Sodium-cooled Fast Neutron Reactors Research Articles

Simulation Analysis Based on Sodium-Cooled Fast Reactor Fuel Assembly Under Blockage Accident

This paper employs computational fluid dynamics to investigate the fuel assemblies within a sodium-cooled fast neutron reactor. To begin with, we developed computational models for seven wire spacer fuel rods under both normal operating conditions and transient blockage conditions. We conducted separate analyses to assess the impacts of normal operation, blockage thickness, and blockage area. This work allowed us to acquire data on the heat transfer properties and the flow field variations for the coolant, blockages, and fuel rods across different conditions. Subsequently, we leveraged these flow field alterations to examine the resulting temperature distributions. Analyses were conducted to evaluate the effects of normal operation, blockage thickness, and blockage area. The research acquired the heat transfer characteristics and flow field distribution variations of the coolant, blockages, and fuel rods under different operational conditions and utilized these variations to analyze the temperature distribution. Through research analysis, the following conclusions have been reached. Under normal operating conditions, the temperature and flow fields of the fuel components exhibit cyclic variations along the axial length, corresponding to the pitch of the wire spacers. Heat exchange between the internal and external subchannels occurs independently, which further substantiates that the incorporation of wire spacers strengthens lateral flow disturbances. This effect is more pronounced within the internal subchannels, thereby leading to a marked difference in the flow fields on either side of the wire spacers. In the case of blocked conditions, an increase in blockage thickness and area both lead to higher temperatures for the coolant, the blockages themselves, and the fuel rods. The temperature in the recirculation zone behind the blockages also rises with increasing blockage thickness and area, although the magnitude of this increase is not significant. After the onset of blockage conditions, the instantaneous temperature tends to increase. As time progresses, the instantaneous temperature fields following the blockage do not display greater fluctuations in temperature change than those observed after the system has stabilized. For the temperature parameters of the convective field, differences arising from an increase in blockage area are more significant than those caused by an increase in blockage thickness. When blockage area and blockage thickness are increased by the same multiple, an increase in blockage area results in a higher temperature peak. Increasing the area of blockage necessitates a longer duration for both the temperature and the velocity fields to revert to equilibrium.

Applicability to DEMO breeding blankets of neutron measurements techniques from fission reactors

Neutron measurements are routinely performed in fission reactors. A lot of experience has been gained at CEA in designing, prototyping and testing neutron sensors and associated acquisition systems in several nuclear facilities with a wide range of demanding conditions. In this paper, the applicability of this experience to neutron diagnostics in the breeding blankets of DEMO is addressed, as a complement to the already planned diagnostic strategy.The breeding blankets are a key component of a fusion reactor such as DEMO. The expected performances and operational constraints are demanding, nevertheless some of those, as the neutron flux range and high temperature, are close to what is encountered in sodium-cooled fast neutron reactors.The strengths and drawbacks of neutron sensors are presented with respect to the performances and constraints. Self-powered neutron detectors and fission chambers benefit from a vast return of experiment in fission reactors and received recent developments in the framework of studies for sodium-cooled fast neutron reactors. The innovative optical fission chamber is discussed for its electromagnetic immunity and high dynamic range demonstrated up to seven decades. Dosimetry with activation foils is also considered for its ability to provide an absolute calibration and spectral unfolding.

Economics of new nuclear power plants – assessment of investments into Generation III, small modular reactors and non-light-water reactors

