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AbstractFuel-cladding chemical interaction is recognized as a significant challenge in metallic fuel/steel cladding systems due to the formation of low-melting-point intermetallic eutectic compounds between fuel and steel cladding constituents. To address this, the study explores diffusion barrier coatings applied via metal-organic chemical vapor deposition, chosen for its low processing temperature under 600°C, thus preventing thermal degradation of steel cladding. In this study, we successfully developed thin, dense coatings ranging from a few to several micrometers in thickness. These coatings are composed predominantly of a mixture of V2C and vanadium carbide (VC) phases. Following the coating process, the T91 ferritic/martensitic (F/M) steel substrates remained intact with no noticeable decarburization or reduction in microhardness near the VC coating. Further testing through diffusion couple experiments at 550°C for 100 h revealed that an 8-µm-thick VC coating layer can effectively prevent interdiffusion between cerium and T91 F/M steel. Leveraging optimized processing conditions on flat coupon samples, this deposition was also applied to an archived EBR-II HT-9 steel cladding. These results demonstrate promising applications for sodium-cooled fast reactors (SFRs). [ABSTRACT FROM AUTHOR]

The annular fuel has two cooling surfaces inside and outside, which can fully take away the heat of the fuel assemblies. It is urgent to conduct research on sodium-cooled fast reactor annular fuel assemblies. The Prandtl number Pr of liquid sodium is much less than 1 (Pr ˂˂ 1). The problem is that using the traditional constant turbulent Prandtl number Prt of 0.85 ~ 0.9, which will greatly affect the accuracy of the numerical prediction of the thermal-hydraulic properties of liquid sodium. In order to address this issue, the turbulent heat transfer characteristics of the bare and wire spacers 7-pin annular fuel assembly for a sodium-cooled fast reactor are investigated by a SST k-ω-kθ-εθ model. The open-source CFD software OpenFOAM is used for this calculation. In order to verify the validity of the present calculation method, some classical liquid metal heat transfer correlations are used to compare and analyze the results of the present calculation method (kθ-εθ model), the results of the Reynolds analogy hypothesis (Prt = 0.85 model), and the results of the turbulent Prandtl number correlation (Prt = kays model). It is shown that the SST k-ω-Prt = 0.85 model overestimates the Nusselt number Nu of the annular fuel assembly. The SST k-ω-kθ-εθ model agrees well with the SST k-ω-Prt = Kays model of the bare and wire spacers 7-pin annular fuel assembly. It suggests that the numerical calculation results of the SST k-ω-kθ-εθ model are reliable in both the simple and complex fuel assembly models. In brief, the SST k-ω-kθ-εθ model is more capable of predicting the turbulent heat transfer characteristics of the bare and wire spacers 7-pin annular fuel assembly in sodium-cooled fast reactors. In addition, the thermal-hydraulic characteristics of annular fuel assemblies with wire spacers for sodium-cooled fast reactors have been investigated. This research can provide a reference for the study of turbulent heat transfer characteristics of annular fuel assemblies in sodium-cooled fast reactor. [ABSTRACT FROM AUTHOR]

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

One of the issues in the development of the sodium-cooled fast reactor (SFR) using transuranic (TRU) metallic fuel is the absence of criticality benchmark experiment that faithfully mocks up the nuclear characteristics of the target design for validation of the reactor core design code and its uncertainty quantification (UQ). This study aims to quantify the criticality uncertainty of a typical TRU burner with metallic fuel by using the standard upper safety limit (USL) estimation framework based on the similarity analysis of existing benchmark experiments but elaborated in two aspects:1) application of two-sided rather than one-sided tolerance interval and 2) inclusion of additional uncertainty to account for fission products and minor actinides not included in the benchmark experiments. To conduct the similarity analysis and evaluate the nuclear-data induced uncertainty, existing, well-verified computing codes were integrated, including the nuclear data sampling code SANDY, the nuclear data processing code NJOY, and the continuous-energy Monte Carlo code McCARD. Finally, using the SFR benchmark database comprising both publicly available and proprietary benchmark experiments, the criticality uncertainty of the TRU core model with metallic fuel was evaluated.

In case of a hypothetical severe accident in a Sodium-cooled Fast Reactor (SFR), coolability of the debris bed in the post-accident phase plays a vital role in mitigating the accident and ensuring the structural integrity of the reactor vessel. Few numerical studies are reported in literature, in which the boiling heat transfer in debris bed is expressed as equivalent heat conduction using similarity law between heat conduction and two-phase heat transfer. However, these studies assumed steady state mass conservation for the boiling zone and neglected the gravity force. Hence, a detailed study has been carried out for various particle sizes and porosities of SFR debris to investigate the influence of above considerations. The effect of gravity on debris bed coolability is studied using steady state model of Lipinski, which showed that gravity has a non-negligible effect, for particle size of 0.3 mm and porosity of 0.5. However, the gravitation force was found to have a negligible effect in dryout heat flux estimation for the bottom cooled configuration. A transient numerical model is developed for simulating the boiling phenomena in debris beds and validated with the published experimental results. The assumption of steady state mass conservation is verified by carrying out transient analysis, which indicated early prediction of the dryout inception. For time dependent heat generation case, the unsteady mass conservation predicted higher DHF compared to constant heat generation.

