameters needed to check the extended criterion. Perturbations have been applied to these four states. The extended criterion, proposed in this paper, has predicted flow stabilization for 3 states whereas it has predicted a flow excursion (sudden flow rate decrease leading to SA dry-out) for the last state. In this latter case, the classical Ledinegg criterion, which relies only on pressure drop evolution at constant power, has predicted a stable configuration. Then, the extended criterion predictions have been verified with MACARENa simulations, by slightly and temporary reducing the driving pressure head of the flow rate in the studied SA from each of the four initial states. For stable states, when setting again the pressure head of the flow to the value of the initial state, the initial flow rate is recovered. For the last initial state (predicted as unstable by the extended criterion), after a convergence period carried-out by imposing the SA steady flow rate, the solving of the transient momentum equation have shown that a flow excursion occurs in the MACARENa simulation as expected from the extended criterion. Consequently, the extended criterion proposed in this paper has been verified and could enable to define stable and unstable domain of couple flow rate/power in order to design future cores.

Verification of the multi-group diffusion code AZNHEX using the OECD/NEA UAM Sodium Fast Reactor Benchmark

AZNHEX is a novel 3D neutron diffusion code for nuclear core analysis with hexagonal-z geometry, and it is part of the Mexican project on the development of domestic nuclear analysis software “AZTLAN Platform”. Currently, the code is under development but some important steps have been made in the verification procedure of the code. A composite nodal method for prismatic hexagons has been developed and applied to solve the neutron diffusion equations for cores built by the union, side by side, of hexagonal prisms. A standard finite element method is used starting by the strong form to get the corresponding weak form.The verification and validation process of the AZNHEX code started with some academic exercises. In the case of this paper, in order to move to much more realistic scenarios, the OECD/NEA UAM Sodium Fast Reactor Benchmark has been used together with the SERPENT code for further verification. One of the exercises of the OECD/NEA Sodium Fast Reactor Benchmark is focused on the neutronic characterization of global parameters (keff, sodium void worth, Doppler, etc.) and feedback coefficient evaluation. This Benchmark is intended to confirm the ability of participants and their neutronic codes to provide generally consistent results when analyzing SFR core characteristics and thus, it represents a very good exercise for the AZNHEX development team.The differences in keff of AZNHEX versus the well validated deterministic codes DYN3D and PARCS are 70 pcm and 113 pcm respectively when using the same set of XS previously generated with the stochastic code SERPENT. Also, the radial power distribution compared between the deterministic codes over the main diagonal of the core presented very good agreement among them. In other exercises, differences in the order of 200 pcm, in the worst cases, were found when comparing integral parameters with SERPENT. The AZNHEX results give the confidence to keep developing the code in order to convert it into the standard domestic tool for hexagonal-z geometry core analysis. Future developments on the code will be focused on the implementation of discontinuity factors and the thermal expansion effects in an operating core as well as the time dependent implementation.

Validation of the DYN3D-Serpent code system for SFR cores using selected BFS experiments. Part II: DYN3D calculations

The capability of the DYN3D-Serpent codes system to simulate highly heterogeneous sodium-cooled fast reactor cores has been studied. The BFS-73-1 and the BFS-62-3A critical assemblies were chosen for the investigation. The study was performed in two parts. In the first part of the study, a 3D full model of each of the assemblies was simulated using the Serpent Monte-Carlo (MC) code, and the basic neutronic characteristics were evaluated and compared against experimental values. In the second part of the study, which is the subject of this paper, the assemblies were modeled using the DYN3D nodal diffusion code. The few-group cross-sections for the DYN3D analysis were generated using the Serpent MC code. The generation of effective few-group cross-sections of such assemblies is quite a challenge due to the substantial heterogeneity of the assemblies configuration. Therefore, the use of homogenization techniques was considered and evaluated. Initially, the GET and SPH techniques were applied for the analysis of the BFS-73-1 assembly core fuel rods, and of selected fuel rods from the BFS-62-3A assembly. Then, the SPH method was implemented and demonstrated for a pin-by-pin calculation of the BFS-73-1 assembly. It was shown that the GET and the SPH method noticeably improve the prediction accuracy of the DYN3D code. The results of the DYN3D pin-by-pin calculation with the SPH correction agree very well with that of the full assembly Serpent results, which in turn agree very well with the experimental data.

System codes benchmarking on a low sodium void effect SFR heterogeneous core under ULOF conditions

This paper discusses system codes benchmarking activities on an ASTRID-like heterogeneous fast core under a representative design basis accident condition: the unprotected loss of flow accident (ULOF). The paper provides evidence that all the system codes used in this exercise are capable to simulate the transient behavior of heterogeneous SFR cores up to the initiation of sodium boiling. As a proof of this, a comparison of steady-state results and dynamic simulation results for a ULOF transient (simulated using system codes in combination with neutron point kinetics) are provided and discussed in this paper. The paper contains a brief description of the system codes (TRACE, CATHARE, SIM-SFR, SAS-SFR, ATHLET, SPECTRA, SAS4A) used by the participants (PSI, CEA, EDF, KIT, GRS, UPVLC, NRG, KTH), assumptions made during the simulations, as well as results obtained.

New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses

Abstract The reactor dynamics code DYN3D being developed at the Helmholtz-Zentrum Dresden-Rossendorf is currently under extension for Sodium cooled Fast Reactor analyses. This paper provides an overview on the new version of DYN3D to be used for SFR core calculations. The current article shortly describes the newly implemented thermal mechanical models, which can account for thermal expansion effects of the reactor core. Furthermore, the methodology used in Sodium cooled Fast Reactor analyses to generate homogenized few-group cross sections is summarized. The conducted and planned verification and validation studies are briefly presented. Related publications containing more detailed descriptions are outlined for the completeness of this overview.

On the use of the SPH method in nodal diffusion analyses of SFR cores

A number of recent studies successfully demonstrated the feasibility of using Monte Carlo code Serpent to generate few-group cross sections (XS) for full core nodal diffusion analyses of SFR cores. The current study investigated the potential of the SPH method, applied to correct the few-group XS produced by Serpent, to further improve the accuracy of the nodal diffusion solutions. The procedure for the generation of SPH-corrected few-group XS is presented in the paper. The performance of the SPH method was tested on a large oxide SFR core from the OECD/NEA SFR benchmark. The reference SFR core was modeled with the DYN3D and PARCS nodal diffusion codes using the SPH-corrected few-group XS generated by Serpent. The nodal diffusion results obtained with and without SPH correction were compared to the reference full-core Serpent MC solution. It was demonstrated that the application of the SPH method improves the accuracy of the nodal diffusion solutions, particularly for the rodded core state.