Abstract. Given the need to combat climate change, as recently stressed again within the IPCC (2023) report, and the recent energy price increases for electricity and natural gas through the war in Ukraine (ECB, 2022), investments into new nuclear power plants are a considered option for future energy systems in some countries. In this paper, we discuss economic aspects of such investments, differentiating between three different types of reactor technology, as described in the following: i. At present, the only viable option for an investment would be “Generation III” reactors, i.e., light-water reactors with high capacities (in the range of or above 1000 MW). The most recent projects of that type have been very expensive, though, and there is a controversy about whether future ones will become competitive (Wealer et al., 2021; Duan et al., 2022). Economic questions relate to economies of scale and the differences in costs between western reactors (USA, Europe) and those in Russia and China.ii. In some countries, the development of and subsequent investments in light-water reactors of small power rating (<300 MW) are pursued (e.g., in the US, Canada and the UK). These are sometimes called “small modular reactors” in the recent literature (Chu, 2010; IAEA, 2022). These concepts are surrounded by high uncertainty, and the paper proposes a methodology for economic analysis, based on previous literature (Rothwell, 2016; Roulstone et al., 2020; Boarin et al., 2021).iii. A third option for newly built reactors is represented by non-light-water reactors, amongst which the classical sodium-cooled fast neutron reactor (“fast breeder”) is the most advanced type as well as high-temperature reactors and molten-salt reactors. With the establishment of the GenIV International Forum in 2001, 14 member states, including the USA, China, Russia, the EURATOM states, and the United Kingdom, have joined forces with the shared objective of further developing non-light-water reactor concepts. The paper provides a methodology to assess the competitiveness of fast reactors and extends it to other non-light-water reactors. The paper concludes with an assessment of the economics of new nuclear power plants going forward. Particular consideration is given to the aspects of decommissioning from the very outset, i.e., the planning of the new reactor. In that context, the paper will address the interdependencies between technology choices and storage issues, for example volume composition.

Multi-scale structural investigation of uranium-plutonium mixed oxides (U1-yPuy)O2-x with high plutonium content

Uranium-Plutonium mixed oxides U1-yPuyO2-x are considered as the reference fuels for Sodium-cooled Fast neutron Reactors (SFRs). Nevertheless, there is a lack of experimental data on their thermal properties for a Pu content above 0.45. The first step to fill this gap is to manufacture samples with high plutonium contents, which is the purpose of this work. Here, the manufacturing of U1-yPuyO2 samples with y = 0.60, 0.65 and 0.70 by a co-milling powder metallurgy process is detailed. Relying on an optimization of each manufacturing step (blending, co-milling, sieving, pelletizing and sintering), in order to obtain dense, monophasic, stoichiometric and homogenous materials.A multi-scale characterization strategy was used to determine sample microstructural properties including density measurements, Thermal-Ionization Mass Spectrometer (TIMS), ceramography, Electron Probe Micro Analyses, µ-Raman Spectroscopy and X-Ray Diffraction. The results obtained show that dense (> 95% TD), homogenous (chemical distribution of cations and oxygen), monophasic and stoichiometric (O/M = 2.00) samples have been achieved. These properties correspond to those needed to optimize the thermo-physical properties measurements. This also shows that the coupling of multi-scale characterizations is mandatory to evaluate the structural and microstructural properties of this kind of material.

On developing a practical safety culture for the advanced reactor workforce: Experiences of working with sodium

The Versatile Test Reactor is meant to fill a need in the advanced nuclear reactor development community. Accelerated nuclear testing of fuels and materials is needed to allow designs to proceed. To accomplish this, a fast neutron spectrum is needed. To have a fast neutron spectrum, the reactor coolant needs to not only have excellent heat transfer properties to sufficiently cool the reactor core, but also not significantly slow neutrons down following their birth in the fission process. Sodium is an excellent coolant choice that can fulfill these needs and many others. Sodium is the sixth most abundant element in the Earth’s crust and can be acquired easily through simple chemical separation methods. Sodium, however, reacts when in contact with water and air, but with proper precautions, these scenarios can be avoided. Nuclear reactors that utilize molten sodium as a primary heat transfer medium [a.k.a. sodium-cooled fast-neutron reactors, or SFRs, Figure 1 (1)] provide unique challenges in their daily operation, and therefore require focused training for researchers, scientists, staff, and employees alike to understand these unique challenges. Therefore, a comprehensive understanding of sodium properties is needed by all personnel involved in the design, construction, operation, experiment planning and execution of SFRs. Additionally, several commercial advanced reactor designs are under development which will use sodium for the reactor coolant. These advanced reactor designs include the Natrium reactor from TerraPower and GE-Hitachi (2), the ARC-100 reactor from Advanced Reactor Concepts (3), and the Aurora reactor from Oklo (4).