With the aim of “carbon peaking” and “carbon neutrality” in China, hydrogen energy will play a crucial role of in the future. The third-generation nuclear power (GEN-Ⅲ) features a relatively low reactor outlet temperature, with the primary application of its generated electric energy to direct electrolysis of water to produce hydrogen. In contrast, the fourth-generation nuclear power (GEN-Ⅳ) has higher thermal efficiency and reactor outlet temperature. Considering the matching of reactor outlet temperature with the hydrogen production process, cogeneration can be employed to effectively reduce the cost of hydrogen production. Sodium-cooled fast reactor (SFR) is one of the GEN-Ⅳ reactors, which has the advantage of high reactor outlet temperature, high efficiency and high-tech maturity, and can provide the thermal power and electrical energy for the hydrogen production process. SFR also have advantages of proliferation of nuclear fuel and burn long-lived radioactive nuclides. The closed fuel cycle formed by the development of fast reactors and pressurized water reactors can effectively achieve the sustainable development of nuclear energy. Hydrogen production could be considered in the design of the next generation million kilowatts-level fast reactor (CFR1000), making the hydrogen production process more economical and reduction of carbon emissions, simultaneously. Considering the reactor outlet temperature and the aim of zero-carbon emission during the hydrogen production process, the results reveal that the most suitable hydrogen production methods for SFR are the thermo-chemical cycle (copper-chlorine cycle) and conventional electrolysis. For conventional electrolysis, hydrogen production efficiency is relatively low, but the technology maturity is much higher compared with the other methods. For the thermo-chemical cycle (copper-chlorine cycle), technology maturity is relatively low but has higher hydrogen production efficiency, which could make the hydrogen production process more economical in the future. Based on the hydrogen economic evaluation program HEEP (V2021) which was developed by the IAEA, an economic study on the hydrogen production cost per kilograms of CFR1000 was conducted, which was verified by the fourth-generation reactor economic evaluation software G4-ECONS V3.0 developed by GIF. The cost of the hydrogen production by conventional electrolysis coupled with the CFR1000 is 3.05 / � � , � � � � ℎ � � � � � � � ℎ � � � � � � � � � � � � � � � � � � � � � � � � � � � � � � � � � � ℎ � ℎ � � ℎ � � � � − � ℎ � � � � � � � � � � � � � 4.83 /kg. However, with the investment ratio reduction of SFR, and the potential impact of carbon sinks in the future, the cost of hydrogen production will be further explored. HEEP calculation results show that a decrease in the overnight investment of CFR1000 could significantly reduce the cost of hydrogen production. Overall, SFR shows great promise in the field of hydrogen production in the future.

The Sodium-Cooled Fast Reactor (SFR) is a promising concept chosen in the Generation IV International Forum as a possible design for pursuing the sustainable use of nuclear energy. Its core consists of multiple hydraulically isolated assemblies, with a tightly packed triangular lattice array of fuel pins enclosed in a hexagonal duct present within each assembly. Helical wire spacers are wrapped along the axis of the rods to maintain a gap between them, inducing a secondary flow, increasing the channel mixing, and enhancing convective heat transfer. In this study, a direct numerical simulation campaign is conducted for a simplified 7-pin wire wrapper geometry, with Reynolds numbers ranging from $Re$ Re = 1000 to 10 000 and a Prandtl number of $Pr$ Pr = 0.005, to investigate heat transfer in low-flow conditions. The wire wrapper case is compared to a bare bundle case with seven pins. The results are discussed, and heat transfer predictions are compared between our numerical results and classic correlations. An anisotropy invariant map is obtained for the above-mentioned cases, and turbulent kinetic energy and turbulent heat flux budgets are computed and analyzed. Our findings provide unique insights into the flow behavior within a wire-wrapped bundle. [ABSTRACT FROM AUTHOR]

For all the nuclear reactor systems, quantitative assessment of the accident management (AM) effects against long-term external hazards became one of the essential issues after the lesson learned from the Fukushima Daiichi Nuclear Power Plant accident. However, the influence from the safety systems' stochastic and dynamic shifting between multiple working states, which is related to the interaction with the adjacent components/systems in general, has not been accounted for yet. Therefore, this research aims to develop a dynamic probability risk assessment tool considering repairable multi-component interdependency for investigating the AM influences on the multi-state safety systems under long-term external hazards. Based on the newly proposed methodology in this research via integrating the Petri net (PN) model with the continuous Markov chain Monte Carlo (CMMC) method, a framework applying PN-CMMC methodology to a severe accident analysis code, SPECTRA, had been originally constructed. Different AM influences on the multi-state decay heat removal systems against long-term volcanic ashfall were also quantitatively confirmed, indicating that halving the repairing time is more influential in suppressing the core damage frequency than doubling the number of adjacent electricity support systems. Therefore, the PN-CMMC-SPECTRA framework can further assess the uncharted dynamic multi-state concerns, leading to a safer AM strategy. [ABSTRACT FROM AUTHOR]

