Sodium-cooled Fast Reactor Research Articles

Validation of the Hybrid Turbulence Model in Detailed Thermal-Hydraulic Analysis Code SPIRAL for Fuel Assembly Using Sodium Experiments Data of 37-Pin Bundles

In the study of safety enhancements on advanced sodium-cooled fast reactors (SFRs) by the Japan Atomic Energy Agency (JAEA), it has been essential to clarify the thermal hydraulics under various operating conditions at high and low flow rate conditions in a fuel assembly (FA) with wire-wrapped fuel pins to assess the structural integrity of the fuel pin that achieves a high-performance core with high burnup ratio and high power density. A finite element thermal-hydraulic analysis code named SPIRAL has been developed by JAEA to analyze the detailed thermal-hydraulic phenomena in the FA of a SFR. In this study, numerical simulations of 37-pin bundle sodium experiments at different Reynolds (Re) number conditions, including a transitional condition between laminar and turbulent flows and turbulent flow conditions, were performed to validate the developed hybrid k-ε/kθ-εθ turbulence model equipped in SPIRAL to consider the low Re number effect near the wall in the flow and temperature fields. The temperature distributions predicted by SPIRAL were consistent with those measured in the sodium experiments at the Re number conditions. Through the validation study, the applicability of the hybrid turbulence model in SPIRAL to the thermal-hydraulic evaluation of sodium-cooled FAs in a wide range of Re numbers was confirmed.

Solar and sodium fast reactor-based integrated energy system developed with thermal energy storage and hydrogen

Hybridizing nuclear and renewable energy systems helps enhance the benefits of both sources and eliminate their disadvantages. In order to provide resilience, energy systems need to be designed in a complementary manner to achieve an uninterrupted energy supply. The current study proposes a hybridization of a sodium fast reactor with a concentrated solar plant and molten salt energy storage system. By considering the community requirements, additional subsystems are added that use process heat and power to generate more useful commodities. The proposed nuclear and renewable hybrid energy system generates heat, power, hydrogen, fresh water, and cooling effect for communities in order to meet their needs in a sustainable fashion. The proposed system is a potential alternative to currently available fossil fuel-driven systems. Intermittency and flexibility issues of such systems are overcome with hybridization and integration. Apart from the other commodities, hydrogen is produced and supplied to the community by considering hydrogenization along with the electrification of transportation. Specifically, a parabolic trough collector-type concentrated solar plant, a sodium-cooled fast reactor, a molten salt energy storage unit, a lithium bromide absorption refrigerator, a solid-oxide electrolyser, a multi-effect distillation-type desalination unit, and a district heating system are considered to be used in the integrated system. The proposed system is investigated by considering the first and second laws of thermodynamics, which assess energy and exergy inputs and outputs of the system and subsystems. In order to analyze the system with realistic conditions, actual data for California, the United States, is considered. For a community with 240243 residents, a sodium fast reactor with 1.5 GWth capacity and parabolic trough collectors with 0.5 GWth capacity are considered, along with a 4 GWh of molten salt energy storage system. Annual average energy and exergy efficiencies are found to be 63.54 % and 57.96 %. The maximum energy efficiency is calculated as 84.4 %, and the maximum exergy efficiency is found as 81.28 %. Apart from the current community needs, 53,285.15 tonnes of hydrogen is produced annually.