MULTIPHYSICS MODELISATION OF AN UNPROTECTED LOSS OF FLOW TRANSIENT IN A SODIUM COOLED FAST REACTORS USING A NEUTRONIC-THERMAL-HYDRAULIC COUPLING SCHEME

Sodium cooled fast neutron reactors (SFR) are one of the selected reactor concepts in the framework of the Generation IV International Forum. In this concept, unprotected loss of cooling flow transients (ULOF), for which the non-triggering of backup systems is postulated, are regarded as potential initiators of core melting accidents. During an ULOF transient, spatial distributions of fuel, structure and sodium temperatures are affected by the core cooling flow decrease, which will modify the spatial and energy distribution of neutron in the core due to the spatial competition of neutron feedback effects. As no backup systems are triggered, sodium may reach its boiling temperature at some point, leading to local sodium density variations and making the transient fluctuate in a two-phase flow physics where thermal-hydraulics and neutronics may interact with each other. The transient phenomenology involves several physic disciplines at different time and spatial scales, such as core neutronics, coolant thermal-hydraulics and fuel thermo-mechanics. This paper presents the results of thermal-hydraulic/neutronic coupled simulations of an ULOF transient on the SFR project ASTRID. These coupled calculations are based on the supervisor platform SALOME to link the neutron code APOLLO3® to the system thermal-hydraulic code CATHARE3. The physical approach used by the coupling to describe the neutron kinetic is a quasi-static adiabatic one, updating the normalized spatial power distribution periodically by performing static neutron calculations, while a point kinetic model associated to a neutron feedback model calculates the power amplitude variations.

Conception of delayed neutron detector blocks in a sodium cooled fast reactor

The primary vessel of a sodium-cooled fast neutron reactor (SFR) must remain free of any contaminant released by a cladding failure in a fuel assembly. A comprehensive detection system of cladding failures has then to be considered right from the beginning of the reactor design. One sub-system, called delayed neutron detection (DND), aims to search for delayed neutron precursors (DNP) through an ex-vessel analysis of a sodium sample, which goes to a container, usually named a pot, surrounded by neutron detectors. This paper presents a methodology to optimize the hydraulic part of the design of the DND detection block. From a detection scenario that identifies the physical parameters in play, a computational model is derived, which couples fluid mechanics and neutron transport. Different types of pot designs are considered, for which the sodium flow velocity, DNP concentration and detector activities are assessed. A new type of pot with a coil shape is proposed and shown to have the better performances with respect to the detector response to delayed neutrons.

Characterization of high-temperature nuclear fuel-coolant interactions through X-ray visualization and image processing

High-resolution X-ray imaging was employed at the JAEA MELT facility to visualize a kilogram-scale interaction between a jet of high temperature molten stainless steel and sodium, mimicking a sodium-cooled fast-neutron reactor in the aftermath of a core-disruptive accident. A novel software, SPECTRA, has been developed for the quantitative characterization of jet quenching and fragmentation. Tracking and 3D reconstruction of the melt phase traversing the imaging window enabled the detection of 72% of the debris mass recovered post-experiment. The rebounding of melt fragments confirmed a solid outer crust at the melt-coolant interface, while a thermal fragmentation event induced rapid vapor expansion. Jet fragmentation is best explained by the vaporization of coolant entrained within the melt jet generating an internal over-pressure sufficient for fragmentation of the crust. Thermal fragmentation produced missiles with velocities exceeding that of the jet, and a bimodal debris size distribution of coarse jet shells and finer fragments.