Boron carbide (B4C)–stainless steel (SS) eutectic reaction behavior is one of the most important issues in the core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). In this study, the immersion experiments using B4C pellets with molten SS were conducted to evaluate the CDA sequences such as contact event of solid B4C with degraded core materials including SS at very high temperature. The immersion experiment aims at understanding the kinetic behavior of solid B4C–liquid SS reaction based on the reduced thickness of B4C pellet after the experiment in the temperature ranges from 1763 to 1943 K, which is higher than the temperature of solid B4C–solid SS reaction. Based on the kinetic consideration of the reaction rate constants for solid B4C–liquid SS reaction, it was found that similar temperature dependency was identified between solid B4C–liquid SS and solid B4C–solid SS. Besides, the reaction rate constants of solid B4C–liquid SS were smaller than those of solid B4C–solid SS extrapolated in higher temperature region by two or more orders of magnitude due to two different evaluation method for B4C side/SS side. It was confirmed that this difference was reasonable through the consideration of previous reaction tests in solid–solid contact for B4C side/SS side. [ABSTRACT FROM AUTHOR]

Evaluating severe accidents in Generation IV sodium-cooled fast reactors (SFRs) is challenging due to the eutectic reaction between boron carbide (B 4 C) and stainless steel (SS), which results in boron migration and enhanced neutron absorption. A novel quantitative visualization method using radiative heating to observe eutectic behavior and resulting melt structure during boron migration. Experiments, replicating real-scale B 4 C-SS control rods, were conducted in an integrated test facility under core disruptive accidents (CDAs) conditions, introducing a new approach. This study explores the long duration melting (candling phenomenon) of boron within stainless steel, shedding light on relocation behavior not previously studied. We identified two distinct failure mechanisms: the separation of SS from the B 4 C pellet, resulting in the formation of a later melting drop, and the fracture of the B 4 C pellet into multiple pieces, possibly due to thermal stress. The visualization technique and thermal interfacial resistance analysis precisely captured the eutectic temperature, aligning well with prior research. Our study yield insights into eutectic melt relocation behavior by systematically characterizing relocated solidified eutectic melt with a cooling rate of 50 °C/s in stagnant argon environment via scanning electron microscopy-energy dispersive X-ray spectroscopy (SEM-EDS), X-ray diffraction (XRD), and microhardness testing. Our study shows that due to high chromium content, (Cr,Fe) 2 B phases are observed in XRD analysis instead of just Fe 2 B phase. Other phases present in our solidified eutectic melt include γ-Fe (fcc), (Cr,Fe) 23 (B,C) 6 (cubic), and (Cr,Fe) 5 B 3. The atomic composition within these zones was ascertained through SEM-EDS analysis. Our study shows varying hardness values in distinct phases, reflecting differing boron concentrations within micrograph zones. Chromium promotes (Cr,Fe) 2 B and (Cr,Fe) 5 B 3 boride formation, transforming the structure from tetragonal to orthorhombic with increased chromium content, thereby increasing hardness. This validates B 4 C-SS eutectic mixture relocation and diverse boride phase formation under extreme reactor conditions. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

In the event of core disassembly accident (CDA) in a sodium-cooled fast reactor (SFR), the molten material from the core migrates and interacts with the low-temperature coolant in the lower plenum, resulting in the formation of core debris that eventually settles to forms a debris bed. Improper placement and inadequate cooling of the debris bed can result in the failure of the pressure vessel. To efficiently disperse the core debris and minimize contact between the lower head of the pressure vessel and the molten material, a core catcher was specifically designed as a passive preventive and mitigating device. The structure of the core catcher directly impacts the shape and distribution of the debris bed, which subsequently affects its re-criticality and long-term decay heat removal capability. This study aims to perform a numerical investigation to optimize the structural design of the core catcher, with a specific focus on the mechanisms and laws of the chimney structure of the core catcher on the formation and distribution of the debris bed. The improved discrete element method (DEM), which is based on dimensionless stiffness and dimensionless damping coefficients, was employed to numerically simulate the formation process of the debris bed on the core catcher. This method incorporates source terms into the momentum equation of debris particles to account for the influence of coolant on their motion. Validation experiments were conducted to verify the capability of the numerical simulation algorithm used in optimizing the design of the core catcher. In this paper, the effects of three factors on particle motion and debris bed formation mechanisms were investigated by varying the vertical projection side length of the chimney cover, inclination angle of the cover, and spacing between chimneys. The research findings suggest that the selection of appropriate values for these parameters has a substantial impact on the shape and distribution of the debris bed. Therefore, it is recommended to establish reasonable chimney design parameters to achieve optimal performance in practical applications. Furthermore, this study reveals that during the formation of the debris bed, the debris particles demonstrate a secondary scattering effect, which plays a vital role in enhancing the uniformity of the debris bed. This study will provide valuable engineering references for optimizing the design of the core catcher.

