Sodium-cooled Fast Reactor Research Articles

Studies on In-situ regeneration of cold trap of a Bench-Top sodium loop

A cold trap (CT) is one of the crucial modules of the sodium coolant circuit of a sodium-cooled fast reactor as it controls hydrogen, oxygen and other impurities in liquid sodium within the threshold levels in dynamic mode. The major impurities are crystallized as NaH and Na2O in CT and get loaded during its continuous service. The CT at the secondary sodium circuit needs regeneration at regular intervals of time. In-situ regeneration of the CT during the reactor operation would be cost-effective while generating power. A bench-top sodium loop was employed to carry out the present study. The regeneration of CT was carried out by adopting a thermo-vacuum process in which the NaH decomposed into sodium and hydrogen. During the regeneration process, hydrogen release behavior as a function of time with varying temperature and cover gas pressure was investigated by deploying in-house developed online hydrogen sensors. An electrochemical hydrogen meter (ECHM) to measure hydrogen in liquid sodium in the ppb range, a semiconducting metal oxide-based hydrogen sensor (SMOHS) and a polymer electrolyte membrane-based hydrogen sensor (PEMHS) to measure hydrogen in the argon cover gas were used. Simultaneous measurement of hydrogen in both liquid sodium and cover gas phase played a central role in analyzing the hydrogen distribution phenomenon. The quantity of total hydrogen evolved and decomposed NaH was determined after the successful regeneration of the cold trap.

STEM and HRTEM study on matrix microstructure and oxide particles in 11Cr ferritic/martensitic ODS steel

Fe–Cr oxide dispersion strengthened (ODS) steel is one of the candidate structural materials for high burn-up fuel cladding of sodium-cooled fast reactor and fusion reactor blanket due to its superior high temperature strength and excellent resistances to corrosion, creep and irradiation. The matrix structure and oxides in 11Cr ferritic/martensitic (F/M) ODS steel (Fe–11Cr–0.13C–1.3W–0.4Ni–0.3Ti–0.35Y2O3) have been characterized by scanning transmission electron microscopy (STEM) and high resolution transmission electron microscopy (HRTEM). The matrix of 11Cr F/M ODS steel is composed of tempered martensite (major) and residual ferrite (minor). The dispersion morphology of oxides of residual ferrite is much better, relative to that of tempered martensite, because the oxides in the former have much smaller mean diameter and inter-particle spacing and, however, significantly higher number density and volume fraction. For tempered martensite and residual ferrite, phase identifications on 41 and 69 nanoparticles with peak number fraction and, consequently, most significant contributions to the macroscopic properties, were accomplished, respectively. The proportions of Y–Ti–O, Y–Cr–O, TiO2 and Y2O3 in tempered martensite are 58.5%, 12.2%, 19.5% and 9.8%, respectively. The proportions of Y–Ti–O, Y–Cr–O, TiO2 and Y2O3 in residual ferrite are 65.3%, 8.7%, 15.9% and 10.1%, respectively. There is no considerable difference in phases and their corresponding number fractions of nanoparticles between tempered martensite and residual ferrite. Coherency and crystallographic orientation relationship between the lattice of oxides and that of the residual ferrite matrix are ubiquitous and, moreover, such coherency and crystallographic correlation between the lattice of oxides and that of the tempered martensite matrix were only occasionally detected. The mechanisms of the formation and polymorphic transition of various kinds of oxides, the structure origins of the much better dispersion morphology of the nanoparticles of residual ferrite, relative to that of tempered martensite, and, moreover, the strengthening mechanisms in the 11Cr F/M ODS steel were discussed.

Numerical simulation of natural circulation characteristics under different DHX layout schemes in pool-type SFR during station blackout accident

Decay heat removal system (DHRS) is an essential system for pool-type sodium-cooled fast reactor (SFR) to ensure natural circulation and remove decay heat during station blackout (SBO) accident. Decay heat exchanger (DHX) is the key equipment in DHRS. In China Experimental Fast Reactor (CEFR), DHX layout scheme consisting of two DHXs in hot pool (DHX Layout 1) is adopted. Different DHX layout scheme is designed in the advanced pool-type SFR, for example DHXs are installed in both cold and hot pool (DHX Layout 2). It is necessary to analyze the influence of different DHX layout schemes. In this paper, two DHX layout schemes were proposed. Numerical simulation was carried out, of which method was preliminarily validated. The numerical results show that DHX Layout 2 can reach long-term cooling stage of natural convection earlier than DHX Layout 1 with strengthened flow rate, resulting in more uniform temperature distribution in the hot pool. There is a local asymmetrical temperature distribution in the cold pool for DHX Layout 2. These findings by numerical simulations provide important numerical reference for the subsequent DHRS design of the advanced pool-type SFR.

