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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 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]

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]

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]

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]

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]

Melting of the nuclear core is one of the severe accident scenarios in a Sodium-cooled Fast Reactor (SFR). During such an event, molten corium may come into contact with the coolant sodium. This interaction of the molten fuel and the coolant is commonly termed molten fuel–coolant interaction (MFCI) in the nuclear industry. In this study, a numerical analysis is carried out to study the solidification of a molten fuel droplet in the liquid sodium pool. In the first part of the study, the effect of constant internal heat generation on the solidification of the droplet is evaluated with convective heat dissipation prescribed at the droplet surface. The internal heat generation (decay power) and the heat transfer coefficient are varied as parameters, and the time required for complete solidification of the molten droplet is obtained. Based on the results, the freezing of the droplet is categorized into three regimes: conduction limited, transition, and internal heat generation dominated regimes. It is observed that the solidification process of nuclear fuel droplets generated during MFCI is not influenced by internal heat generation and lies in a conduction-limited regime for decay power level prevailing in a medium-sized SFR. Hence, in the next part of the study, the numerical analysis is carried out by incorporating the time-dependent decay power and the temperature-dependent heat transfer coefficient in the computational model by developing user-defined subroutines depicting a realistic scenario of an accident. The results of the analysis show that because of the high subcooling of sodium, film boiling is ruled out; nucleate boiling with a maximum heat transfer rate occurs briefly. The heat transfer coefficient then declines as the interface temperature between the droplet and the sodium decreases rapidly until the natural convective regime is reached. A parametric study on the droplet diameter is also carried out by varying the diameter from 0.5 to 10 mm, spanning the typical particle size spectrum expected during MFCI. [ABSTRACT FROM AUTHOR]

• A new solid body formation model in particle method was developed. • The solid bodies in the new model are formed by inter-particle attraction force. • The new model can simulate creation, formation, and motion of solid bodies. • The solid bodies by the new model can diminish and disappear by melting. • The mechanism based on the inter-particle attraction force is common with that in real world. In severe accidents of SFRs, molten core materials can discharge from the core, and the jet can impinge on the lower structure plate. After the jet impingement, fragmented discharged materials can form ring-shape solidification. A fundamental experiment was conducted to simulate the behavior. In order to simulate the behavior of solid body creation and motion, a new solid body formation model by inter-particle attraction force in particle method was developed. The advantage of the new model is that it can simulate creation, formation, and motion of solid bodies without any artificial treatment as solid bodies. The movable solid bodies by the new model have any size, shape, and number, and they are created and grown by solidification, and diminish and disappear by melting. The mechanism based on the inter-particle attraction force is common with that in real world where interatomic attraction force is the cause of solid body formation. [ABSTRACT FROM AUTHOR]

Nuclear energy-producing nations are almost universally experiencing delays in the commissioning of the geologic repositories needed for the long-term isolation of spent fuel and other high-level wastes from the human environment. Despite these problems, expert panels have repeatedly determined that geologic disposal is necessary, regardless of whether advanced reactors to support a "closed" nuclear fuel cycle become available. Still, advanced reactor developers are receiving substantial funding on the pretense that extraordinary waste management benefits can be reaped through adoption of these technologies. Here, the authors describe why molten salt reactors and sodium-cooled fast reactors -- due to the unusual chemical compositions of their fuels -- will actually exacerbate spent fuel storage and disposal issues. Before these reactors are licensed, policymakers must determine the implications of metal- and salt-based fuels vis a vis the Nuclear Waste Policy Act and the Continued Storage Rule. [ABSTRACT FROM AUTHOR]

The VTR is a 300-MW(thermal) sodium-cooled fast reactor (SFR) designed for the specific purpose of delivering unique testing capabilities to enable the advancement of all reactor technologies. With its flux level, irradiation volume, and operational flexibility, the VTR will enable accelerated testing of materials, fuels, and various components needing irradiation testing. Proven SFR technologies and design approaches have been leveraged in designing the VTR core, ensuring the highest possible readiness level. This resulted in the VTR using ternary metallic fuel and delivering fast flux levels in excess of 4 × 1015 n/cm2·s over large useful volumes, corresponding to about 60 dpa/year in steel. As part of the design efforts, the VTR core performance has been determined for a representative configuration, ensuring that the reactivity control systems offer sufficient shutdown margins, that the core can be safely cooled in all situations, and that reactivity feedback coefficients are conducive to a favorable safety behavior. Furthermore, the incorporation of features such as fuel assembly storage in the shield region supports the flexible and reliable operation of the VTR. Additional design work has been ongoing as well. This includes thorough shielding performance evaluations to ensure safe operation of the VTR, verification and validation of the design tools used to achieve compliance with Nuclear Quality Assurance (NQA-1) requirements, early assessment of the impact of irradiation experiments on the core performance envelope and associated margins, and in-depth uncertainty quantification efforts to quantify the anticipated range of performance characteristics. An experimental program supporting the VTR core design has been set up, with the current focus being on thermal-hydraulic experiments. The purpose of this experimental program is to obtain confirmatory measurements to serve directly as part of the core design basis or as part of the validation cases supporting the simulation tools used. [ABSTRACT FROM AUTHOR]

To support the design efforts of advanced sodium-cooled fast reactors (SFRs), a series of computational fluid dynamics (CFD) simulations are performed to investigate the pressure change along various flow passages in the proposed SFR system. The simulations are carried out with the state-of-the-art spectral element flow solver, Nek5000. Two specific case studies are presented in this paper: the flow exiting the axial neutron reflector channels and the flow entering the fuel pin bundle. Due to the high Reynolds numbers expected, a Reynolds-averaged Navier-Stokes (RANS) approach is necessary to model the turbulence. A newly developed regularized k − ω RANS model is adopted in the related CFD calculations. The first case study explores the effect of Reynolds number on the pressure change when flow exits the reflector channels. The pressure change in this case has two major contributors: the change due to wall friction and the Bernoulli effect. It is noted that the nondimensional pressure loss follows a log-linear trend up to Re = 105, and then the trend is flattened. In the second case study, the advanced NekNek coupling capability is tested where an integral domain can be divided into multiple subdomains with coupling interfaces, which would greatly ease the meshing process of complex engineering geometries and potentially save computational resources. The preliminary results obtained so far confirm the consistency between the NekNek results and those produced by regular Nek5000 simulation. The presented work demonstrates the readiness and flexibility of the related CFD techniques, which is part of the broader effort to leverage cutting-edge CFD to inform the advanced nuclear reactor designs. [ABSTRACT FROM AUTHOR]