For many years, sodium-cooled fast reactors have occupied the most important part of the closed fuel cycle. In order to improve the economy of sodium-cooled fast reactors, the nuclear industry around the world is actively increasing fuel burnup as much as possible. The behavior simulation of fuel elements under high fuel burnup is a key issue in the design and reliability of fuel elements. In this case, it is necessary to develop computer code that can accurately analyze fuel behavior to evaluate the behavior and reliability of high-fuel fuels, and as a safety analysis tool to evaluate the performance and behavioral evolution of fuel elements under steady-state, transient and accident conditions. For the above reasons, the Chinese Institute of Atomic Energy has developed FIBER, a performance analysis code for fuel elements of sodium-cooled fast reactor. The code consists of two main parts:The first part is used to analyze the temperature distribution, the thermal deformation and fission gas release; The other part is used to analyze the mechanical behavior of fuel elements. In the thermal analysis part, the axisymmetric finite volume method is applied to the entire length of the fuel element. The code has the ability to calculate thermal conductivity, gap heat transfer, coolant heat transfer, fission gas release, fuel restructure, solid fission product migration, and plenum pressure. In the mechanical analysis part, the axisymmetric finite element method is applied to the entire length of the fuel elements. The code can simulate the phenomena of thermal expansion, densification, irradiation swelling, pellet cracking, elasticity, plasticity, creep, and PCMI. The thermal analysis part and the mechanical analysis part are coupled, and the convergence of temperature and deformation is obtained in each time step through iteration. FIBER code consists of many theoretical models, empirical models, and parameters that control the calculation process. However, fuel behavior cannot be explained only by a simple combination of these models, because fuel behavior is the result of the coupling of many phenomena. Therefore, as many cases as possible must be used for code verification to determine the appropriate model and parameter selection. The irradiation data of UO2 and MOX of the Russian BN600 reactor were obtained through research. The two fuel elements operated in the Russian BN600 for 559 days, with maximum fuel burnup of 11.8at% and maximum irradiation damage of 78 dpa. The FIBER code was used to analyze the above two fuel elements. the calculation results of fission gas release rate, irradiation deformation, gap, columnar region, are compared with the irradiation data. The comparison results show that the FIBER code is effective for evaluating the irradiation deformation, columnar crystal region, and fission gas release performance of high burnup fuel elements.

The inherent safety of reactor is emphasized in the design of the sodium-cooled fast reactors (SFRs). For accident mitigation, inherent safety and passive measures are adopted to reduce the demand for power sources and enhance safety and economy. SFRs can effectively prevent the unprotected accidents. The probability of core meltdown is very small. However, in order to prevent the large radioactive release, SFRs still consider the mitigation measures for large-scale core meltdown. For pool-type sodium-cooled fast reactors, there is a large amount of sodium with high heat capacity in the reactor vessel. It is advantageous to install the core catcher in the reactor vessel for collecting and cooling the core melt. FRTAC is a liquid metal fast reactor safety analysis code developed by China Institute of Atomic Energy (CIAE). The Experimental Breeder Reactor-Ⅱ (EBR-Ⅱ) shutdown heat removal test 45R (SHRT-45R) is an unprotected loss-of-flow event conducted by Argonne National Laboratory (ANL). The primary circuit natural circulation test of Phenix Reactor is a main pump shut down with scram event conducted by French Alternative Energies and Atomic Energy Commission (CEA). The FRTAC code simulated the SHRT-45R test and the natural circulation test performed during the Phenix end-of-life. The results showed that the calculated value is in good agreement with the experiments. This indicated that FRTAC code can be used for thermal-hydraulic simulations under natural circulation of SFR. In this paper, FRTAC code was used to analyze the natural circulation heat transfer after the core meltdown. And the core melt in-vessel retention scheme was studied. The analysis object is a 1 500 MWt pool-type sodium-cooled fast reactor. The reactor has four independent decay heat removal systems (DHRSs). The DHRS consists of three loops, namely a primary loop, an intermediate loop and an air cooler loop. According to the research, the relocation time for the core melt to the core catcher is about 12 hours. At this time, the decay heat is about 9 MW. The natural circulation in the reactor vessel can effectively cool the core melt and transfer the decay heat into the sodium pool. The passive DHRSs can export the heat from the sodium pool to the atmosphere. The maximum temperature of the sodium pool does not exceed 450℃ during the cooling of the core melt. Therefore, the integrity of the core catcher and the reactor vessel will not be affected.