Resonance self-shielded multigroup atomic displacement cross section for assessing neutron irradiation damage

Accurate calculation of neutron-induced displacement damage is a key step for the subsequent investigations on neutron irradiation damage. Due to the use of multigroup displacement damage cross sections, the resonance self-shielding effect of neutron-induced nuclear reaction must be taken into account. In addition to the method proposed by the authors in 2019, two additional methods are proposed in the present work to consider the self-shielding effect on the calculation of neutron irradiation-induced displacement damage. The two newly proposed methods allow directly using self-shielded multigroup nuclear reaction cross sections without infinite-dilution cross sections, even though they are not as consistent as the first one in the case of weak resonance self-shielding effect. The resulting resonance self-shielding corrections on the displacement damage of 56Fe in a typical sodium-cooled fast reactor are quite close for the three different methods. The slight differences (within ∼ 2 %) are much smaller than the self-shielding corrections, which are about −11.3 % for the dominant neutron elastic scattering-induced damage and −10.4 % for the total displacement damage.

Modeling of sodium-cooled fast reactor core – A sensitivity study on number of core channels and axial meshes

System thermal–hydraulic codes are used in the dynamics analysis of sodium-cooled fast reactors (SFRs). In such simulations, the accuracy of the reactor core model is very important to establish plant safety. Coupled physical phenomena, namely, heat transfer, fluid dynamics, and neutronics, need to be simulated in the core. Due to the complexity, a simplified core modeling approach has been traditionally followed in system dynamics codes. Firstly, steady-state core neutronics calculations are carried out using a 2D axisymmetric model to obtain perturbation worth and power distributions. Later, this data is mapped to several core channels (usually 10 to 20), each representing a set of fuel subassemblies (SAs) for thermal-hydraulics calculations.In this paper, a more sophisticated core modeling approach has been followed to understand the importance of such an exercise. A 3D neutronics model was used to obtain subassembly-wise perturbation worths and power distributions. For thermal-hydraulics calculations, the number of core channels was increased with the idea of one subassembly per channel. The number of axial meshes in each channel was also increased. The traditional and new modeling approaches were used to simulate the benchmark problem of FFTF LOFWOS test#13 for comparison. Steady-state parameters from both the new and the old models were comparable. Notably, there was a peaking of the axial power profile in the new model compared to the old model, which resulted in slightly higher peak fuel temperatures in the new model. However, the difference in the evolution of transient parameters between the two models was insignificant, and both models compared reasonably well with the benchmark data.

General methodology for gas monitoring of cladding failure in sodium-cooled fast reactors

A gas detection system (DRG) is designed in the sodium-cooled fast reactor (SFR) to limit the release of fissile material into the coolant. The system monitors a fission gas signal in the cover gas resulting from a cladding failure. We developed a general methodology to describe the entire detection process, which can be applied to arbitrary SFRs with different fuel and core designs and contribute to the conception choice and optimization for future design. We defined a corresponding equation consisting of several modules, including fission gas quantity accumulated in the fuel gas plenum, the gas leakage flux through the crack, the transportation from the fuel pin to the detector, and the detector sensitivity to different isotopes. We presented a mathematical model to determine the fission gas activity in the detector as a function of the crack size on the cladding. The methodology is applied and modeled in the PHENIX reactor for illustration purposes. With the deduced methodology, we analyzed the entire system and discussed the future design of the DRG system.

Experimental study on interface characteristics of liquid sodium-water vapor interaction in SWR

When the sodium-water boundary in the sodium-cooled fast reactor (SFR) fails, water vapor will enter the liquid sodium and cause the sodium water reaction (SWR) accident. SWR may lead to many serious consequences for power plant. Thus, it is a non-negligible factor to be considered in the design and operation of SFR. In this study, experiments on the interaction between liquid sodium and water vapor were carried out. The reaction phenomena, temperature of liquid sodium, morphology of reaction product and other characteristics of the chemical were summarized and analyzed in this paper. The experiment was conducted under different initial sodium temperatures (300–450 °C) and different mass fraction of water vapor (0.1–0.9 at 200 °C) conditions. Visualization scheme was used in this experiment and the reaction process was recorded by a high-speed camera. The results indicate that both mass fraction of water vapor and initial temperature of liquid sodium are important factors in the reaction process. The mass fraction of water vapor affects the actual reaction rate. The initial temperature determines the form of reaction products (liquid NaOH) on the surface of liquid sodium. For the experiments with initial temperature below 400 °C, the reaction products appear as bubbles and foams. While the initial experiment temperature exceeds 400 °C (including 400 °C), bubbles and foams gone, replaced by a thin film. By analyzing the conditions of bubble generation, the surface tension of liquid NaOH is an important cause, which can be affected by the temperature and bounding water molecules. Furthermore, according to the measurement results of sodium temperature, the reaction products in the form of a large concentration of foams have a strong blocking effect on mass transfer and greatly slow down the actual reaction rate. This research has important reference significance for the study of the mechanism of liquid sodium and water vapor reaction.