An ingenious prototype Fuel Assembly Inner DUct Structure (FAIDUS) design is conceptualized under the Generation IV safety philosophy to prevent large-scale corium pool formation following total coolant flow blockage accident. The early failure of the duct wall provides a built-in guideway for molten material relocation out of the core region, ensuring inherent safety robustness. Therefore, understanding the duct wall failure mechanism is critical for evaluating accident mitigation potential of the FAIDUS design. In this regard, a transient enthalpy-based multiphase computational model is developed to investigate the heat transfer and fluid flow behavior involved during the duct wall failure in 3-D domain. The numerical model is employed initially to simulate wall failure in the EAGLE ID1 in-pile test. The model results are validated against the in-pile test data. Subsequently, FAIDUS duct wall failure in a prototype fast reactor fuel subassembly (SA) is investigated employing the present model. The study reveals that the large heat flux from the molten pool to the duct is the main reason for the duct wall failure. The duct failure starts with an initial rupture across the corners of the hexagonal-shaped duct and progresses to total failure of the duct wall by the expansion of the rupture over the whole duct wall area with time. The outer hexcan wall sustains only minor thermal damage during the duct wall failure process. The present study explores the associated thermal-hydraulics, molten pool dynamics, crust formation behavior, and event sequences involved in duct wall failure in great detail. The simulation findings suggest that the FAIDUS SA design facilitates premature failure of the slim duct wall prior to the hexcan wall failure. This could potentially enable early discharge of the molten corium away from the core region, limiting the accident progress. • A comprehensive 3-D enthalpy-porosity based multi-phase thermal hydraulic model is developed. • The model is validated against EAGLE ID1 in-pile experimental data. • The rates of damage progression in inner duct and hexcan wall of a prototype SFR fuel SA due to flow blockage are predicted. • Early failure of the thin inner duct and its efficacy in effective relocation of molten fuel out of active core are studied. [ABSTRACT FROM AUTHOR]

The advanced nuclear reactor industry in North America includes more than 50 companies and labs, which collectively have attracted some $1.3 billion in private capital, as well as government grants and other assistance. Proponents of advanced nuclear reactors say that they are essential to help humans stop heating the planet with carbon dioxide emissions, and that they can do so without the safety, security, and cost concerns posed by older nuclear technology. Detractors say the advanced nuclear industry will never take off, and particularly not without government action that puts a price on carbon dioxide emissions, helping low- and no-carbon energy sources compete economically with fossil fuels. The author interviews company leaders, academics, scientists, and regulators to determine which companies are most likely to succeed. [ABSTRACT FROM PUBLISHER]

• Steam generators (SG) in SFRs are operated at higher temperatures and higher pressure margins than the LWR counterparts. • Other failures, e.g. fatigue, stress corrosive crack, fretting, etc., are far less likely than creep during normal operation. • Creep dominates SFR SG degradationsgiven the high temperature and can be described by Larson-Miller parameter (LMP). • The proposed prognostic model estimates the failure probabilities by sampling LMR and temperature and their uncertainties using Monte-Carlo methods. • The model successfully predicts the failure probability within a given time and suggests how adjusting the workload can protect the component. This paper presents a prognostic model for sodium-cooled fast reactor (SFR) steam generators (SGs). The purpose of the model is to estimate the remaining useful life of SFR SGs and thus to support the decision-making of autonomous control. SFR SGs are of great interest for plant integrity due to their harsh operation environment. They operate at higher temperatures and higher coolant-to-steam pressure differences than those of current light water reactors (LWR). The severity of the SFR SG failure consequences, which include water-sodium contact, is another reason. Understanding its failure mechanisms is important for the development of its prognostic model. Based on our literature review and physics-based analysis, we concluded that creep would be a dominant degradation mode of SFR SGs due to SFR's high-temperature and high-pressure environment. Thus, creep is the focus of the prognostic model. Various other failure modes were also investigated in this study. The mechanical fatigue due to flow-induced vibration is usually observed in early developmental phases and would not be an important issue during normal operation. The thermal fatigue due to thermal stripping is occasionally observed in other components in SFRs but does not affect SG integrity. Pure water stress corrosive crack, fretting, and so on are commonly observed in LWR SGs but are significantly less important in SFR SGs because of the high temperatures, high pressure differences, and chemical properties of liquid sodium. Based on these investigations, a prognostic model focusing on creep failures was developed. It estimates the failure probability profile by sampling the Larson-Miller parameter (LMP) and temperature and associated uncertainties through Monte Carlo methods. Two case studies were presented. The first one demonstrated the model's capability to calculate the failure probability for a new specimen within a given time. If the tube is working under temperatures of 500C ± 3 and pressures of 15.2 MPa, which leads to LMP of 20,100 ± 50, the failure probabilities within 50, 70, and 90 years are approximately 0.1 %, 3.2 %, and 17.5 %. The second demonstrated how adjusting the workload can help to protect the integrity of the component. For an old tube reaching 99 % of its lifespan, continuing to run at temperatures of 500C ± 3 and LMP of 20,100 ± 50 leads to a failure probability of about 29.0 % within a year. If the temperature is reduced by 5C, the failure probability can be reduced to 2.5 %, and, if the pressure is also reduced such that the LMP is increased by 100, the failure probability can further be reduced to 0.12 %. [ABSTRACT FROM AUTHOR]

In this work, a Rotational Fuel-shuffling Breed-and-Burn fast reactor with nitride fuel and sodium coolant (RFBB-NS) was analyzed. The purpose of the study was to design an RFBB-NS, including assembly duct and control rod assemblies, and to show the feasibility. Monte Carlo simulation code Serpent was used for neutronic and fuel burnup analyses and COMSOL Multiphysics was used for heat removal analysis. The results showed that the practical design of RFBB-NS cannot reach criticality due to the higher presence of absorbing materials such as assembly ducts. Therefore, the core was modified to increase the neutron economy by applying a lead–bismuth reflectors. This allowed the reactor to reach criticality and an equilibrium state after fuel shuffling. Afterward, burnup performance analysis showed that the average fuel discharge burnup was ∼ 298 MWd/kgHM and, that the Pu concentration in the discharge fuel was lower than in previous studies. However, the radiation damage to the cladding material was slightly higher than the constraint value in this study. [ABSTRACT FROM AUTHOR]