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

The finned-tube sodium-to-air heat exchanger plays a crucial role in ensuring the safety of prototype Gen IV sodium-cooled fast reactors. It is designed to actively operate with power, but it is also important to consider its passive performance to ensure availability during power-out conditions. This study focused on addressing this limitation through a natural convection experiment using a dedicated facility, and the obtained results were compared with the safety analysis code MARS-LMR. The overall temperature difference was observed to be within an acceptable range of less than 5 %. While the code calculations aligned well with the experimental results at higher temperatures, they underestimated the experimental results at lower temperatures. These findings can serve as a basis for other codes employing different heat transfer correlations and for future reactor licensing purposes. [ABSTRACT FROM AUTHOR]

Due to the presence of sodium, it is a challenging task to achieve the reliable and safe operation of steam generators in a sodium-cooled fast reactor (SFR). Water flow oscillations in a two-phase flow system worsen the tube integrity. An accurate prediction of two-phase pressure drop is essential in designing steam generators to operate in a stable regime. Toward this, experiments have been carried out on an industrial-size 19-tube model sodium-heated steam generator of 5.5-MW capacity to understand two-phase pressure drop characteristics at various operating conditions. The measured data are used to estimate the two-phase frictional pressure drop. The concept of a two-phase friction multiplier has been used in the present study. A significant variation in the two-phase frictional multiplier is seen with steam quality, whereas the variation of the two-phase friction multiplier is insignificant at saturated steam condition. Based on the experiments, complemented by computational model, a correlation has been developed for the two-phase frictional multiplier as a function of steam quality for sodium-heated once-through straight-tube steam generators. [ABSTRACT FROM AUTHOR]

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

The SPECTRA code has been developed as an integrated safety analysis tool for comprehensive analyses of in- and ex-vessel phenomena during a severe accident (SA) in a sodium-cooled fast reactor (SFR). The individual modules, such as sodium fire and sodium-concrete reaction, have been implemented into SPECTRA to simulate specific phenomena in an accident of SFRs. However, SPECTRA has no capability for the overlapped event involving sodium fire and sodium-concrete reaction in the same compartment because the sodium pool and the floor concrete are modeled in each module independently. The capability of SPECTRA is enhanced by integrating the analyses of sodium pool fire and concrete ablation for overlapped events of the ex-vessel phenomena. A new shared module for the sodium pool and floor concrete region, named “PFC module” was introduced in the previous study. The sodium fire module is connected to the PFC module in this study. The integrated model is validated through the benchmark analysis of the F7-1 pool fire experiment. The calculation results of the temperature and combustion rate show good agreement with the experimental result. A demonstration analysis is also conducted for an overlapped event of the ex-vessel phenomena including both concrete ablation and sodium pool fire. SPECTRA can simulate a reasonable heat and mass transfer behavior during the overlapped event.

The concept of the inherent safety characteristic on thermal-hydraulic of the pool-type sodium-cooled fast reactor was proposed and its formation mechanism was analyzed. That is, the excellent physical properties of sodium (high thermal conductivity, large difference value between melting point and boiling point at atmospheric pressure, and good fluidity) combined with the reasonable design of pool-type reactor (large sodium capacity, height arrangements between heat exchangers as well as cold sodium pool and core), make the reactor has a characteristic: In the early stage of reactor scrams, the sodium and metal components in the reactor vessel can be used as the effective heat sink of the core decay heat, and the heat generated in core can be removed into the sodium pool in time by heat conduction and natural circulation flow to ensure the core safety. Based on the actual reactor natural circulation test data of BN-600 and Phenix reactors, the natural circulation experiment data in the test device with sodium as working fluid and the natural circulation condition predictive calculation of the sodium-cooled fast reactors with typical layout using the different codes, the characteristic and its formation mechanism were demonstrated. The core decay heat can be transferred to the pool in time under the natural circulation condition because the good thermal conductivity of sodium and the natural circulation flow due to the reasonable reactor design result in the establishment of a good heat exchange relationship among assemblies by the wrapper walls and the sodium in inter-wrappers, then, on the one hand, the heat of the assemblies is discharged by the combined action of the sodium in the wrappers and inter-wrappers, on the other hand, the outlet temperature differences among assemblies are leveled to a certain extent, then the peak temperature is lowered. Based on the above research, for the pool-type sodium-cooled fast reactor, in the early stage of reactor emergency shutdown, the reactor’s inherent thermal-hydraulic safety characteristic alone can ensure the core safety without other heat sink, and in the later stage, it is only necessary to establish an effective heat sink from outside the reactor before the sodium temperature in the reactor rises to the safety limit. The effective heat sink can be the specially arranged decay heat removal system or the reactor vessel conventional heat loss item etc. And for the typical decay heat removal system setting heat exchangers in cold sodium pool or hot sodium pool, the system cannot play a role to removal core heat in the early stage of operation, at this time it must to rely on the inherent thermal-hydraulic safety characteristic as heat sink to ensure core safety.