On physics of a hypothetical core disruptive accident in Multipurpose hYbrid Research Reactor for High-tech Applications – MYRRHA

The sensitivity of the reactivity of a fast reactor core to changes in its geometry and/or fuel relocation calls for particular attention with regard to criticality events. A category of these events, the so-called Core Disruptive Accidents (CDAs), are intensively studied in the safety assessment of Sodium-cooled Fast Reactors (SFRs), and more recently also in the case of other systems. Differences between SFRs and Heavy Liquid Metal Fast Reactors (HLMFRs) are significant and therefore warrant an understanding of phenomena and the development of models specific to HLMFRs. This paper provides a qualitative overview of the physics relevant to the investigation of a CDA in HLMFR, with a particular application to the Multipurpose hYbrid Research Reactor for High-tech Applications – MYRRHA. At first, a core compaction mechanism viable for an HLMFR has been postulated. In what follows, simulation by an already existing severe accidents code, as well as modelling based on fundamental physics and engineering, have been performed. It is demonstrated that, for a linear insertion of reactivity due to hypothetical core compaction, the reversal of reactivity evolution happens due to the Doppler effect and the thermal expansion of core materials. Subsequent expansion by fuel melting terminates the prompt-critical event and makes the system delayed-supercritical. Successive fuel and/or coolant boiling is responsible for the hydrodynamic disassembly of the core and it therefore effectively terminates the transient.

Optimization and Improvement of Sodium Heated Once-through Steam Generator Transient Analysis Code Based on the JFNK Algorithm

The sodium heated once-through steam generator (OTSG) is a vital barrier separating sodium and the water loop in the sodium-cooled fast reactor (SFR). In view of the timeliness requirement of OTSG operation performance evaluation, Fortran95 programming language is used to optimize and improve the solution algorithm of the home-made transient analysis code, named TCOSS, for SFR sodium heated OTSG, which is modified into the JFNK algorithm for solving large sparse nonlinear matrices. It includes a matrix preconditioning module, a Krylov subspace formation module, a GMRES algorithm module and an inexact Newton iteration module. The correctness and efficiency of the algorithm model were verified using benchmarks such as the B1-B transient, ETEC shutdown experiments and seven-tube prototype experiments. The calculation speed was increased by more than four times.

Temperature dependence of O solubility in liquid Na by atomistic simulation of Na(l)-Na2O(s) interfaces using corrected machine learning potential: a step towards simulating Na combustion.

Liquid Na combustion is a significant safety concern in sodium-cooled fast reactors. Atomistic simulations are an alternative to experiments for studying detailed mechanisms of complex combustion processes. However, accurate simulations of the interfaces involved in combustion are challenging even for density functional theory (DFT), because the systematic error between different chemical systems cannot be fully cancelled. Herein, we report the achievement of a key milestone in atomistic simulation of liquid Na combustion, which involves the development of a machine learning (ML) moment tensor potential that allows accurate simulation of interface systems between liquid Na and solid Na2O. The ML potential is trained by using supervised and active learning to ensure DFT-level accuracy. An empirical correction is then applied to achieve experimental accuracy by reducing systematic error. Consequently, the basic properties of liquid Na and solid Na2O are accurately simulated. In addition, with empirical correction, experimental O solubility data for liquid Na at 350-900 K are reproduced by using interface molecular dynamics simulations and a thermodynamic model. The temperature dependence of the enthalpy and entropy of the Na2O solution and their effect on O solubility are evaluated. The results show that, despite the increase in solution enthalpy with temperature, O solubility increases more rapidly than the linear Arrhenius plot due to the effect of solution entropy. The results of this study indicate that, with appropriate correction, ML potentials can achieve near-experimental accuracy, beyond the accuracy of DFT, in interface simulations and material properties calculations, paving the way for sodium combustion simulations in the future.