Hydraulic Suspended Passive Shutdown Subassembly (HS-PSS) is a passive nuclear safety technology employed in Sodium-Cooled Fast Reactors (SFR). During Sodium Fast Reactors (SFR) operation, this component is often subjected to stress cycles caused by flow-induced vibration (FIV), which may lead to material failure and potentially threaten the reactor's safety. Therefore, it is necessary to investigate and evaluate the FIV behavior of the HS-PSS in fast reactors. Different from the conventional used before, HS-PSS is a new type of control rod assembly with no existing studies on FIV. In this study, we established a full-scale experiment for FIV of the HS-PSS in different flow conditions and with various movable component positions (600 mm, 800 mm, 1200 mm). Modal experiments were conducted to determine the natural frequencies. FIV test results show that the highest amplitude occurs when the movable component is 800 mm, with a noticeable acceleration response near the first natural frequency. To explain the experimental observations, numerical simulations have been conducted to analyze the HS-PSS flow field characteristics. The simulations have considered different positions of the movable component at a fixed flow rate of 16.281 m3/h and different flow rates with the movable component positioned at 800 mm. The results indicate significant changes in the flow distribution within the HS-PSS and the hydraulic forces applied to the inlet surface of the movable component when the movable component is set at 800 mm. The experimental and numerical analysis results of this study have provided valuable insights for analyzing and assessing the FIV behavior of the HS-PSS in Sodium-Cooled Fast Reactors. • A test bench was constructed to study the flow-induced vibration of a Hydraulic Suspended Passive Shutdown Subassembly. • Experiments have discovered the location of special movable parts that respond significantly to vibrations. • The flow field simulation results are consistent with the experimental phenomena, explaining the cause of the vibration response. [ABSTRACT FROM AUTHOR]

• Demonstration plant of sodium-cooled fast reactor. • A design and lifecycle simulation/evaluation system. • Safety design and evaluation methods. • Codes and standards taking into account transformation to risk-informed and performance-based framework. • Separation and transmutation of minor actinides in nuclear fuel cycle. A sodium cooled fast reactor (SFR) is one of the most relevant and decarbonized energy supply system with higher sustainability on natural resources, footprint, and waste management. It was planned in a "strategic roadmap" of fast reactor decided by Inter-Ministerial Council for Nuclear Power Japan in 2022 to start a conceptual design of a demonstration reactor from 2024 with a background of accumulated knowledge and experiences of SFR development. For example, a design and lifecycle simulation/evaluation system named ARKADIA has been developed to accelerate such design works. It will enable to take into account plant lifecycle, e.g., operation and maintenance, to the plant design and optimize it based on simulations and knowledgebase. This paper shows research progresses of ARKADIA, safety design and evaluations, codes and standards, fuel cycle, and SFR development projects in Japan. [ABSTRACT FROM AUTHOR]

• A single high-temperature strength equation expressing the mechanical strength in different deformation and rupture modes was derived for 9Cr-ODS TMS cladding tubes. • This equation can predict the rupture life of the cladding tubes under various stresses and temperatures over time. • The reason why the equation can be applied to different deformation and rupture modes is considered to be the effect of the fine-grain matrix of 9Cr-ODS TMS. • The study suggested the thermal activation process is dominant even for the high-temperature deformation modes exceeding yield stress in the 9Cr-ODS TMS with the fine-grained matrix. This paper discusses the applicability of Straalsund et al.'s technique for combining the Larson–Miller parameter (LMP) and life-fraction rule to form a single high-temperature strength equation for 9Cr-oxide-dispersion-strengthened (ODS) tempered martensitic steels (TMS). It uses the extensive dataset on creep rupture, tensile, and temperature-transient-to-burst tests of 9Cr-ODS TMS cladding tubes in the α-phase, α/γ-duplex, γ-phase matrices, which are accumulated by the Japan Atomic Energy Agency so far. The technique is adequately applicable to 9Cr-ODS TMS cladding tubes. A single high-temperature strength equation expressing the mechanical strength in different deformation and rupture modes (creep, tensile, temperature-transient-to-burst) is derived for 9Cr-ODS TMS cladding tubes. This equation can predict the rupture life of the cladding tubes under various stresses and temperatures over time. The applicable range of the high-temperature strength equation is specified in this study and the upper limit temperature for the equation is found to be 1200 °C. At temperatures higher than 1200 °C, the coarsening and aggregation of nanosized oxide particles and the γ to δ phase transformation are reported in previous studies. The high-temperature strength equation can be well applied to the creep and tensile strength in the α-phase matrix, the creep strength in the γ-phase matrix and the temperature-transient-to-burst strength in both phases except for the low equivalent stress (43 MPa) at temperatures exceeding 1050 °C. The mechanism of the notable consistency between creep and tensile strength in the α-phase matrix is discussed by analyzing the high-temperature deformation data in the light of existing deformation models. The study suggested the thermal activation process is dominant even for the high-temperature deformation modes exceeding yield stress in the 9Cr-ODS TMS with the fine-grained matrix. [ABSTRACT FROM AUTHOR]

HT9 cladding was manufactured through successive cold working and heat treatments, and its creep strains were obtained at various stresses and temperatures in the range of 10–120 MPa and 843–953 K, respectively, for up to 20,000 h. Primary and secondary creep regimes were observed, whereas only a few specimens at extreme conditions experienced a tertiary creep regime and creep rupture. The reference creep models, including the theta projection and modified Garofalo equation for HT9 cladding, did not fit the experimentally obtained creep strain curves; the manufactured HT9 cladding showed higher creep resistance than the reference models. Hence, a new correlation was developed by deriving various constants, such as the activation energy, stress exponent, and material constants, from the measured creep strains. The creep correlation successfully predicted the creep behavior of HT9 cladding manufactured. Obtained extensive creep strains over a range of stresses and temperatures, closely approximate actual operating conditions, and developed correlation in this study are valuable in various applications using HT9, providing essential data for analyzing and estimating creep resistances of cladding during the operation of nuclear reactors. [ABSTRACT FROM AUTHOR]