After core meltdown of the sodium-cooled fast reactor (SFR), the jet of molten corium interacts with liquid sodium, which is called molten fuel-coolant interaction (MFCI). The molten corium is fragmented and solidified. The formed debris is then relocated on the core catcher or in the lower head to form debris bed. The morphological structure of the debris bed has an important influence on the cooling characteristics. However, due to the opacity of liquid sodium, it is difficult to investigate the relocation behavior of melt debris flow through experiments. In this paper, a CFD-DEM coupling model was established to simulate the relocation behavior of melt debris flow. The model was validated by experimental data. Then, the relocation behavior of UO2 melt debris was simulated, and the effect of debris diameter was studied. The liquid sodium has an effect of stiction on the front of debris flow, forming a mushroom like shape which becomes smaller as the diameter of the debris increases. The liquid sodium near the central axis is most obviously affected by the debris flow, and the velocity of the liquid sodium reaches the maximum value at the upper position of the debris catcher. As the debris diameter increases, the velocity gradient and peak velocity of the liquid sodium decrease, and the vortices formed in the vicinity of debris catcher intensify, contributing to the diffusion of momentum and energy. For different debris diameters, after the debris flows are immersed into liquid sodium, the front motions tend to converge. Then, the debris flow reaches the catcher with a nearly uniform velocity. The projection curve of the top surface of the debris bed satisfies the sine function. The accumulation angle of the debris bed with debris diameter of 1.5 mm is smaller than that of 2.0 mm and 2.5 mm. Thus the porosity of the debris bed is reduced, which is less conducive to the cooling by natural circulation. The debris temperature varies little in the argon atmosphere, while drops sharply at the moment of immersion into liquid sodium. With the increase of debris diameter, the cooling effect of the liquid sodium on the debris flow is enhanced, and the average temperature drop rate of the debris flow decreases due to reduced release rate. This study can provide a guidance for the research on the cooling characteristics of core debris bed, and has a reference value for the design and arrangement of core catcher and the formulation of other mitigation measures of severe accident in SFR.

The effectiveness evaluation of safety measures against severe accident is necessary for restart of experimental sodium-cooled fast reactor Joyo in Japan. These safety measures correspond to those in defense-in-depth (DiD) level 4. In the previous study, an internal event level-1 probabilistic risk assessment (PRA) at power was performed to calculate frequencies of the accident sequences of failure of safety measures in DiD levels 1 to 3, to identify dominant accident sequence groups, and to identify dominant accident sequence for selecting important accident sequences in each accident sequence group, which are needed for implementing the effectiveness evaluation of safety measures in DiD level 4. Based on this, the present study implemented internal event level-1 PRA at power to show quantitatively reduction of those occurrence frequency by the safety measure in the DiD level 4. As a result, the frequency of each accident sequence group decreased significantly, and the total frequency of the accident sequence groups decreased to about 1×10-6 /reactor-year, which is about 1/1000 times the one estimated in the previous study. The protected loss of heat sink was the largest contributor in all the accident groups, and a dominant accident sequence in each accident group was also identified in this study.

This study aims to understand the time-dependent change in the occurrence rate of leak from steam generator (SG) tubes in sodium-cooled fast reactors (SFRs). The target SFRs in the present paper are Phenix in France and BN-600 in Russia. By reviewing publicly available literature that show data from the SFRs, we have investigated the numbers of tube-to-tubeplate welds and tube-to-tube welds, heat transfer areas of tube base metal, operating hours of SGs, dates when SG tube leak occurred, locations of leak, and corrective actions taken after tube leak events, such as replacement of the module, in which a leak occurred. Based on these, we have estimated the time to leak and quantitatively analyzed the time-dependent change of the occurrence rates of SG tube leak for each of the above-mentioned parts by hazard plotting method. The results show the time-dependent change in the rates of both Phenix and BN-600. For Phenix, the rate of leak at tube-to-tube welds shows a significant increase after 40,000 hours. This can be caused by thermal stress repeatedly exerted on the materials. In addition, the decrease in the rate was found, which is probably thanks to improved welding and SG operating conditions. For BN-600, it seems that in many cases, the probable cause of the leak was initial defects that developed to failure during the early stage of reactor operation, and that no special countermeasure was taken in the later stages. Therefore, it would be natural to assume that the rate simply decreased over time.