A CFD-based simulation of fluid flow and heat transfer in the Intermediate Heat Exchanger of sodium-cooled fast reactor

A three-dimensional computational fluid dynamics (CFD)-based simulation for the Intermediate Heat Exchanger (IHX) of a pool-type sodium-cooled fast neutron reactor has been performed in this paper, which investigates the flow-field and temperature distributions of liquid sodium on both primary and secondary sides at full power conditions. Typically, some simplifications are applied to reduce the difficulty of computation and render the problem more tractable. A proposed method is that treat a number of tubes as porous media that replicates the pressure loss and heat transfer characteristics of the detailed model. The heat exchanger model of the ANSYS FLUENT code is used to analyze the heat transfer phenomena inside of the IHX and the resistance coefficients of porous media have been predicted by empirical relationships which vary with height and flow directions. Two separate models, with different size of inlet and outlet area for shell-side sodium, have been developed in this paper. By comparing the results from the different simulations, the model with larger inlet area for primary and non-uniform inlet velocity distribution of secondary sodium satisfies the design requirements with respect to the permissible upper limit of temperature difference and local radial velocity.

Sodium purification systems for NPP with fast reactors (retrospective and perspective views)

Coolant purification systems constitute integral element of any reactor facility. A number of new tasks emerge in the current day reactor engineering, one of which is the task to enhance productivity and capacity of such systems. Retrospective analysis of coolant purification systems in nuclear reactors from BR-5 to BN-1200, such as cold traps and sorbents for purification from cesium ensuring successful operation of nuclear power installations equipped with sodium-cooled fast neutron reactors and hot traps representing promising concept to be used for sodium purification from oxygen on NPPs demonstrated that there exist capabilities for enhancing characteristics of such systems. Main directions of efforts for improving characteristics of purification systems for future NPPs equipped with sodium-cooled reactors were suggested based on the results of the implemented studies. Cold trap must be included as the mandatory element of the purification system of the reactor primary cooling loop both in normal and in standby operational modes. Optimal solutions on the purification from radioactive impurities and non-radioactive corrosion products as applied to fast sodium-cooled reactors require implementation of purposeful R&D efforts.

Actinide Oxidation State and O/M Ratio in Hypostoichiometric Uranium-Plutonium-Americium U0.750Pu0.246Am0.004O2-x Mixed Oxides.

Innovative americium-bearing uranium-plutonium mixed oxides U1-yPuyO2-x are envisioned as nuclear fuel for sodium-cooled fast neutron reactors (SFRs). The oxygen-to-metal (O/M) ratio, directly related to the oxidation state of cations, affects many of the fuel properties. Thus, a thorough knowledge of its variation with the sintering conditions is essential. The aim of this work is to follow the oxidation state of uranium, plutonium, and americium, and so the O/M ratio, in U0.750Pu0.246Am0.004O2-x samples sintered for 4 h at 2023 K in various Ar + 5% H2 + z vpm H2O (z = ∼ 15, ∼ 90, and ∼ 200) gas mixtures. The O/M ratios were determined by gravimetry, XAS, and XRD and evidenced a partial oxidation of the samples at room temperature. Finally, by comparing XANES and EXAFS results to that of a previous study, we demonstrate that the presence of uranium does not influence the interactions between americium and plutonium and that the differences in the O/M ratio between the investigated conditions is controlled by the reduction of plutonium. We also discuss the role of the homogeneity of cation distribution, as determined by EPMA, on the mechanisms involved in the reduction process.

Effect of the fuel element bundle statistical characteristics on the evaluation of temperature in the sodium-cooled fast-neutron reactor core

Different fuel element bundle models used to calculate the coolant and fuel cladding temperatures inside fuel assemblies have been analyzed as applied to sodium-cooled fast-neutron reactors. The drawbacks of the existing models have been identified. A bundle model based on an experimental study into the actual arrangement of the fuel elements within the AF shroud has been proposed. The model's capabilities and advantages, as compared to conservative models, have been shown with regard for the need to raise the reliability of the fuel cladding working temperature estimation.