Impingement wastage experiment with SUS 316 in a printed circuit steam generator

The sodium cooled fast reactor (SFR) is one of the Gen-IV reactors with the most operating experience accumulated. Although the technology level is the most mature among the Gen-IV reactors, there is still a safety problem that has not been solved, which is the sodium-water reaction. Since sodium and water are separated only by a heat transfer tube with a thickness of only a few mm, there is inherently a risk of a sodium-water reaction (SWR) accident in the SFR. In this study, it is attempted to quantitatively evaluate the resistance of SWR accidents by replacing the shell and tube steam generator with printed circuit steam generator (PCSG) as a method to mitigate the SWR accident. To do this, a CATS-S (Compact Accident Tolerance Steam Generator-SWR) facility was designed and built. And for the quantitative evaluation of accident resistance, a methodology for measuring the impingement wastage rate was established. As a result of this research, the impingement wastage rate caused by SWR generated in a PCSG was measured first time. It was confirmed that the impingement wastage phenomenon was suppressed in the PCSG, and the accident resistance was higher than that of the SWR through comparison with the experimental results performed in the existing shell and tube steam generator. In conclusion, a PCSG is more resistant to impingement wastage as a result of the SWR accident than existing shell and tube steam generators, and it is estimated that a PCSG can mitigate SWR accidents, an inherent problem of SFR.

Characteristics of debris resulting from simulated molten fuel coolant interactions in SFRS

Sodium cooled Fast Reactors (SFR) are built with several engineered safety features and hence a severe accident such as a core melt accident is hypothetical with a probability of <10−6/ry. However, in case of such accidents, the mixture of the molten fuel and structural materials interacts with sodium. This phenomenon is known as Molten Fuel Coolant Interaction (MFCI) and results in fragmentation of the melt due to various instabilities. The fragmented particles settle as a debris bed on the core catcher at the bottom of the reactor vessel, and continue to generate decay heat. Characteristics of the debris particles play a vital role in heat transfer from the bed and need thorough investigation. The size, shape, and physical state of the debris depend on the associated fragmentation mechanism, superheating of the melt, and sodium temperature. Experiments have been conducted by releasing simulated corium, a molten mixture of alumina and iron generated by the aluminothermy process at ∼2400 °C into liquid sodium, to study the fragmentation phenomena. After the experiment, the fragmented debris was retrieved and the particle size distribution was determined by sieve analysis. The debris was subjected to microscopic investigation for obtaining morphological characteristics. Based on the characteristics of debris, an attempt has been made to assess of fragmentation mechanism of simulated corium in sodium.

Model for core catcher ablation and cavity shape in the film regime

During core meltdown in a nuclear reactor the corium may propagates toward the lower head and hit the vessel or structural elements as a coherent liquid jet. This could occur for instance in sodium fast reactors (SFR) if discharge tubes and an in-vessel core-catcher are used as mitigation devices. This can result in the ablation of the core-catcher and potential loss of its integrity. During ablation, a cavity is formed within the solid. Therefore the ablation of a solid structure by a liquid jet is studied here, with a focus on the cavity shape to improve future core-catcher designs. During the cavity formation, liquid flows over the cavity being formed as a liquid film in the film ablation regime, and ultimately results in the formation of a liquid pool. In the film ablation regime, two types of cavity shapes have been identified previously. To describe the process of the cavity formation, a mathematical model valid in this film ablation regime is here proposed which links the local melting rate to the local curvature of the cavity. It is applied with two sets of assumptions applicable to the two types of cavity previously identified. For the first assumption set, the heat transfer coefficient varies along the liquid/solid interface while the liquid temperature is held constant, it is representative of laminar boundary layer growth in laminar film. For the second assumption set, the heat transfer coefficient is held constant while the temperature varies along the liquid/solid interface, this is consistent with a turbulent liquid film. The two modeling approaches are compared with experimental data acquired from the HAnSoLO experimental setup. Numerical predictions show promising agreements with experimental observations, particularly for the laminar boundary layer growth zone of the cavity.