Structural analysis of a reactor vessel in a sodium-cooled fast reactor under extremely high temperature conditions

Structural analysis of a reactor vessel in a sodium-cooled fast reactor under extremely high temperature conditions

A New Proposal Generalized Predictive Control Algorithm With Polynomial Reference Tracking Applied for Sodium Fast Reactors

This paper proposes a generalized predictive control (GPC) with constraints and orthonormal Laguerre functions using the simplified model of the primary system (reactor core and intermediate heat exchanger (IHX)) of a prototypical sodium fast reactor (SFR). This paper develops a multiple-input multiple-output (MIMO) GPC with input constraints able to track polynomial references of any degree applied in coolant temperature difference across the core and fractional power. The manipulated variables of the GPC-SFR are the reactivity and the sodium flow rate of the primary and secondary pipes. Moreover, orthonormal Laguerre functions and step down condition number techniques were also applied to avoid the numerical ill-conditioning issue in quadratic programming of large systems. Thus, a GPC type-2 was designed to control fractional power, coolant temperature difference across the core and sodium tank temperature of the SFR primary system when temperature references change according to a linear ramp after reaching their steady-state operation, sustaining 100% power operation on the reactor. In order to analyze the load tracking capability of the GPC-SFR type-2, the load following from 100% fractional power (FP) to 60% FP at 0.8% FP/min rate is simulated. Constraints on the rate of coolant temperature difference across the core and reactivity were applied for the design safety. For comparison criteria, this paper compares the GPC-SFR type-2 with the GPC-SFR type-1, i.e, standard model predictive control (MPC), to verify the viability and superior performance of the proposal regarding: (a) ramp-tracking capability of temperature and load; (b) the rejections of a reactivity disturbance of -1 cent and a secondary sodium inlet temperature disturbance of +10°F; and (c) a simulation with uncertainty in reactor design. The simulations show that the GPC-SFR type-2 overcome the GPC-SFR type-1 robustness and performance.

Review of Euratom projects on design, safety assessment, R&D and licensing for ESNII/Gen-IV reactor systems

Five Euratom projects launched since 2017 in support of the development of ESNII/Generation-IV reactor systems are briefly presented in the paper in terms of key objectives, results, and recommendations for the future. These projects focus on various aspects of the following ESNII/Generation-IV systems: Sodium Fast Reactor, Gas Cooled Fast Reactor, Supercritical Water Cooled Reactor, and Molten Salt Fast Reactor. The paper does not consider EU projects focused on the Gen-IV reactor technologies based on the use of heavy metals as a coolant because these projects are reviewed in a different paper.

Thermodynamic Stability of Volatile Radionuclides in Cladding Failure Accident of Sodium-cooled Fast Reactor

Thermodynamic Stability of Volatile Radionuclides in Cladding Failure Accident of Sodium-cooled Fast Reactor

Development of probabilistic risk assessment methodology against strong wind for sodium-cooled fast reactors

Development of probabilistic risk assessment methodology against strong wind for sodium-cooled fast reactors

Analysis on heat transfer performance sub-channels of sodium-cooled fast reactor fuel assemblies based on entransy

In this paper, the symmetric heat transfer performance of sodium-cooled fast reactor fuel assemblies was analyzed and studied. The model is analytically optimized based on sub-channel calculations. The deviations of the numerical simulation results from the pre-existing experimental data in the literature are within 10 %, with an average deviation of 2.5 %, which tested the reliability of the model. The calculated results demonstrated that the distribution of the axial power, temperature, and coolant of the reactor core is approximately symmetric M-shape. The reactor core coolant has a monotonic increase in axial distribution with the cladding temperature and the temperature peaks all appear at the reactor core outlet. The individual fuel assemblies' internal temperature is relatively sensitive to the axial power distribution, and there are troughs around the imports and exports. The simulated results showed that the center temperature of the hottest rod reactor core block reached 965.65 K. This pa- per provides a better guide to understanding the overall heat transfer effect by optimizing the heat transfer model.

Experimental study on boiling two-phase of liquid sodium along a 7-rod bundle – Part Ⅱ: Heat transfer characteristics

Sodium boiling will occur in some hypothetical accident cases (such as additional reactivity introduction and heat discharge capacity deterioration, and Local fault propagation). Thus sodium boiling experiments are necessary for the safety analysis and the development of serious accident programs in sodium-cooled fast reactors. In the present study, the boiling two-phase heat transfer experiments of liquid sodium were carried out on the boiling liquid sodium test loop. The experimental results showed that the entire boiling process can be divided into four stages. The heat transfer coefficient of the boiling two phases in the rod bundle channel increases gradually with the increase of heat flux and grows very slowly with the increase of system pressure. Based on the 54 groups of rod channels obtained, the new correlation of transfer heat coefficient of boiling two-phase liquid sodium was developed in a 7-rod bundle. The new correlation can well predict the experimental data of boiling two-phase liquid sodium obtained by other scholars, and the prediction error is within ±50\\%. Finally, mean relative deviation (MRD), absolute mean relative deviation (MARD), and root mean square deviation (RMSRD) were introduced to evaluate the accuracy of the correlation quantitatively.