Thermally-induced reaction between liquid sodium (Na(l)) and calcium hydroxide (Ca(OH)2(s)) was investigated as one of the possible component reactions when Na(l) was reacted with concrete materials under a postulated severe accidental condition in a sodium-cooled fast reactor. The Na(l)–Ca(OH)2(s) reaction was traced using a differential scanning calorimetry (DSC), placed in an argon substituted glove box. An exothermic DSC peak appeared in the temperature range of 550–700 K was attributed to the Na(l)–Ca(OH)2(s) reaction to form a mixed phase comprised of calcium oxide (CaO(s)), sodium hydroxide (NaOH(s,l)) and sodium oxide (Na2O(s)). Based on morphological analyses of reacting system, a physico-geometrical reaction model was proposed: the reaction initiates at the initial contact area of Na(l) and Ca(OH)2(s) and proceeds by the movement of the reaction interface toward Ca(OH)2(s), where the product layer is composed of CaO(s), NaOH(s,l), and Na2O(s). Therefore, the Na(l) diffusion through the product layer is the necessary process to promote the reaction between Na(l) and Ca(OH)2(s), for which a significant influence of the melting of NaOH(s) that has on the overall kinetic behavior is expected. This melting occurs midway through the Na(l)–Ca(OH)2(s) reaction at 594 K. The overall Na(l)–Ca(OH)2(s) reaction was analyzed kinetically using the DSC curves recorded at different heating rates. A partially overlapping two-step reaction feature was evidenced, in which the primary and secondary reaction steps were characterized by the apparent activation energy values of 128 and 138 kJ mol−1, respectively. The multistep feature of the Na(l)–Ca(OH)2(s) reaction could result from the melting of NaOH(s) in the product layer. [ABSTRACT FROM AUTHOR]

The objective of the present study is to show that it is feasible to establish the breed-and-burn (B&B) mode of operation with rotational fuel shuffling in the S-PRISM core based on neutronic and thermal-hydraulic analyses. The results quantified the impact of major core design choices on the criticality of a core that uses sodium as a coolant material and HT9 steel as structural material. The design variables examined include the binary metallic fuel U-Zr with different weight percentages of zirconium as well as different core heights and fuel rod pitch-to-diameter ratios (P/Ds) in the fuel assembly. We found that a core using the binary metallic fuel U-Zr with 2 wt% zirconium, with a core height of 200 cm, a P/D of 1.086, and a core power of 400 MW(thermal), could overcome some major design constraints. It was also found that with shuffling intervals of 1125 to 1250 days, the core with rotational fuel shuffling was critical in the equilibrium state, and the possible average discharged burnup was from 274.8 to 305.3 GWd/ton HM. Reactor characteristics such as neutron flux and power profile were almost stable during the equilibrium cycle. A steady-state thermal-hydraulic analysis was performed for the hottest channel in the core. It revealed that both the fuel and cladding maximum temperatures were less than the melting point of the fuel and the chemical interaction temperature of the HT-9, respectively. The mixed coolant outlet temperature was somewhat below the temperature usually observed in sodium-cooled fast reactors. Thus, it appears that the S-PRISM can be principally designed to have a B&B core. [ABSTRACT FROM AUTHOR]

A one-dimensional (1-D) thermal stratification (TS) model was recently developed in our research group to predict the TS phenomenon in pool-type sodium-cooled fast reactors. This paper performs uncertainty quantification (UQ) of the 1-D TS model to evaluate its performance by considering the aleatoric uncertainties that existed in the model parameters and to identify the plausible sources of the epistemic uncertainties. The Latin hypercube sampling-Monte Carlo method (LHS-MC), which is elaborated with an example in this paper to facilitate its understanding and implementation, is used for the UQ process. The advantages of LHS-MC, including both better stability and better accuracy than the conventional random sampling-Monte Carlo method with fewer realizations, are demonstrated in this paper. In total, 648 temperature measurements acquired from nine experimental transients performed in a university-scale Thermal Stratification Experimental Facility are used to evaluate the performance of the computational 1-D TS model. The UQ result shows that 77.5% of the experimental data can be predicted by the 1-D TS model within uncertainty ranges, which indicates the good performance of the computational model when the aleatoric uncertainties are correctly captured. The rest 22.5% of the experimental data are found located outside of the uncertainty ranges, which reveals the existence of the epistemic uncertainties caused by the lack of understanding of the TS phenomenon and defects in the 1-D model. The simple jet model currently employed by the 1-D TS model is thought to be one of the attributors to these defects. [ABSTRACT FROM AUTHOR]

This work presents a demonstrational application of genetic algorithms (GAs) to solve sample optimization problems in the generation IV nuclear reactor core design. The new software was developed implementing novel GAs, and it was applied to show their capabilities by presenting an example solution of two selected problems to check whether GAs can be used successfully in reactor engineering as an optimization tool. The 3600 MWth oxide core, which was based on the OECD/NEA sodium-cooled fast reactor (SFR) benchmark, was used a reference design [1]. The first problem was the optimization of the fuel isotopic inventory in terms of minimizing the volume share of long-lived actinides, while maximizing the effective neutron multiplication factor. The second task was the optimization of the boron shield distribution around the reactor core to minimize the sodium void reactivity effect (SVRE). Neutron transport and fuel depletion simulations were performed using Monte Carlo neutron transport code SERPENT2. The simulation resulted in an optimized fuel mixture composition for the selected parameters, which demonstrates the functionality of the algorithm. The results show the efficiency and universality of GAs in multidimensional optimization problems in nuclear engineering. [ABSTRACT FROM AUTHOR]

In the steam generator of a sodium-cooled fast reactor, heat transfer occurs from the primary sodium flow side at high temperature to the secondary water flow side at low temperature. Preventing the interaction between sodium and water in the heat exchanger is important to avoid a reactive explosion. To do so, the concept of the copper bonded steam generator, which consists of heat-exchanger modules with cross-flow-type arrangement, was recently proposed. In this cross-type arrangement, the sodium flow should be as uniform as possible throughout the horizontal channels in terms of the heat transfer efficiency between the sodium and water sides. This study aims to design a perforated plate leading to uniform sodium flow distribution throughout the 66×33 channels. The geometrical design and location of the perforated plate are optimized using numerical simulation. The designed plate improves the flow distribution uniformity at a slight cost of pressure loss. [ABSTRACT FROM AUTHOR]

Summary: Closed type assembly leads to complicated core thermohydraulics which are significant for design and safety analysis in sodium‐cooled fast reactor (SFR). Most of the major phenomena can be dealt with by general parallel channel network and two‐dimensional model for inter‐wrapper flow (IWF) in system code transient thermal‐hydraulic code for analysis of SFR (THACS). A main deficiency left is the overestimation of inter‐assembly heat transfer effect due to lump temperature message that one‐dimensional assembly model can only provide. A methodology was thus developed to compute wall‐subchannel temperature for simulating energy transportation through duct wall. Intra‐assembly thermohydraulics were classified as two mechanisms and non‐dimensional parameter was introduced for characterizing each of them. Ratio between linear heat flux density through hexagonal wrapper surface and pin surface qw/p was employed to account for radial heat transfer. Flow redistribution effect arouse in mixed and natural convection was concretized by a importance symbol of buoyancy force, Richardson number Ri. Two criteria were also proposed for dividing flow state into three parts and specific model for each section was defined. Two validation cases against 3‐D computational fluid dynamics (CFD) simulation of China experimental fast reactor whole core and test data from plant dynamics test loop with direct heat exchanger experiment were performed with THACS. Noticeable improvement after incorporation of this model demonstrated its value in system analysis. Although the procedure for parameter computations is geometry related, this method can be reproduced to any type of assembly directly. [ABSTRACT FROM AUTHOR]