In order to design fast reactors, it is necessary to consider high cycle fatigue of structural materials up to 1×109 cycles. Since the fatigue usage factor (UF) is calculated down to the order of 0.01 in fast reactor design, it is necessary to develop fatigue curves applicable up to 1×1011 cycles to evaluate high cycle fatigue at 1×109 cycles. In this study, the applicability of the best-fit fatigue curve of JSME up to 1×1011 cycles was verified in order to develop a high cycle fatigue evaluation method for Mod.9Cr-1Mo steel, which is a candidate material for fast reactor structural materials. In the applicability verification, high cycle fatigue tests were first conducted at 500 °C, 550 °C, and 600 °C under strain-controlled conditions, and ultrasonic fatigue tests were conducted at 550 °C. The test results were then evaluated and showed that there was no rate dependence and no difference by product types for high cycle fatigue strength of Mod.9Cr-1Mo steel at high temperatures. It was also clarified that there was no difference in high cycle fatigue strength of Mod.9Cr-1Mo steel at high temperatures even if the failure mode changed from surface origin to internal origin. A best-fit fatigue curve was developed from test data obtained in high cycle fatigue tests up to 1×109 cycles. And based on the high cycle fatigue mechanism discussed from the ultrasonic fatigue test results, the best-fit fatigue curve of JSME was confirmed to be applicable to Mod.9Cr-1Mo steel up to 1×1011 cycles regardless of product types.

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 paper provides a better guide to understanding the overall heat transfer effect by optimizing the heat transfer model. [ABSTRACT FROM AUTHOR]

Since the leakage of sodium in an SFR (sodium-cooled fast reactor) causes an explosion upon reaction with air and water, sodium leakages represent an important safety issue. In this study, a novel technique for improving the reliability of sodium leakage detection applying DDVR (dynamic data validation and reconciliation) is proposed and verified to resolve this technical issue. DDVR is an approach that aims to improve the accuracy of a target system in a dynamic state by minimizing random errors, such as from the uncertainty of instruments and the surrounding environment, and by eliminating gross errors, such as instrument failure, miscalibration, or aging, using the spatial redundancy of measurements in a physical model and the reliability information of the instruments. DDVR also makes it possible to estimate the state of unmeasured points. To validate this approach for supporting sodium leakage detection, this study applies experimental data from a sodium leakage detection experiment performed by the Korea Atomic Energy Research Institute. The validation results show that the reliability of sodium leakage detection is improved by cooperation between DDVR and hardware measurements. Based on these findings, technology integrating software and hardware approaches is suggested to improve the reliability of sodium leakage detection by presenting the expected true state of the system.

To understand the eutectic reaction mechanism and the relocation behavior of the core debris is indispensable for the safety assessment of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). This paper addresses reaction products and their distribution of the eutectic melting/solidifying reaction of boron carbide (B4C) and stainless-steel (SS). The influence of the existence of carbon on the B4C-SS eutectic reaction was investigated by comparing the iron boride (FeB)-SS reaction by Raman spectroscopy with Multivariate Curve Resolution (MCR) analysis. The scanning electron microscopy with dispersive X-ray spectrometer was also used to investigate the elemental information of the pure metals such as Cr, Ni, and Fe. In the B4C-SS samples, a new layer was formed between B4C/SS interface, and the layer was confirmed that the formed layer corresponded to amorphous carbon (graphite) or FeB or Fe2B. In contrast, a new layer was not clearly formed between FeB and SS interface in the FeB-SS samples. All samples observed the Cr-rich domain and Fe and Ni-rich domain after the reaction. These domains might be formed during the solidifying process.

Under severe accident in sodium-cooled fast reactor (SFR), molten corium may fall into the coolant and cause molten fuel-coolant interaction (MFCI). The molten corium is fragmented and solidified to produce debris during MFCI. The debris is relocated on the core catcher or lower head to form core debris bed. The coolability and recriticality of core debris bed are significantly affected by the fragmentation characteristics of molten corium jet. Based on the linear stability theory, kinematic equations and modified Laplace’s law at the contact interface, the growth of surface disturbance of molten corium jet was derived to consider the effect of boiling and solidification behavior. The fragmentation model of molten corium jet was established. Then the criteria for fragmentation of molten corium jet under typical conditions were acquired. The model developed considers the effects of three factors: the relative velocity between melt and sodium, sodium boiling and melt solidification. Among them, the relative velocity is reflected by the classical K-H instability, and sodium boiling and melt solidification are reflected by the impact pressure coefficient Cpb and bending stiffness Ds. Furthermore, the proposed model was compared with COSA experiment results. For the three molten aluminum experiments under non-boiling conditions, when Ds is around 2.0×10-3 N·m, the predicted mass median diameter (DMM) are in good agreement with the experimental values of test Al-1 and test Al-4. While the experimental results of Al-2 are in good agreement with the results of model with Ds at 1.0×10-2N·m because the initial temperature of aluminum is lower and the solidified layer is thicker. For the two stainless steel experiments under boiling conditions, when Cpb is in the range of 106-108 Pa·m-1 and Ds is in the range of 10-5-10-6 N·m, the predicted DMMs are in good agreement with the experimental values. Since sodium is an opaque medium, it is difficult to obtain the visualization results of jet fragmentation, such as the thickness of solidified layer (which can be used to calculate Ds) and the interface disturbance amplitude (which can be used to calculate Cpb) during the fragmentation process. In the follow-up, the key parameters of the fragmentation process can be obtained by numerical simulation method (such as MPS). Then the model can be revised and improved in detail. The research in this work not only provides a reliable tool for the evaluation of severe accident but also offers important guiding significance and reference value for design of mitigation measurements of SFR.