Validation of the fast reactor plant dynamics analysis code Super-COPD using FFTF loss of flow without scram test #13

The Japan Atomic Energy Agency (JAEA) has been developing a plant dynamics analysis code named Super-COPD for design studies and safety analyses of sodium-cooled fast reactors. Since there are a few validation studies of a thermal–hydraulic model in a plant, including a point kinetics model considering reactivity feedback for anticipated transient without scram (ATWS) events classified into the beyond design basis events (BDBEs), further validation studies under BDBE conditions are needed to improve the prediction accuracy of the model. Therefore, JAEA has participated in the international benchmark analysis of loss of flow without scram (LOFWOS) test #13 performed at the Fast Flux Test Facility (FFTF) as one of passive safety demonstration tests hosted by the International Atomic Energy Agency. In the first step, the transient calculation with major reactivity feedback mechanisms and imposing measured power history was conducted with Super-COPD. The results were compared with the measured data. Several challenges were identified about the thermal–hydraulic model in the core and heat transport systems and the reactivity model in the point kinetics model. Consequently, the transient calculation was performed with modified models considering the radial heat transfer, flow redistribution, and interwrapper flow in the core. The time-dependent change and the second peak of the outlet temperature in the specified assembly were predicted accurately because the core thermal–hydraulics was reproduced well. Furthermore, the assembly bowing reactivity model was developed to improve the net reactivity, thereby resulting in a good agreement with the measured data on the reactor power, temperature, and flow rate in the primary loop. Thus, this paper proves the validation of Super-COPD for the LOFWOS event in the BDBE.

Comparison of Zirconium Redistribution in BISON EBR-II Models Using FIPD and IMIS Databases with Experimental Post Irradiation Examination

Metallic fuels have seen increased interest for future sodium fast reactors due to their material properties: high thermal conductivities and advantageous neutronic properties allow for greater fission densities. One drawback to typical metallic fuels is zirconium redistribution, which impacts this advantageous material and its neutronic properties. Unfortunately, the processes behind zirconium migration behavior are understood using first principles, so before these fuels are implemented in future fast reactors, characterization and fuel qualification regimes must be completed. These activities can be supported through the use of robust modeling using the most accurate empirical models currently available to fuel researchers around the world. The tool that allows researchers to model this complex coupled thermo-mechanical behavior and nuclear properties is BISON. Additionally, BISON model parameters need to be compared against PIE measurements. The current work utilizes two fuel pins from EBR-II experiment X441 to optimize various model parameters, including porosity correction factor, thermal conductivity, phase transition temperature, and diffusion coefficient multipliers, before implementing the final model for seven fuel pins with differing characteristics. To properly evaluate the BISON simulations, the results are compared to PIE metallography data for each fuel pin, to ensure the zirconium redistribution is properly reflected in the simulation results. Six out of seven analyzed fuel pins demonstrate good agreement between the metallography images and BISON results, showing alignment of the Zr-rich, Zr-depleted, and moderately Zr-enriched zones at various axial heights along the fuel pins. Further work is needed to refine the model parameters for general pin use.

Investigation on the thermal hydraulic characteristics of sodium-supercritical CO2 in compact heat exchange channel

Sodium-cooled fast reactor coupled with supercritical CO2 Brayton cycle has broad development prospects owing to the high thermal efficiency, smaller components and compact footprint. As one of the key components, sodium-supercritical CO2 compact heat exchanger plays a vital role in improving the operation performance of coupled power systems. In this paper, a numerical model for coupling heat transfer of sodium-supercritical CO2 in a straight-channel compact heat exchange channel is established and the prediction accuracy of the model is verified with experimental results. The influence of structural parameters and flow parameters on the resistance characteristics and heat transfer performance of cold and hot channels is systematically analyzed. The results show that the structure of semi-circular cross-section with an ABAB layout performs best in heat transfer performance. For the sodium channel, performance evaluation criteria gradually grow with the increase of sodium inlet velocity and slowly decline with the supercritical CO2 inlet velocity. For the supercritical CO2 channel, performance evaluation criteria decrease with the increase of sodium inlet velocity and increase with the supercritical CO2 inlet velocity. Reducing the supercritical CO2 inlet temperature could effectively improve the thermal hydraulic performance of the sodium-supercritical CO2 compact heat exchange channel.