Summary: To evaluate residual heat removal capacity of next generation China sodium‐cooled fast reactor (SFR) and provide data for code validation, an integral natural circulation experiment was performed on a scaled water platform which had same structure with reactor primary system. A modular code THACS focusing on safety analysis of SFR was validated by this test. Some necessary development and verification for physical property, frictional loss, and heat transfer of water were carried out first. Whole simulation of an unprotected loss‐of‐flow (ULOF) case was divided into two phases and computation lasted until 5000 s of test time during second stage. Agreement in light of system temperature distribution in first stage provided a suitable initial condition for subsequent transient. Satisfied flow rate curve was reproduced with decaying pressure head and valve characteristic as input when pump was stopping. All natural circulation paths were predicted, including original flow in primary loop through main pipe and reversed flow in reactor vessel cooling system (RVCS). Transition to operative mode of direct heat exchanger (DHX) in direct reactor auxiliary cooling system (DRACS) and its cooling power were also forecasted successfully. This work means a significant advance for THACS in the field of engineering application of China SFR. [ABSTRACT FROM AUTHOR]

Summary: Accuracy and computational expenses are contradictory when performing simulation to hot plenum where multi‐dimensional phenomena voice loudly in sodium‐cooled fast reactor (SFR). A 2‐dimensional method was preferred in light of the cylindrical shape and drive force behind significant thermal stratification phenomena during transient case. One barrier related to buoyancy effect on turbulent flux transportation has been removed before. Asymmetric arrangement of components in circumferential direction makes another difficulty to reproduce actual flow field within hot plenum of pool‐type SFR. A series of special techniques for spatial discretization were implemented to restore geometry reality. Treatments about upper core structure, main heat exchanger and auxiliary heat exchanger were evaluated based on criterion of predicting well‐mixed inner plenum and stratification like outer plenum. Finally, optimal model came to an 8‐block partition scheme in which whole profiles of upper core structure and auxiliary heat exchanger were described. Validation was carried out through a normal state experiment on a water platform with reduced‐scale structure. Agreement of relative temperature distribution demonstrated ability of this model for tackling spatial layout in pool‐type SFR with the 2‐dimensional library. All real‐physics and efficient features indicate that this 2‐dimensional methodology is an excellent tool for engineering application when designing or evaluating a new SFR concept. [ABSTRACT FROM AUTHOR]

Thermal stratification (TS) is a thermal-fluid phenomenon that can introduce large uncertainties to nuclear reactor safety. The stratified layers caused by TS can lead to temperature oscillations in the reactor core. They can also result in damages to both the reactor vessel and in-vessel components due to the growth of thermal fatigue cracks. More importantly, TS can impede the establishment of natural circulation, which is widely used for passive cooling and ensures the inherent safety of numerous reactor designs. A fast-running one-dimensional (1-D) model was recently developed in our research group to predict the TS phenomenon in pool-type sodium-cooled fast reactors. The efficient 1-D model provided reasonable temperature predictions for the test conditions investigated, but nonnegligible discrepancies between the 1-D predictions and the experimental temperature measurements were observed. These discrepancies are attributed to the model uncertainties (also known as model bias or errors) in the 1-D model and the parameter uncertainties in the input parameters. In this study, we first recognized through a forward uncertainty analysis that the observed discrepancies between the computational predictions and the experimental temperature measurements could not be explained solely by input uncertainty propagation. We then performed an inverse uncertainty quantification (UQ) study to reduce the model uncertainties of the 1-D model using a modular Bayesian approach based on experimental data. Inverse UQ serves as a data assimilation process to simultaneously minimize the mismatches between the predictions and experimental measurements, while quantifying the associated parameter uncertainties. The solutions of the modular Bayesian approach were in the form of posterior probability density functions, which were explored by rigorous Markov Chain Monte Carlo sampling. Results showed that the quantified parameters obtained from the inverse UQ effectively improved the predictive capability of the 1-D TS model. [ABSTRACT FROM AUTHOR]

• Past findings on debris bed behaviors are summarized for enhanced SFR safety. • Debris bed formation & self-leveling mechanisms and characteristics are clarified. • Study at more realistic conditions and future development direction are prospected. Understanding the Core Disruptive Accident (CDA) is crucial for the safety analyses of Sodium-cooled Fast Reactors (SFR), despite its exceedingly low possibility. During a CDA, the potential release and subsequent relocation of molten materials from the core could result in the formation of a debris bed inside the reactor vessel due to their quenching and fragmentation from the interactions with the subcooled sodium coolant. Debris beds with varying shapes are expected to form due to the different flow-regime characteristics during their formation. Further, the heat decay from this high-temperature debris might trigger sodium coolant boiling, instigating the debris bed self-leveling behavior, which flattens the debris beds. Considering the importance of the debris bed shapes to the debris transfer, heat removal and neutronic criticality, lots of knowledge about the mechanisms and characteristics of debris bed formation and self-leveling behaviors has been gathered in recent years through extensive experimental, modeling and numerical outcomes. This study aims to systematically and critically summarize these past explorations into the debris bed formation and self-leveling behaviors, offering a beneficial direction for future research on the improved safety evaluations in SFR severe accidents. [ABSTRACT FROM AUTHOR]

[Display omitted] • GRESS code is developed for analyzing fission gas release and swelling in SFR. • Grain size and gas bubble diameter exhibit rapid variations within a specific range of fuel temperature. • The relatively minimal variation in grain size and gas bubble diameter at higher fuel temperatures result in a decrease in gas bubble swelling. • FGR and intragranular gas bubble swelling increase over time during transient operation. • The results demonstrate the coupling and interaction of fission gas behavior with each other. In order to simulate the behaviors of Fission Gas Release (FGR) and swelling for Sodium-cooled Fast Reactor (SFR) under normal and transient conditions, a G as RE lease and S welling S imulations (GRESS) code is developed using COMSOL Multiphysics. GRESS is mainly based on a previous model called the Operational Gas RElease and Swelling (OGRES) model, in which the FGR and swelling are primarily related to burnup, time and temperature. GRESS assumed that there is a spherical grain at the fuel pellet. The grain is presented by a 2D axisymmetric geometry in COMSOL Multiphysics. Fission gas atom generation, gas atom diffusion, gas atom diffusion into the bubble, bubble resolution, gas bubble nucleation, gas bubble migration to the grain boundary, bubble pressure, gas bubble swelling, grain growth, oxygen to metal atomic ratio, fuel porosity, fuel thermal conductivity, heat generation rate, pore velocity, pore concentration, plutonium migration and so on are considered in GRESS code. The output results include grain size, gas bubble diameter, gas bubble concentration, the amount of gas atoms per bubble, gas atom diffusion coefficient, bubble diffusion coefficient, surface diffusion coefficient (varying with temperature), FGR (varying with time), and swelling (varying with time and radius). The comparison of the results with other modeling outcomes and experiments is performed, which shows a good agreement. [ABSTRACT FROM AUTHOR]