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For a sodium-cooled fast reactor, the capability for stable cooling and avoiding re-criticality on the debris bed is essential for achieving in-vessel retention when severe accidents occur. However, an unexploited uncertainty still existed regarding the compound effect of the heterogeneous configuration and dynamic particle redistribution for the debris bed's criticality and cooling safety assessment. Therefore, this research aims to develop a numerical tool for investigating the effects of the different transformations of the heterogeneous configurations on the debris bed's criticality/cooling assessment. Based on the newly proposed methodology in this research, via integrating the Discrete Element Method (DEM) with Computational Fluid Dynamics (CFD) and Monte-Carlo-based Neutronics (MCN), the coupled CFD–DEM–MCN solver was constructed with the originally created interface to integrate two existing codes. The effects of the different bed configurations' transformations on the bed safety assessments were also quantitively confirmed, indicating that the effect of the particle-centralized fissile material had the dominant negative effect on the safety margin of avoiding re-criticality and particle re-melting accidents and had a more evident impact than the net bed-centralized effect. This coupled solver can serve to further assess the debris bed's safety via a multi-physics simulation approach, leading to safer SFR design concepts. [ABSTRACT FROM AUTHOR]

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

A Sodium-cooled Fast Reactor (SFR) is one of the optimized candidates in Generation IV nuclear reactor systems, but safety is an essential issue for SFR development and application. Most knowledge was accumulated through SFR safety investigations, especially for Core Disruptive Accidents (CDAs). During the CDA of SFRs, the molten materials in the core region are likely to discharge into subcooled sodium and form a debris bed on the lower region of the reactor vessel. Noticing that elaboration on the characteristics and mechanisms of Debris Bed Formation (DBF) behavior should be essential for the subsequent analysis of debris bed coolability and accident progression through various experimental and modeling studies, much knowledge was obtained during the past decades. Motivated to promote future investigations on CDAs of SFRs, the previous experiments and modeling studies on DBF behavior are systematically reviewed and discussed in this paper. The experimental results showed that the flow-regime and accumulated-bed characteristics during DBF were influenced by varying parameters and realistic conditions. Through the modeling studies, several empirical models were proposed for predicting the flow regime and accumulated-bed characteristics in DBF. In addition, to promote further development of research, the future prospects concerning DBF behavior are also described. [ABSTRACT FROM AUTHOR]

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.

Thermal striping is a complex thermal–hydraulic phenomenon caused by fluid temperature fluctuations that can also cause high-cycle thermal fatigue to the structural wall of sodium-cooled fast reactors (SFRs). Numerical simulations using large-eddy simulation (LES) were performed to predict and evaluate the characteristics of the temperature fluctuations related to thermal striping in the upper internal structure (UIS) of the prototype generation-IV sodium-cooled fast reactor (PGSFR). Specific monitoring points were established for the fluid region near the control rod driving mechanism (CRDM) guide tubes, CRDM guide tube walls, and UIS support plates, and the normalized mean and fluctuating temperatures were investigated at these points. It was found that the location of the maximum amplitude of the temperature fluctuations in the UIS was the lowest end of the inner wall of the CRDM guide tube, and the maximum value of the normalized fluctuating temperatures was 17.2%. The frequency of the maximum temperature fluctuation on the CRDM guide tube walls, which is an important factor in thermal striping, was also analyzed using the fast Fourier transform analysis. These results can be used for the structural integrity evaluation of the UIS in SFR.

The purpose of this study is to clarify the possibility of designing a small rotational fuel-shuffling breed-and-burn fast reactor (RFBB) with nitride fuel and sodium coolant based on neutronic and heat removal analyses. In these reactor analyses, uranium nitride fuel with a helium bond and sodium coolant was applied to the RFBB, whose thermal power is 450 MW. The structural and cladding materials are oxide dispersion-strengthened ferritic steel. Calculation results showed that the core with rotational fuel shuffling achieved an equilibrium state at criticality near unity, and the average discharge burnup of discharged fuel was 187 MWd/kg heavy metal. In this equilibrium state, reactor characteristics, such as neutron flux and the power profile, were almost stable, and the maximum displacements-per-atom value was slightly higher than 650. A steady-state heat removal analysis was performed for the hottest channel in the core, revealing that the fuel temperature was lower than the operational limit temperature and that the cladding temperature was lower than its melting temperature. However, it was slightly higher than the suggested value of 600°C for retaining nitride fuel integrity for high burnup. It was shown that the core radius could be smaller than that of the metal-fueled core of the previous study. [ABSTRACT FROM AUTHOR]

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