Fuel cycle cost comparison between lead and sodium cooled fast reactors

Sodium and lead coolants are the most recognized coolants for Liquid Metal-cooled Fast Reactor (LMFR). The two coolants have significantly different physical, thermal and chemical properties, which results in major core design differences between lead- and sodium-cooled systems. This work attempts to quantify the impact of the coolant type on the fuel cycle cost of a LMFR by comparing the Levelized Fuel Cycle Cost (LFCC).The paper is composed of two studies with the goal of comparing the advantage of the superior neutronic performance of lead versus the advantage of the tight lattice that can be achieved with sodium coolant. In the first study, the influence of the neutronic differences between lead and sodium on the fuel cycle cost are isolated and quantified. This is achieved by comparing the fuel cycle performance of a Lead-cooled Fast Reactor (LFR) core model to that of a Non-optimized SFR (NSFR) model. The two models have a similar geometry, fuel loading, fuel type, and structural materials. The only major difference between the two cores was the coolant (i.e. lead vs. sodium). Based on this comparison, it is shown that the better neutronic performance of lead coolant results in longer cycle length for the LFR core. It is also shown that this difference in cycle length can result in up to 30.8% better LFCC for a lead-cooled system compared to a sodium-cooled one. In the second study, a Sodium-cooled Fast Reactor (SFR) is designed that is optimized for the sodium coolant. The design process for the Long-life core Sodium Fast Reactor (LSFR) realizes the compatibility of sodium coolant with structural materials and allows for a more compact design. The design parameters and constraints of the design are presented, and the objective of the design is to reduce the LFCC. In this study, the design with the sodium coolant offers a 2% reduction in the LFCC, along with a smaller core size.The overall results show that if you compare sodium and lead using the same reactor geometry, the neutronic benefits of lead coolant can lead to a 30% reduction in the LFCC. However, if you optimize the geometry to take advantage of the higher power density allowed with sodium coolant, the sodium coolant can offer a 2% LFCC reduction or a smaller core size.

On the role of low frequency modes in the thermal and momentum mixing in parallel plane jets

Parallel plane jet mixing has been the subject of extensive study and its physics are characterized well in literature. However, the physics of parallel plane jet mixing in relatively small confined regions is not well known. In this study we investigate the effects of re-circulatory flow on thermal and momentum mixing in confined parallel plane jets. Studying confined jet interaction is important for a variety of engineering applications. One of these applications is sodium cooled fast reactors (SFR). It is common in SFRs to have coolant stream mixing in a confined domain, often the upper plenum. The Reactor Cavity Cooling System (RCCS) facility at the University of Michigan models this problem in a simplified enclosure with isothermal parallel plane jets as the incoming coolant streams. A Large Eddy Simulation (LES) of this simplified plenum was developed using optimal boundary conditions, proven mesh convergence, and consistent resolution. This model was then validated against experimental results from the RCCS facility. After model validation, flow structures in the jet interaction were discerned by means of Proper Orthogonal Decomposition (POD). Comparison of 3D POD to a clipped 2D POD region reveals that re-circulatory flow can lead to dominant oscillations within the jet mixing region. Subsequently, Power Spectrum Density (PSD) and Continuous Wavelet Transforms (CWT) were used to ascertain frequency-based phenomena that might occur in this mixing region. Frequency analysis uncovered a dominant low-frequency oscillation in both velocity and temperature PSDs. These frequencies fall within the known ranges of thermal striping in sodium. The data generated in this study can be used for the benchmarking of less computationally intensive approaches. This data will be made available upon request to ebm5351@psu.edu.