• The thermal stress induced cracking of nuclear fuel droplet fragmenting in sodium is studied. • Fracture Bifurcation Theory (FBT) and Minimum Potential Energy Principle (MPEP) approaches are adopted for the analysis. • Parametric study carried out for different droplet sizes, crust thickness and cooling conditions. • The minimum size of the droplet prone to catastrophic failure is found out using FBT. • Fine fragment detached from the solidified crust due to thermal stress cracking evaluated using MPEP. During Molten Fuel Coolant Interaction (MFCI) occurring in a Sodium-cooled Fast Reactor (SFR) following a severe accident, the corium jet breaks up into coarse droplets due to hydrodynamic instabilities and fine flaked debris are generated due to the thermal stress cracking occurring on the solidifying droplet. In particular mixed oxide core debris quenched by liquid sodium are more prone to thermal stress cracking because of the excellent heat transport properties of liquid sodium, which leads to a large temperature gradient across the crust of the ceramic fuel. In this paper, thermal stress initiated cracking of solid crust of the nuclear fuel droplet is studied using Fracture Bifurcation Theory (FBT) and Minimum Potential Energy Principle (MPEP). FBT is able to predict the minimum crack length needed for crack propagation or catastrophic failure depending on the computed K I values. On the other hand, MPEP can predict the periodic and hierarchical cracking pattern that can develop on the core debris. This helps in the prediction of number of cracks and the crack length that can develop on the solid crust. Parametric studies have been carried out by varying MOX droplet sizes, crust thickness and cooling conditions and calculating K I by semi-analytical means and comparing it with K IC. The analysis revealed that solidifying droplets with radius equal to or more than 2 mm are susceptible to thermal stress initiated crack growth and catastrophic breakage even if a few tens of micron size initial crack develops on its surface. Then a particular case of 2 mm radius core debris is considered and its cracking pattern is studied by applying MPEP. This optimal crack configuration is obtained by minimizing potential energy for a given range of crack length and spacing. The analysis has been carried out with ANSYS APDL code which calculates the strain energy for a given initial crack length and spacing. The 2D computational model is validated with a similar study carried out for Alumina disc quenched in water and the results are in good agreement. The minimum fine fragment size predicted by the analysis for mixed oxide debris is a flake like debris with equivalent radius of 41 µm when approximated to be a sphere. The particle size correlates well with the fine debris sizes reported in literature of MFCI experiments. [ABSTRACT FROM AUTHOR]

Thermal stratification is an important phenomenon in the process of natural circulation decay heat removal in sodium-cooled fast reactors (SFR). It affects the establishment of the natural circulation and the flow rate of the natural circulation in a station black out (SBO) accident, and therefore affects the safety of the reactor. In safety evaluation, programs are often required to perform analysis for various conditions and long-term transients. However, it is difficult to meet this requirement by adopting a multi-dimensional method or one-dimensional and multi-dimensional coupling method. Therefore, this paper proposes applying a thermal stratification model to the system analysis program THACS to analyze the thermal stratification phenomenon. First, the thermal stratification model is validated with experimental data. The validated model is applied to analyze a station black out accident in the SFR. The lumped parameter model and thermal stratification model are used respectively for the hot pool. By comparing the calculation results of the two models, the effect of thermal stratification of the hot pool on key parameters such as the natural circulation flow rate of the primary circuit, natural circulation flow rate in the inter-wrapper region, average core outlet temperature, maximum core fluid temperature, and the maximum temperature of cladding was analyzed, and the effect of thermal stratification on natural circulation decay heat removal of the SFR was evaluated. This work can provide a reference for the safety analysis of the reactor and contribute to developing a more conservative safety analysis and evaluation strategy. [ABSTRACT FROM AUTHOR]

For accurate assessment of nuclear heating in fast reactors, a new coupled neutron and gamma heating calculation scheme has been developed based on VARIANT nodal transport solutions of neutron and gamma flux distributions. The MC2-3 code was extended to generate multigroup neutron and gamma cross sections and kinetic energy release in materials (KERMA) factors, and a utility program CURVE was developed to reconstruct detailed pin and duct wall powers from VARIANT output files. The improved heating calculation scheme has been verified against MCNP6 Monte Carlo reference solutions for the Advanced Burner Test Reactor (ABTR) and Experimental Breeder Reactor II (EBR-II) benchmark problems. Compared to the existing coupled heating calculation method based on DIF3D diffusion theory solutions, the new heating calculation scheme utilizes more accurate gamma cross sections and KERMA factors, accounts for the transport effects, and eliminates the approximations in the existing pin power reconstruction scheme. As a result, it produces more accurate assembly and pin power distributions. For both the ABTR and EBR-II problems, the maximum assembly power error was ~1% in fuel assemblies and ~2% in instrumented structure assemblies, and the maximum error in pin segment powers in an axial node of fuel assembly was ~4%. In the blankets of the EBR-II problem, the maximum error in pin segment powers was increased to ~8%, mainly due to the lower power level and the relatively large error in the nodal power of the VARIANT solution. [ABSTRACT FROM AUTHOR]