The ability of silicide coating to delay the catastrophic oxidation of vanadium under severe conditions

V–4Cr–4Ti vanadium alloy is a potential cladding material for sodium-cooled fast-neutron reactors (SFRs). However, its affinity for oxygen and the subsequent embrittlement that oxygen induces causes a need for an oxygen diffusion barrier, which can be obtained by manufacturing a multi-layered silicide coating. The present work aims to evaluate the effects of thermal cycling (using a cyclic oxidation device) and tensile and compressive stresses (using the three-point flexure test) on the coated alloy system. Tests were performed in air up to 1100°C, which is 200°C higher than the accidental temperature for SFR applications. The results showed that the VSi2 coating was able to protect the vanadium substrate from oxidation for more than 400 1-h cycles between 1100°C and room temperature. The severe bending applied to the coated alloy at 950°C using a load of 75MPa did not lead to specimen breakage. It can be suggested that the VSi2 coating has mechanical properties compatible with the V–4Cr–4Ti alloy for SFR applications.

Oxide dispersion strengthened ferritic steels: a basic research joint program in France

AREVA, CEA, CNRS, EDF and Mécachrome are funding a joint program of basic research on Oxide Dispersion Strengthened Steels (ODISSEE), in support to the development of oxide dispersion strengthened 9–14% Cr ferritic–martensitic steels for the fuel element cladding of future Sodium-cooled fast neutron reactors. The selected objectives and the results obtained so far will be presented concerning (i) physical–chemical characterisation of the nano-clusters as a function of ball-milling process, metallurgical conditions and irradiation, (ii) meso-scale understanding of failure mechanisms under dynamic loading and creep, and, (iii) kinetic modelling of nano-clusters nucleation and α/α′ unmixing.

Investigation of Nuclear Data Libraries with TRIPOLI-4 Monte Carlo Code for Sodium-cooled Fast Reactors

The Sodium-cooled fast neutron reactor ASTRID is currently under design and development in France. Traditional ECCO/ERANOS fast reactor code system used for ASTRID core design calculations relies on multi-group JEFF-3.1.1 data library. To gauge the use of ENDF/B-VII.0 and JEFF-3.1.1 nuclear data libraries in the fast reactor applications, two recent OECD/NEA computational benchmarks specified by Argonne National Laboratory were calculated. Using the continuous-energy TRIPOLI-4 Monte Carlo transport code, both ABR-1000 MWth MOX core and metallic (U-Pu) core were investigated. Under two different fast neutron spectra and two data libraries, ENDF/B-VII.0 and JEFF-3.1.1, reactivity impact studies were performed. Using JEFF-3.1.1 library under the BOEC (Beginning of equilibrium cycle) condition, high reactivity effects of 808 ± 17 pcm and 1208 ± 17 pcm were observed for ABR-1000 MOX core and metallic core respectively. To analyze the causes of these differences in reactivity, several TRIPOLI-4 runs using mixed data libraries feature allow us to identify the nuclides and the nuclear data accounting for the major part of the observed reactivity discrepancies.

In vessel detection of delayed neutron emitters from clad failure in sodium cooled nuclear reactors: An estimation of the signal

The detection of clad failures is mandatory in sodium-cooled fast neutron reactors in compliance with the “clean sodium” concept. An in-vessel detection system, sensitive to delayed neutrons from fission products released into the primary coolant by failures, partially tested in SUPERPHENIX, is foreseen in current SFR projects in order to reduce significantly the delay before an alarm is issued. In this paper, an estimation of the signal received by such a system in case of a failure is derived, taking the French project ASTRID as a working example. This failure induced signal is compared to that of the contribution of the neutrons from the core itself. The sensitivity of the system is defined in terms of minimal detectable surface of clad failure. Possible solutions to improve this sensitivity are discussed, involving either the sensor itself, or the hydraulic design of the vessel in the early stage of the reactor conception.