Uncertainty reduction of sodium void reactivity using data from a sodium shielding experiment

ABSTRACT This study investigated the feasibility of reducing the uncertainty associated with fast-reactor-core design by sharing an experimental database between different fields (e.g. reactor physics and radiation shielding) using data assimilation techniques. As the first step in this study, we focused on the ORNL sodium shielding experiment and investigated the possibility of using the experimental data to reduce the uncertainty in sodium void reactivity (SVR), which is the most important safety parameter for sodium-cooled fast reactors. A sensitivity analysis based on the Generalized Perturbation Theory was performed for the sodium shielding experiment. Using the sensitivity coefficients evaluated here and those of the sodium void reactivity previously evaluated by the JAEA, we showed that sodium shielding experimental data can contribute to the uncertainty reduction of SVR by adopting the cross-section adjustment method. Based on this study, the uncertainty reduction effect is expected to be significant, especially for SVR dominated by neutron-leakage phenomena. Although new reactor physics experimental data on SVR may be difficult to obtain, the results of this study suggest that data from sodium shielding experiments can partially substitute for this role. This study demonstrated the value of the mutual use of integral experimental data in fast reactor designs.

Attractiveness of Fissile Material in Nuclear Wastes from Different Fuel Cycles

Nuclear fuel cycle advancements will result in new types of fissile material, including nuclear wastes, that require security and safeguards. Nuclear wastes may be more vulnerable for diversion by non-state actors, and chemical processing to recover fissile material is not an insurmountable challenge. Previous work has applied a figure of merit (FOM) to assess material attractiveness and security risks. This analysis applies the material attractiveness FOM to wastes produced by fuel cycles from the Fuel Cycle Evaluation and Screening (FCES) study. Two aspects of security risk are studied: (1) the time before the fissile material in the waste becomes attractive and (2) the number of waste packages required to obtain a critical mass of fissile material. Two fuel cycles are presented to highlight detailed results: (1) once-through use of low-enriched U in light water reactors (LWRs) and (2) continuous recycle of Pu in sodium fast reactors (SFRs). Increasing LWR used nuclear fuel (UNF) package loading increases the time to attractiveness, but the larger packages contain enough Pu for multiple critical masses. The high-level waste (HLW) from processing the SFR fuels has similar FOM behavior but longer time to attractiveness due to the concentration of fission products. More HLW packages are required to obtain a critical mass; that number can be further increased by increasing the separation efficiency. Extended to all FCES fuel cycles, the minimum time before attractiveness is generally lower for UNF than for HLW because radioactivity is concentrated in HLW. For nearly all fuel cycles that produce UNF, only one package is required to recover enough fissile material for a critical mass. Notably, some advanced fuel cycles produce HLW, of which only two packages need to be recovered to obtain a critical mass, even when the target fissile material is recycled. Going forward, an assessment of the security risks posed by fissile material in nuclear wastes will need to quantify the challenge posed by separations. Ultimately, the assessment could inform security and response measures; whether any of the observations might affect these measures could be an area for future work. Finally, future analysis could study whether different fuel cycle wastes are more attractive for use in radiological dispersal devices or radiological exposure devices.

Effects of irradiation on the insolubility of sodium fast reactor (U,Pu)O2 mixed oxide with a very high amount of plutonium

Sodium fast nuclear reactors might operate with MOX fuels containing up to 45% of plutonium, nevertheless, the reprocessability of spent nuclear fuel is sparsely investigated in the literature for plutonium contents higher than 30%. The dissolution step is a critical stage for MOX fuels since plutonium oxide exhibits a much slower dissolution kinetics as compared to uranium oxide. In particular, many authors suggested a significant change in the dissolution behavior of mixed oxides for plutonium contents between 30% and 50%. In this paper, the dissolution tests of unirradiated and irradiated CAPRA pellets with an initial plutonium content of 45% is described. Up to our knowledge, this is the first dissolution test reported in the literature with such high plutonium content in irradiated nuclear fuel. The fresh unirradiated pellets appears almost insoluble due to a very slow dissolution kinetics. Although the insolubility remains high after irradiation, a beneficial effect of irradiation on the dissolution behavior is observed with a depletion factor of five. The dissolution rates observed are very consistent with previous studies reported in the literature. The results show that complementary steps such as fuel pre-treatment or depletion steps should be considered in order to recycle spent nuclear fuel highly enriched in plutonium.