Experimental study on boiling two-phase of liquid sodium along a 7-rod bundle – Part Ⅰ: Pressure drop characteristics

The two-phase pressure drop characteristics of liquid sodium are very significant for the safety and serious incident program development analysis of sodium-cooled fast reactor (SFR). In this paper, the boiling two-phase pressure drop experiments of liquid sodium were conducted in a 7-rod bundle with the mass flux range of 80 ∼ 390 kg/(m2·s), heat flux up to 120 kW/m2 and the absolute pressure range of 7.5 ∼ 100 kPa. The experimental results show that the frictional multiplier of boiling two-phase liquid sodium in the rod bundle channel decreases with the increase ofXLM. Some existing two-phase pressure drop correlations in the literatures were assessed and compared with the experimental data obtained in the present study. Results showed that these correlations could not predict the current experiments well because of the different geometries and measurements of high temperature environment. The new correlation for the two-phase friction multiplier of liquid sodium was developed in the 7 rod bundle. The predicted value of the new correlation was within 30 % of the experimental data in the rod bundle, which sufficiently demonstrates that the new correlation developed in this paper can well predict the boiling two-phase frictional pressure drop of liquid sodium in the rod bundle.

Numerical studies of power oscillations induced by sodium boiling in ESFR after ULOF transient

This paper investigates numerically a new phenomenon of power oscillations, which is found to take place after unprotected loss of flow (ULOF) in a large power European Sodium Fast Reactor (ESFR) with a sodium plenum above the core.The oscillations are induced by sodium boiling and re-flooding events and can trigger a power excursion, if the oscillation amplitude becomes stronger and the reactivity exceeds the prompt criticality during transients. Numerical results show that after the sodium boiling onset in and in the vicinity of the sodium plenum, a negative reactivity effect is introduced, that reduces the power, and the sodium boiling is stopped. When liquid sodium re-floods into the voided region, introducing a positive reactivity effect, the power increases, and then sodium boils again. This process repeats, i.e. power oscillations take place. A long numerical simulation shows that the oscillations may disappear as the cover gas pressure increases up to a certain value during the transient. This gives us an idea that an increased cover gas pressure could prevent the oscillation for the considered case, simply because the sodium boiling is suppressed. An additional transient simulation with a higher initial cover gas pressure confirms this idea.

Numerical simulation of operational transients in sodium-cooled fast reactors

Reactor startup and shutdown are planned plant transients that are expected numerous times in the life of a nuclear reactor. These transients are desired to be carried out quickly for economic benefits and at the same time safely, which demands slower execution. The various plant design and safety constraints to be met during the execution of these transients in a typical medium-sized sodium-cooled fast reactor (SFR) were discussed. Operational procedures for the execution of these transients were reported in detail. A “wait and raise” mode for the startup and a “wait and lower” mode for the shutdown were adopted. During these procedures, primary and secondary loop flow rates were decided to be maintained at a nominal 100 % flow to avoid thermal stratification issues in sodium pools. Feedwater flow will be controlled based on sodium temperature at the steam generator (SG) outlet. The dynamics analysis for a reference SFR design was then carried out using plant dynamics code DYANA-P. The modeling details and approximations for simulating the startup and shutdown transients using the DYANA-P code were discussed. Two operational parameters, namely, control rod movement distance and waiting time between control rod movements, were optimized as 2.3 mm and 2.75 min, respectively, for the reactor's beginning of equilibrium cycle (BOEC) core configuration. The startup and shutdown durations were both estimated as ∼20.5 h each. The hot pool temperature change rate was estimated as ∼19 °C/h, which is less than the allowable limit of 20 °C/h. The SCRAM parameters, namely, reactor period and reactivity, were less than their alarm limits, thus, not interfering with the planned operations. The design and operational constraints of various plant systems were also satisfied. Parametric studies were also carried out for a few parameters: the fuel-clad interaction and the core burnup conditions. The assumptions of fuel being stuck to the clad (in feedback reactivity calculation) and the BOEC core configuration gives a higher power raising rate for a given control rod stroke and waiting time.