In vessel detection of delayed neutron emitters from clad failure in sodium cooled nuclear reactors: Information treatment

With appropriate techniques, the information brought by the in-vessel instrumentation of nuclear reactors may betray rather subtle departures from normal state indicating an abnormal situation at early stages, to improve both safety and availability. This paper takes the case of the detection of delayed neutrons released by a clad failure in sodium-cooled fast neutron reactors as an illustrative application of hypothesis testing. Two methods are discussed, seeking either for a counting excess of for a departure from stationarity, yielding similar results. The question of false alarms is addressed by a Bayesian approach that takes into account the prior probability of failure. The use of several identical sensors is shown to be effective.

Analysis of the experimental neutron noise from the PHENIX reactor

This paper deals with the interpretation of the neutron noise measurements in sodium-cooled fast neutron reactors and the problem of the core vibration monitoring. Given that the compaction of a SFR core results in a positive reactivity coefficient, monitoring any core movement is mandatory. Although several sodium-cooled reactors have been operated throughout the world, only the French PHENIX reactor matched the needs in terms of instrumentation and available data.This paper presents an analysis oriented towards the core compaction monitoring, of the measurements performed on the PHENIX reactor recorded in the SARA system.The main result is the observation of the neutron noise spectra as a function of power: as already proposed in the early years of PHENIX and SUPERPHENIX (SPX), the spectrum reflects the mechanical vibration of the fuel assembly lattice. The cross correlation with measurements such as vibration, sonar and temperature do not provide significant additional information to confirm or disprove this interpretation of the neutron noise spectrum: the temperature fluctuations lie in a distinct frequency domain, the sonar and vibration measurements on the control rods suffer from high detection noises.This paper also highlights that the interpretation of the noise measurements depends on the recording of the raw data, allowing data post analysis, development of new interpretative techniques, and feedback in terms of design of the instrumentation. As such, the SARA system is an example to reproduce, as far as the sustainability of the knowledge is concerned.

Laser absorption spectroscopy for xenon monitoring in the cover gas of sodium cooled fast reactors

Xenon is a fission product of uranium that must be monitored on line and in real time in the cover gas of a sodium cooled fast neutron reactor (SFR) for safety reasons. In this paper, tunable diode laser absorption spectroscopy is applied to quantify low concentrations of natural xenon in argon. Optical detection is carried out at 12 145 cm−1 (823 nm) from the 6s[3/2]02 Xe electronic metastable state, populated in a low pressure DC glow discharge. Optimal discharge parameters (110 Pa, 1.3 W) were experimentally determined to maximize the laser absorption signal. Measurement stability, argon pressure spectral broadening, isotope quantification, and linearity of the optical measurement with respect to the xenon concentration in the gas sample were carefully investigated. Under optimized experimental conditions, the detection limit of xenon over argon molar ratio is estimated as 60 ± 10 parts per billion at 1 Hz, corresponding to a xenon number density of 1.6 × 109 atoms.cm−3 along the laser beam.

Simulation of mixing different temperature jets using CABARET method

Mathematical modeling of mixing jets with different temperatures discharged from the fuel assemblies of a sodium-cooled fast-neutron reactor BN is presented. Numerical simulations are performed using the CABARET method. The aim is to verify the applicability of the CABARET method to the calculations of liquid flows with a small Prandtl number (~ 0.01) at relatively high Reynolds numbers (~105). For this purpose, the fluid flowing through the geometrically complex channel of the experimental setup has been investigated. Mesh decomposition and finite-difference grid generation techniques are described. The comparison between the numerical and experimental data shows that with the proposed approach the time-averaged local temperatures in the non-isothermal flow of liquid metal can be predicted with a sufficient accuracy, and a good qualitative description of the pulsation characteristics of the flow can be provided.