Technological readiness of alternative reactor concepts

Abstract. Some alternative reactor concepts are frequently discussed in the literature as relevant for the disposal of high-level radioactive waste. We present results on the technological readiness of and research and development on those reactor concepts from a project commissioned by the Federal Office for the Safety of Nuclear Waste Management on novel reactors (Pistner et al., 2023). The focus is on 7 technology lines and 10 specific reactor concepts within those lines. These are sodium-cooled fast reactors (SFRs), lead-cooled fast reactors (LFRs), gas-cooled fast reactors (GFRs), very high-temperature reactors (VHTRs), super-critical water-cooled reactors (SCWRs), molten-salt reactors (MSRs) and accelerator-driven systems (ADSs). For all technology lines considered, extensive research and development work has been taking place for several decades and in some cases since the middle of the last century. Depending on the technology line, technical test stands for individual phenomena have been built and operated and so have smaller experimental reactors and larger demonstration reactors. Nevertheless, until today no commercially competitive reactor concept exists in the field of alternative reactor concepts. The most extensive technical experience is available for the SFR and VHTR technology lines. However, proof of reliable operation under economic boundary conditions is still required for SFRs and VHTRs. For other technology lines, neither has technical feasibility in the form of a demonstration reactor been demonstrated so far (LFR, MSR) nor are more extensive findings from smaller experimental reactors available (GFR, SCWR, ADS). To plan, license, construct and operate such experimental and demonstration reactors, a period of at least 1 to 2 decades must be assumed for each reactor project and probably substantially more based on historical experience. It has been shown that developers' schedules are often characterized by overly optimistic assumptions, that developments are delayed by years or even decades, and that in many cases specific approaches are discontinued completely because the underlying technological difficulties could not be overcome. Thus, the time still required for the development of alternative reactor concepts is probably in the range of several decades. Against this background, it cannot be assumed that such reactor concepts will be used on a relevant scale by the middle of this century. Additionally we will present general results of this study regarding criteria such as safety, fuel supply and waste disposal, proliferation, and costs and point out advantages and disadvantages of those systems over light-water reactors.

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 (&lt;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.

Utilizing a Virtual Sodium-Cooled Fast Reactor Digital Twin to Aid in Diversion Pathway Analysis for International Safeguards Applications

Digital twin technology can improve the effectiveness of international safeguards inspectors by providing a tool that can perform an accurate acquisition pathway analysis, identify pathway indicators, develop required sensors to detect indicators, and monitor facilities in real time using critical data streams that benefit from this safeguards-by-design approach. Safeguards inspectors are required to visit facilities and verify the nuclear material to ensure no diversion has taken place and to detect facility misuse; however, this analysis and verification effort is time consuming, and with limited funding, it is imperative that time spent at a nuclear facility is focused on key areas. We developed a virtual digital twin of two general sodium-cooled fast reactors and explored diversion and misuse scenarios to determine how a digital twin could provide inspectors with an understanding of how proliferation may occur and where the most likely areas for proliferation would be. For each of the three reactors, an optimization algorithm was able to find core designs that would be difficult to detect via sensors alone; however, a machine learning adapter provided by the digital twin was able to show general trends in where proliferation is likely to take place.

Experiment and new analysis model simulating in-place cooling of a degraded core in severe accidents of sodium-cooled fast reactors

In severe accidents of SFRs, the cooling of the residual core materials, which is called “in-place cooling”, is one of the important factors for In-Vessel Retention (IVR). For the evaluation, behavior of the in-place cooling was analyzed by the SIMMER-III code. In order to understand the in-place cooling, method of Phenomena Identification and Ranking Table (PIRT) was applied. Based on the result, an out-of-pile experiment which focused on the extracted factors was conducted. In the experiment, continuous oscillation of sodium level was observed by sodium vaporization and condensation. Analysis for the out-of-pile experiment was conducted by SIMMER-III, but the results were different between the experiment and the analysis. By investigation of the analysis result, it was revealed that the difference was due to occupation of non-condensable gas. Therefore, an analysis model of inter-cell gas mixing was newly developed, and the agreement was significantly improved by the new model.