• Flow regime in debris bed formation behavior with mixed-size particles is studied. • Good predictability of base model using area mean diameter is confirmed. • Past non-spherical particle extension scheme cannot provide satisfactory results. • New extension scheme for mixed-size particles is suggested and validated. • Different component inertia may cause separation/stratification in reactor accident. During a core disruptive accident of sodium-cooled fast reactors, the fragmented debris released from the reactor core will start to sediment and form debris beds within the lower plenum region of the reactor vessel. To ascertain the characteristics of flow regimes during this debris bed formation process, in our earlier publications based on a great number of experiments conducted using mono-sized spherical and non-spherical solid particles, four typical flow regimes (i.e. the particle-suspension regime, the pool-convection dominant regime, the transitional regime and the particle-inertia dominant regime) were identified. In addition, a flow-regime model, including the base model for predictions of spherical particles and the extension scheme for non-spherical particles, has been successfully developed. In this study, motivated to further understand this behavior at a condition of more realistic particle mixture, a new series of simulated experiments using mixed-size particles is conducted under the gravity-driven release from a nozzle into a static two-dimensional water pool. Through detailed analyses, it is found that the base model, by employing the area mean diameter, can predict the overall regime transition under the condition of mixed-size particles reasonably well. Even if a quantity of effective sphericity is introduced, the previous extension scheme for non-spherical particles lacks the capability to provide satisfactory predictions. To improve the prediction using the form of extension scheme, a new correction scheme based on the volume-mean diameter and the degree of convergence in particle-size distribution is suggested and validated. The conducted analyses in this study also indicate that due to the difference in component inertia, the potential separation and stratification of solid particles are likely to exist during an actual debris bed sedimentation process in a core disruptive accident. [ABSTRACT FROM AUTHOR]

• A robust multi-phase thermal hydraulic model for molten fuel relocation is developed. • Movement of molten fuel in a sodium cooled fast reactor subassembly is analysed. • Penetration length of melt front in sub-channels of fuel subassembly is determined. • Amount of fuel relocated out of core before melt front freezing is quantified. • Design measures to enhance fuel relocation to the core catcher are proposed. A transient, multi-phase enthalpy based computational model is developed to analyse the streaming and freezing characteristics of the molten fuel in a blocked fuel subassembly (SA) during Total Flow Blockage (TFB) at the foot of the SA. During TFB accident progression, a molten material pool is formed following the meltdown of fuel pins. The molten fuel has the potential to relocate axially through the available subchannels in the downward direction. The mathematical model comprises of two modules; a Fluid Dynamics module and a Heat Transfer module. The model adopts both implicit and explicit type of finite difference techniques employing a variable grid system for boundary tracking. The mathematical model has been validated against benchmark experimental data and SIMMER III code prediction reported in open literature. The developed model is employed to simulate the relocation of molten fuel in a prototype fast reactor fuel SA under accident condition. The simulation results indicate that the molten fuel is unable to penetrate the axial blankets in fuel SA without freezing and forming a blockage mainly owing to the rapid freezing of melt. Further, a parametric analysis is carried out to ascertain the influence of critical parameters on the relocation and freezing characteristics. It is observed that the design of the fuel SA has a potential effect on the streaming behaviour and enhancement in fuel relocation is substantially achieved by dwindling the fuel pin size, especially axial length and diameter of the pin, thus preventing the accident progression to the neighbouring SAs. [ABSTRACT FROM AUTHOR]

• Super-COPD code is being developed for design study and safety evaluation of SFR. • Plant dynamics analysis was performed for FFTF Loss of Flow Without Scram test #13. • Multi-channel in core and bowing reactivity models enhanced prediction accuracy. • Reactivity and thermal-hydraulics models of Super-COPD for LOFWOS were validated. 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. [ABSTRACT FROM AUTHOR]

• Numerical simulation of debris bed formation is performed using MPS-DEM method. • High-order MPS method is applied with DEM model for the accurate simulations. • Applicability and reliability of the method are validated by experimental results. • Liquid pool convection induced by falling particles is confirmed in simulations. • Fuel debris bed formation in sodium is simulated to study for CDA analysis of SFR. During a Core Disruptive Accident (CDA) in a Sodium-cooled Fast Reactor (SFR), debris beds may form in the lower plenum of the reactor vessel as a result of the accumulation of discharged debris, which is formed from the solidification and fragmentation of core molten materials during their release due to interactions with the sodium coolant. Since the configuration of debris beds plays a vital role in their long-term coolability and neutronic subcriticality for achieving in-vessel retention, it is essential to understand and assess the Debris Bed Formation Behavior (DBFB). In this study, an advanced Moving Particle Semi-implicit (MPS) method, called least square MPS (LSMPS), is utilized and coupled with the Discrete Element Method (DEM) for the numerical investigations of DBFB characteristics in a two-dimensional liquid pool. The applicability and reliability of the proposed LSMPS-DEM method are validated by comparing simulation results with experiments using simulant materials. The LSMPS-DEM method is then employed to study DBFB in sodium. It is observed that debris properties significantly impact the sedimentation and accumulation processes of DBFB, leading to variations in the characteristics of the formed debris beds. The present study is expected to be useful for further investigations of DBFB under more realistic conditions and for the design improvement of safety devices to mitigate the consequences of CDA in SFR. [ABSTRACT FROM AUTHOR]

• Behavior of in-place cooling of residual core fuel in SFRs was analyzed by SIMMER-III. • Phenomena in the in-place cooling were identified by PIRT. • Out-of-pile experiment simulating the in-place cooling was conducted. • Out-of-pile experiment was analyzed by SIMMER-III. • New model of gas mixing was developed and improved the numerical analysis. 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. [ABSTRACT FROM AUTHOR]

• Validation of THACS and OpenMC for SFR has been enhanced by simulating FFTF LOFWST Test #13. • A new modeling method multi-parallel channel plus local subchannel model is proposed for reactor core. • Important local phenomenon like inter wrapper flow, thermal stratification is analyzed by the specific model. 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. [ABSTRACT FROM AUTHOR]

Understanding the Debris Bed Formation (DBF) behavior is important to the optimized integral evaluation for Core Disruptive Accident (CDA) of Sodium-cooled Fast Reactor (SFR). Aimed at clarifying the characteristics of flow regimes during DBF, extensive experimental works were carried out by releasing solid particles into a water pool. According to the preliminary studies using bottom heating and nitrogen gas injection, it was found that the effect of coolant boiling should be crucial to the flow-regime transition. However, the gas flow rate considered in these studies (0–50 L/min) was not adequate wide to cover the range of vapor flow rate from the possible drastic boiling in actual CDA of SFR. Therefore, in current study, to elucidate the influence of coolant boiling on DBF flow-regime characteristics more comprehensively and provide more extensive data for supporting the predictive model development, a greater number of experiments are conducted with a much-enlarged range of gas flow rates (up to 100 L/min) under varying parametric conditions (including particle size, particle type, water depth and release pipe diameter). Through analyzing the flow-regime transition, an extended functional form for the effect of coolant boiling is successfully proposed. By incorporating this form into our base model, the model predictability can be extended for the DBF at the conditions of coolant boiling induced by accumulated debris in accordance with the verification of the regime map. Present study is believed to be valuable for further developing and verifying the models and codes for SFR safety analysis in China. • Experiments of debris bed formation in CDA of SFR are conducted with gas injection. • Qualitative observations and varying quantitative parametric analyses are performed. • Effect of coolant boiling induced by accumulated debris is studied and confirmed. • An empirical model extension function is developed based on extensive experiments. • The predictability of the extended model is validated through the regime map. [ABSTRACT FROM AUTHOR]