A study on transport and plugging of sodium aerosol in leak paths of concrete blocks

In the event of a severe accident in Sodium Cooled Fast Reactors (SFR), the sodium combustion aerosols along with fission product aerosols would migrate to the environment through leak paths of the Reactor Containment Building (RCB) concrete wall under positive pressure. Understanding the characteristics of sodium aerosol transport through concrete leak paths is important as it governs the environmental source term. In this context, experiments are conducted to study the influence of various parameters like pressure, initial mass concentration, leak path diameter, humidity etc., on the transport and deposition of sodium aerosols in straight leak paths of concrete. The leak paths in concrete specimens are prepared by casting and the diameter of the leak path is measured using thermography technique. Aerosol transport experiments are conducted to measure the transported and plugged aerosol mass in the leak paths and corresponding plugging times. The values of differential pressure, aerosol concentration and relative humidity taken for the study are in the ranges 10–15 kPa, 0.65–3.04 g/m3 and 30–90% respectively. These observations are numerically simulated using 1-Dimensional transport equation. The simulated values are compared with the experimental results and reasonable agreement among them is observed. From the safety assessment view of reactor, the approach presented here is conservative as it is with straight leak paths.

Benchmark analysis of the FFTF LOSWOS test #13 with OpenMC and THACS

In order to support collaborative efforts within international partnerships on the development of sodium-cooled fast reactor (SFR), the International Atomic Energy Agency (IAEA) proposes a coordinated research project (CRP) about the unprotected loss-of-flow-without-scram test (LOSWOS) performed at Fast Flux Test Facility (FFTF) in 2018, which could validate simulation tool and numerical models for SFR. As one of the participants, the Nuclear THermal-hydraulic Laboratory at Xi’an Jiaotong University (NuTHeL) participates with the monte carol code OpenMC and system thermal hydraulic code THACS. This paperdeal with the standalone neutronic modeling by OpenMC and thermal hydraulic modelling by THACS. For neutronic part, the axial homogenous model was modeled to analyze the reactivity feedback coefficients and steady radial power distribution. For thermal hydraulic part, the system model, including 2D CFD outlet plenum model, subchannel model and inter-wrapper flow model, was modeled to analyze both system response and local phenomena during the LOSWOS. In general, both neutronic results and thermal hydraulic results agreed well with other participants or experiments, the reactivity feedback of GEM could provide major negative feedback, which could be used as a novel passive shutdown device. However, the remaining discrepancies were the underestimated peak temperature of PIOTA at the begin of transients, it may be caused by unknow heat transfer coefficients around hexagon tube, which has no correlation and should be further investigated by experiments. In addition, when considering IWF model, the maximum temperature of the outlet for the active fuel region was 10 K larger than the subassemblie’ outlet in the PIOTA, this phenomenon should be carefully treated in loss of flow accidents of SFR.

Comparative analysis of fuel burnup calculations of fourth-generation European fast reactors

Abstract The sustainability of nuclear energy implies the continuous research and development of fast reactor technology. This work represents a comparative analysis of fuel burnup calculations of three commercial-level fast spectrum concepts which are under research and development in Europe. The investigated designs are the European Sodium-cooled Fast Reactor, the European Lead-cooled Fast Reactor, and the European Gas-cooled Fast Reactor. MCNPX transport code was used to design three representative fuel assemblies of these fourth-generation concepts to analyze and compare their neutronic and safety parameters. The neutronic and safety analysis in this work includes the evolution of infinite multiplication factor and fissile inventory, neutron yield, the average energy of neutrons causing fission, neutron energy spectrum, neutron generation time, Doppler reactivity effect, effective delayed neutron fraction, and coolant void reactivity coefficient. The fuel burnup results showed that the ESFR is superior regarding the reactivity swing and breeding of fissile 239Pu. On the other hand, analysis of safety parameters results showed that they depend mainly on fuel composition rather than on other design specifications.