The strong coupling heat transfer between subassemblies and inter-wrapper gaps is closely related to the safety of the sodium-cooled fast reactor (SFR). Therefore, it is very important to carry out a detailed thermal hydraulic analysis of the core. In this paper, the inter-wrapper channel is modeled using the subchannel analysis method. A core thermal-hydraulic analysis code SACOS-Na (Subchannel Analysis Code Of Safety for SFR) is developed by using SIMPLE algorithm. For the subassembly module, a total of three experiments, including ORNL 19-pin experiment, TOSHIBA 37-pin experiment and KNS 37-pin experiment, are adopted to demonstrate the feasibility of the mathematical and physical models and numerical algorithms of the SACOS-Na as well as the rationality of the selected auxiliary models. For the inter-wrapper gap module, WEISS 2-subassembly experiment and CCTL-CFR 3-subassembly experiment are used to verify not only the simulation capability for backflow and strong inter-subassembly heat transfer, but also the heat structure model and the subchannel modeling approach on inter-wrapper gaps. In addition, the prediction accuracy of SACOS-Na has reached the advanced international level by comparing with the worldwide well-known core thermal hydraulic analysis codes. • The subchannel modeling method is adopted to build the mesh for inter-wrapper channels. • The SIMPLE algorithm is used to solve the problems of the reverse flow and inter-subassembly heat transfer occurred in SFRs. • The universal core thermal-hydraulic analysis code SACOS-Na for SFRs is developed. • The accuracy and reliability of the SACOS-Na are discussed from all aspects. [ABSTRACT FROM AUTHOR]

• Severe accident code, AZORES, is coupled with plant dynamics code for SFRs, ADYTUM. • The coupling is performed through LOHS and LORL simulations. • Accident progression between coupled and single AZORES simulations differs significantly. Accident scenarios for the safety evaluation of sodium-cooled fast reactors (SFRs) are categorized into two groups: Anticipated Transient Without Scram (ATWS) and Loss-of-Heat Removal System (LOHRS). The latter is subdivided into two scenarios—Loss-of-Heat Sink (LOHS) and Loss-of-Reactor Level (LORL)—depending on the initiating event. With the development of SFRs, research on ATWS has been paid much attention to due to its rapid accident progression, which could lead to significant radioactive release. Conversely, the significance of LOHRS has increased in recent decades due to lessons learned from accidents at the Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Company. However, compared with ATWS, finding and evaluation tools for LOHRS have been limited. Hence, code development is initiated to enhance the severe accident analysis code AZORES, which the Nuclear Regulation Authority developed for the safety evaluation of SFR accidents. This enhancement is achieved by coupling AZORES with plant dynamics code ADYTUM in order to deal with more detailed thermal hydraulics behavior in reactor cooling system. The paper describes the coupling methodology, focusing on the thermal–hydraulic model as the first step of code development and its application. The coupling has been confirmed to be well achieved through a preliminary analysis of LOHS, comparing the results of coupled and single AZORES. The comparison of the analysis results between LOHS and LORL by the AZORES coupled with ADYTUM has been conducted to identify the key phenomena in accident progression. The comparison shows that the coolant leak area and the pressure drop at the piping of the reactor cooling system affect accident progression in LOHS. Consequently, the coupled AZORES is capable of predicting in-vessel thermal–hydraulic behavior in LOHRS successfully, reflecting accident conditions and plant design appropriately. [ABSTRACT FROM AUTHOR]

• Verification and validation for the accuracy of FASYS is performed. • Basic requirements of FASYS and PIRT in CEFR are established. • Bottom-up, top-down and uncertainty assessment are conducted. • Results indicate FASYS can simulate the accident processes of fast reactor. The fast reactor system analysis code FASYS is developed by CIAE for transient accident analysis in CEFR. By referring to the application scope of FASYS, the basic requirements of FASYS and PIRT in CEFR are established. A comprehensive verification and validation for the accuracy of FASYS is performed, which consist of bottom-up assessment, top-down assessment and uncertainty assessment. The bottom-up assessment applies the static analysis and dynamic analysis to verify the basic function of FASYS. The separate effect tests and integral effect tests are employed in the top-down assessment to validate the availability to simulate model and system. The calculated results confirms that the basic function and model in FASYS are reasonable and credible. The uncertainty assessment results show that FASYS can still calculate IET under the reasonable uncertainty of input parameters, indicating that the FASYS can simulate and analyze the transient accident processes of sodium-cooled fast reactor. [ABSTRACT FROM AUTHOR]

• Effect of interaction mode on local FCI checked experimentally using same couple of melt-coolant materials. • Significant cooling/re-condensation of steam bubbles by surrounding water pool confirmed at MI mode. • Impact of liquid volume delivered, water subcooling and pool depth is totally different within two modes. • Conversion efficiency at MI-mode is significantly lower compared to CI-mode. From the past analyses of Core Disruptive Accidents (CDA) for sodium-cooled fast reactors, it has been recognized that with the assumption of very pessimistic conditions, a large-scale molten-fuel pool, which contains sufficient amount of fuel to achieve the prompt criticality by fuel compaction, might be formed. Due to the potential entrapment of liquid sodium into the fuel pool, local Fuel-Coolant Interaction (FCI) in the pool is considered to be one of the various initial factors resulting in such compactive fluid motions. To ascertain the mechanisms underlying this interaction more reliably, in our earlier publication a number of simulated experiments was carried out by releasing a given amount of liquid water into a molten pool (comprised of a low-melting-point alloy) using the PMCI (Pressurization characteristics in Melt-Coolant Interaction) facility recently developed at the Sun Yat-sen University. In this work, motivated to acquire more enhanced understanding, the effect of interaction mode on local FCIs is further investigated by adding new experimental data within the Melt-Injection (MI) mode. From detailed analyses, it is found that except the existence of a similar impact of melt temperature, much difference in the influence of other experimental parameters such as the liquid volume delivered, water subcooling along with the pool depth is observed. In addition, the performed analyses in this work also confirm that compared to previous Coolant-Injection (CI) mode, the mechanical energy conversion efficiency is much lower within the MI-mode condition. This work provides a large palette of beneficial experimental data and evidence for the better understanding of CDAs and improved verifications of fast reactor safety accident analysis codes in China. [ABSTRACT FROM AUTHOR]